Attachment W to this Enclosure contains Security-Related Information to be Withheld from Public Disclosure In Accordance with 10 CFR 2.390(d)(1) Enclosure to L-PI-14-045 Northern States Power – Minnesota Prairie Island Nuclear Generating Plant Transition to 10 CFR 50.48(c) – NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants 2001 Edition Transition Report Revision 1 – April 2014 Attachment W contains Security-Related Information Withhold Under 10 CFR 2.390(d)(1) Upon removal of Attachment W, this Enclosure is uncontrolled.
Transcript
Attachment W to this Enclosure contains Security-Related Information to be Withheld from Public Disclosure In Accordance with 10 CFR 2.390(d)(1)
Enclosure to L-PI-14-045
Northern States Power – Minnesota
Prairie Island Nuclear Generating Plant
Transition to 10 CFR 50.48(c) – NFPA 805 Performance-Based Standard for Fire Protection for
Light Water Reactor Electric Generating Plants 2001 Edition
Transition Report Revision 1 – April 2014
Attachment W contains Security-Related Information
Withhold Under 10 CFR 2.390(d)(1)
Upon removal of Attachment W, this Enclosure is uncontrolled.
Attachment W contains Security-Related Information
Withhold Under 10 CFR 2.390
Northern States Power - Minnesota
Prairie Island Nuclear Generating Plant
Units 1 & 2
Transition to 10 CFR 50.48(c) - NFPA 805
Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001
Edition
Transition Report
Revision 1
April 2014
Attachment W contains Security-Related Information Withhold Under 10 CFR 2.390
Upon removal of Attachment W, this Enclosure is uncontrolled
Northern States Power - Minnesota NFPA 805 Transition Report
PINGP Page i – Revision 1
TABLE OF CONTENTS
Revision Status ............................................................................................................ iv
Summary of Changes .................................................................................................. vi
Executive Summary ..................................................................................................... ix
Acronym List ................................................................................................................ xi
4.6 Monitoring Program ......................................................................................... 43
4.6.1 Overview of NFPA 805 Requirements for the NFPA 805 Monitoring Program ................................................................................................... 43
4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program ................... 44
4.7 Program Documentation, Configuration Control, and Quality Assurance ........ 49
4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805 ........................................................................................................... 49
4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805 ................................................................................... 51
4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 .... 55
4.8 Summary of Results ......................................................................................... 56
4.8.1 Results of the Fire Area Review .............................................................. 56
4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase....................................................................................................... 57
4.8.3 Supplemental Information –Other Licensee Specific Issues .................... 57
A. NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements ............................................................................................... A-1
B. NEI 04-02 Table B-2 – Nuclear Safety Capability Assessment - Methodology Review ................................................................................................................ B-1
C. NEI 04-02 Table B-3 – Fire Area Transition ..................................................... C-1
D. NEI 04-02 Non-Power Operational Modes Transition ..................................... D-1
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E. NEI 04-02 Radioactive Release Transition ....................................................... E-1
F. Fire-Induced Multiple Spurious Operations Resolution ................................. F-1
G. Recovery Actions Transition ............................................................................ G-1
H. NFPA 805 Frequently Asked Question Summary Table ................................ H-1
I. Definition of Power Block ................................................................................... I-1
J. Fire Modeling V&V ............................................................................................. J-1
K. Existing Licensing Action Transition .............................................................. K-1
L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii)) ... L-1
M. License Condition Changes ............................................................................. M-1
N. Technical Specification Changes .................................................................... N-1
O. Orders and Exemptions .................................................................................... O-1
P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4) ........................................ P-1
Q. No Significant Hazards Evaluations ................................................................ Q-1
R. Environmental Considerations Evaluation ..................................................... R-1
S. Plant Modifications and Items for Implementation.......................................... S-1
T. Clarification of Prior NRC Approvals ................................................................ T-1
U. Internal Events PRA Quality ............................................................................. U-1
V. Fire PRA Quality ................................................................................................. V-1
W. Fire PRA Insights .............................................................................................. W-1
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Revision Status
The following table lists sections of the PINGP NFPA 805 Transition Report and Attachments and identifies their current revision status. This Revision Status section was added to the Transition Report in Revision 1.
Pages designated Revision “0” were included in the September 28, 2012 submittal and have not been changed.
Pages designated “Revision 1” have been updated for the May 1, 2014 Supplement and supersede the following:
• Pages in the September 28, 2012 submittal,
• Pages in Attachment U previously marked “Revised” that were included in the November 8, 2012 submittal addressing the Internal Flooding PRA peer review (ADAMS Accession No. ML12314A144), and
• Pages in Attachment W previously marked “Revised” that were included in the December 18, 2012 submittal that responded to NRC questions in support of Acceptance (ADAMS Accession No. ML12354A464).
Revised text is identified by a vertical bar in the right-hand margin except for Attachment W, which has been completely revised and does not include any change markings, and editorial changes (e.g., spelling or grammar corrections, changes in format or pagination), which are not marked.
Section Revision
Table of Contents 1
Revision Status 1
Summary of Changes 1
Executive Summary 1
Acronym List 1
1.0 Introduction 1
2.0 Overview of Existing Fire Protection Programs 0
3.0 Transition Process 0
4.0 Compliance with NFPA 805 Requirements 1
5.0 Regulatory Evaluation 1
6.0 References 1
Attachment A, NEI 04-02 Table B-1 1
Attachment B, NEI 04-02 Table B-2 1
Attachment C, NEI 04-02 Table B-3 1
Attachment D, Non Power Operational Modes 1
Northern States Power - Minnesota NFPA 805 Transition Report
Attachment T, Clarification of Prior NRC Approvals 0
Attachment U, Internal Events PRA Quality 1
Attachment V, Fire PRA Quality 1
Attachment W, Fire PRA Insights 1
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Summary of Changes
This Summary of Changes section was added to the Transition Report in Revision 1.
Revision Affected Section Change
0 All Initial submittal, September 2012
Revised
(November 8, 2012)
Attachment U, pages U-18, U-19
Updated to address final peer review of Internal Flooding PRA, including closure of Findings in Table U-2
Revised
(December 18, 2012)
Attachment W Revised to correct errors
1 Revision Status Added
Summary of Changes Added
Executive Summary Described Supplement
Acronym List Updated list
1.0, Introduction Described Supplement
4.1.1, Overview of Evaluation Process
Revised EC number
4.2.1.1, Compliance with NFPA Section 2.4.2
Revised to clarify gap analysis between NEI 00-01, Revision 1 and Revision 2; revised EC numbers
4.2.1.2, Safe and Stable Conditions for the Plant
Revised to delete references to PRISM, added control room HVAC discussion
4.2.1.3, Establishing Recovery Actions
Revised EC number; editorial correction
4.2.1.4, Evaluation of Multiple Spurious Operations
Editorial correction
4.3.2, Non-Power Operational Modes
Revised EC number
4.4.2, Radioactive Release Performance Criteria, Results of Evaluation Process
Revised EC number; deleted discussion about preparation of a fire strategy for the Containment Access Facility
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Revision Affected Section Change
4.5, Fire PRA and Performance Based Approaches
Described revision to Fire PRA and focused Peer Review for Hot Gas Layer methodology; also updated the Internal Flooding Peer Review and made an editorial correction
4.6.2, Overview of Post-Transition NFPA 805 Monitoring Program
Revised to clarify that 10 CFR 20 limits will not be exceeded during radioactive releases
4.8.2, Plant Modifications and Items to be Completed During Implementation
Revised EC number for Attachment S, and clarified that there are no outstanding plant modifications with respect to the Fire PRA model other than items in Attachment S; also revised reference to Table S-1 to refer to a completed modification
5.2.1, License Condition Changes
Revise to note that the RG 1.205 license condition was modified
5.5, Transition Implementation Schedule
Added sentence to indicate that Table S-3, item 20 will be completed after completion of the modifications in Table S-2
6.0, References Added correspondence subsequent to the September 28, 2012 submittal, including additional information to support NRC acceptance, the acceptance letter, and letters discussing the May 1, 2014 supplement; also updated Engineering Change (EC) list for the Supplement
Attachment A Added verification of technical bases for “Complies by Previous NRC Approval” statements; also, revised the Compliance Statements for sections 3.3.7, 3.4.3(a)(2), 3.5.6, 3.9.4, and 3.10.9
Attachment B Deleted references to PRISM and added references to Genesis; also updated references and Alignment Basis statements to reflect the revised NSCA model
Attachment C Updated for revised Fire PRA, VFDRs, FREs, recovery actions and modifications
Attachment D Revised EC number and updated to reflect revised NPO model
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Revision Affected Section Change
Attachment E Removed Fire Area (FA) 40, Maintenance Storage Shed, from radiologically controlled areas; also added FA 9 and incorporated several clarifications
Attachment F Updated to reflect additional expert panel review of electrical systems and to identify revised notebooks
Attachment G Revised to state that PINGP does not have a primary control station that meets RG 1.205; also revised Recovery Actions for revised Fire PRA, including new actions
Attachment H Updated FAQ list to add items 12-0063 and 12-0067
Attachment I Revised FA 40 description, deleted YARD items, and revised EC number
Attachment J Updated Fire Modeling V&V
Attachment M Revised License Condition to clarify that it does not apply to equivalency evaluations
Attachment S Revised Modifications for revised Fire PRA and additional fire modeling; added item to Table S-1, revised Table S-2, and revised Table S-3
Attachment U Revised for closure of Finding, and added discussion of planned focused-scope peer review for the new reactor coolant pump seal model
Attachment V Revised for closure of F&Os; deleted PRISM; added focused peer review of HGL methodology
Attachment W Revised Fire PRA results
Northern States Power - Minnesota Executive Summary
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Executive Summary
Northern States Power Company, a Minnesota corporation (NSPM) doing business as Xcel Energy, will transition the fire protection program for Prairie Island Nuclear Generating Plant Units 1 & 2 (PINGP) to a new Risk-Informed, Performance-Based (RI-PB) alternative per 10 CFR 50.48(c) which incorporates by reference National Fire Protection Association Standard 805 (NFPA 805). The licensing basis per 10 CFR 50.48(b) and 10 CFR 50, Appendix R, will be superseded.
The voluntary adoption of 10 CFR 50.48(c) by PINGP does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, General Design Criteria (GDC) 3, Fire Protection. However, compliance with the new rule establishes compliance with these sections. By letter dated June 22, 2011 (ADAMS Accession No. ML111740866), NSPM committed to submit a license amendment request by September 30, 2012, for PINGP to transition to 10 CFR 50.48(c).
The transition process consisted of a review and update of PINGP documentation, including the development of a Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850 as guidance. This Transition Report summarizes the transition process and results. This Transition Report contains information:
• Required by 10 CFR 50.48(c).
• Recommended by guidance document Nuclear Energy Institute (NEI) 04-02 Revision 2 and appropriate Frequently Asked Questions (FAQs).
• Recommended by guidance document Regulatory Guide 1.205 Revision 1.
Section 4 of the Transition Report provides a summary of compliance with the following NFPA 805 requirements:
• Fundamental Fire Protection Program Elements and Minimum Design Requirements.
• Nuclear Safety Performance Criteria, including:
o Nuclear Safety Capability Assessment, Safe and Stable Conditions for the Plant, Establishing Recovery Actions, Evaluation of Multiple Spurious Operations,
o Existing Engineering Equivalency Evaluations,
o Licensing Actions, and
o Fire Area Transitions.
• Non-Power Operational Modes.
• Radioactive Release Performance Criteria.
• Fire PRA and Performance-Based Approaches.
• Monitoring Program.
• Program Documentation, Configuration Control, and Quality Assurance.
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Section 5 of the Transition Report provides regulatory evaluations and associated attachments, including:
• Changes to the Fire Protection License Condition.
• Changes to Technical Specifications, Orders, and Exemptions.
• Determination of No Significant Hazards and evaluation of Environmental Considerations.
The attachments to the Transition Report include details to support the transition process and results.
Attachment H contains the approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this License Amendment Request.
Revision 1 to this Transition Report includes the following:
• Changes to reflect a revision to the Fire PRA, including changes in fire modeling, modifications, recovery actions, and calculated risk values,
• Changes to reflect revisions to the Nuclear Safety Capability Assessment (NSCA) and Non-Power Operations (NPO) models to replace PRISM software,
• Changes to the lists of modifications and implementation items based on additional fire modeling and engineering evaluations, and
• Changes to address generic industry issues addressed in Generic Requests for Additional Information (RAIs) and in revisions to the NEI 04-02 template that were more recent than Revision 1L used in the original submittal.
The entire PINGP NFPA 805 Transition Report document is included with this supplement, to simplify the review effort. Sections and Attachments to the original 2012 Transition Report that have been changed are identified in the Revision Status table, and are briefly described in the Summary of Changes. Revision 1 is not a complete revision to the Transition Report and many attachments have not been changed. Where changes in text or content have been made, the entire section or attachment is marked “Revision 1” in the footer and the affected text is identified by a vertical bar in the right-hand margin. Exceptions include Attachment W, which has been completely revised and does not include any change markings, and editorial changes (e.g., spelling or grammar corrections, format or pagination changes), which are not marked. Sections or attachments that have not been changed are marked “Revision 0.”
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Acronym List
AC Alternating Current ACUBE Advanced Cutset Upper Bound Estimator ADAMS Agencywide Documents Access and Management System ADS Automatic Depressurization System AEC Atomic Energy Commission AF Auxiliary Feedwater AFP Area Fire Plan AFW Auxiliary Feedwater System AHJ Authority Having Jurisdiction ANS American Nuclear Society ANSI American National Standards Institute APCSB Auxiliary Power Conversion Systems Branch AR Action Request ARP Alarm Response Procedure ASD Atmospheric Steam Dump ASEP Accident Sequence Evaluation Program ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials BE Basic Event BKR Breaker BTP Branch Technical Position BTU British Thermal Unit BWROG Boiling Water Reactor Owners Group CAAB Common Area of the Auxiliary Building CAF Containment Access Facility CAFTA Computer Aided Fault Tree Analysis CAS Central Alarm Station CB Control Building (Fire Area) CCDP Conditional Core Damage Probability CC Capability Category CC Component Cooling System CD Condensate System CD Control Damper CCF Common Cause Failure CCW Component Cooling Water CDF Core Damage Frequency CFAST Consolidated Fire and Smoke Transport CFCU Containment Fan Cooler Units CFR Code of Federal Regulations CL Cooling Water CLB Current Licensing Basis CLERP Conditional Large Early Release Probability CO2 Carbon Dioxide
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CPS Common Power Supply CR Control Room CRD Control Rod Drive CRDM Control Rod Drive Mechanism CS Containment Spray CSD Cold Shutdown CSR Cable Spreading Room CST Condensate Storage Tank CT Current Transformer CT External Circulating Water System CTEH Cooling Tower Equipment House CTPH Cooling Tower Pump House CV Control Valve CVCS Chemical and Volume Control System D(1-6) Emergency Diesel Generator DA Deluge Automatic DB Design Basis DBA Design Basis Accident DBD Design Basis Document DC Direct Current DDCLP Diesel Driven Cooling Water Pump DDCWP Diesel Driven Cooling Water pump DDFP Diesel Driven Fire Pump DID Defense-in-Depth DH Decay Heat DG Diesel Generator DM Deluge Manual DPS Dry Pipe System EC Engineering Change ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EDMG Earthquake Damage Mitigation Guide EEE Engineering Equivalency Evaluation EEEE Existing Engineering Equivalency Evaluation EF Error Factor EL Elevation EM Event Monitoring EOF Emergency Operating Facility EOP Emergency Operating Procedure EPM Engineering, Planning, and Management, Inc. EPRI Electrical Power Research Institute ERFBS Electrical Raceway Fire Barrier System ERO Emergency Response Organization ES Equipment Selection ESF Engineered Safety Features ESW Emergency Service Water
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EX Exterior (fire area) EXC Excluding ºF Degrees Fahrenheit F&O Fact and Observation FA Fire Area FAQ Frequently Asked Question FC Fire Compartment FDS Fire Dynamics Simulator FDT Fire Dynamics Tool FIVE Fire Induced Vulnerability Evaluation FHA Fire Hazards Analysis FIF Fire Ignition Frequency FM Factory Mutual FO Fuel Oil FP Fire Protection FPP Fire Protection Program FPE Fire Protection Engineer FPIE Full Power Internal Events FPRA Fire Probabilistic Risk Assessment FR Federal Register FRACQA Functional Responsibilities, Administrative Controls, and Quality
Assurance FRANX Fire Risk Analysis Software Tool FRE Fire Risk Evaluation FSAR Final Safety Analysis Report FSS Fire Scenario Selection ft Feet FV Fussell-Vesely FW Feedwater gal Gallon GDC General Design Criterion GL U.S. NRC Generic Letter GPM Gallons Per Minute HAD Heat Actuated Device HEAF High Energy Arc Fault HEP Human Error Probability HEPA High-Efficiency Particulate Air HFE Human Failure Event HGL Hot Gas Layer HLR High Level Requirement HRA Human Reliability Analysis HRE Higher Risk Evolution HRR Heat Release Rate HSDP Hot Shutdown Panel HSS High Safety Significant HVAC Heating, Ventilation, and Air Conditioning
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HX Heat Exchanger I&C Instrumentation and Controls ID Identification IE Initiating Event IEEE Institute of Electrical and Electronic Engineers IF Ignition Frequency IF Internal Flooding IN U.S. NRC Information Notice IPCEA Insulated Power Cable Engineers Association IPEEE Individual Plant Examination of External Events IPLD Integrated Plant Logic Diagram IS Intake Structure ISDS Ignition Source Data Sheet ISFSI Independent Spent Fuel Storage Installation ISLOCA Interfacing System Loss of Coolant Accident KSF Key Safety Function KV kilovolt KW kilowatt L Liter LA Licensing Action LAR License Amendment Request LCO Limiting Conditions for Operation LE LERF LERF Large Early Release Frequency LFS Limiting Fire Scenario LLC Limited Liability Company LLOCA Large Loss of Coolant Accident LLRW Low Level Radwaste LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LSELS Load Shed and Emergency Load Sequencer LSS Low Safety Significant m meter MAAP Modular Accident Analysis Program MCA Multi-Compartment Analysis MCB Main Control Board MCC Motor Control Center MCR Main Control Room MDAFWP Motor Driven Auxiliary Feedwater Pump MDFP Motor Driven Fire Pump MEFS Maximum Expected Fire Scenario MFW Main Feedwater MG Motor Generator MHIF Multiple High Impedance Fault min minute
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MOV Motor Operated Valve MQH McCaffrey, Quintiere, and Harkleroad MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MSO Multiple Spurious Operation MTTR Mean Time To Repair MV Motor Operated Valve MVSG Medium Voltage Switchgear N/A Not Applicable NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NEPIA Nuclear Energy Property Insurance Association (now NEIL) NIST National Institute of Standards and Technology NFPA National Fire Protection Association NMC Nuclear Management Company, LLC NPO Non-Power Operational NPP Nuclear Power Plant NPSH Net Positive Suction Head NRC U.S. Nuclear Regulatory Commission NSCA Nuclear Safety Capability Assessment NSEL Nuclear Safety Equipment List NSHC No Significant Hazards Consideration NSP Northern States Power NSP Non-Suppression Probability NSPC Nuclear Safety Performance Criteria NSPM Northern States Power - Minnesota NUMARC Nuclear Management and Resource Council NUREG US Nuclear Regulatory Commission Publication NUREG/CR NUREG document prepared by NRC contractors OAB Old Administration Building OCT Overcurrent Trip OMA Operator Manual Action OOS Out Of Service OPEX Operating Experience OS&Y Outside Screw and Yoke P&ID Piping and Instrumentation Diagram PA Preaction PA Public Address PAU Physical Analysis Unit PB Performance Based PBX Private Branch Exchange PC Primary Containment PCD PRA Change Database PCS Power Conversion System PCS Primary Control Station PDS Plant Damage State
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PH Pumphouse PI Project Instruction PINGP Prairie Island Nuclear Generating Plant – Units 1 & 2 PORV Power Operated Relief Valve POS Plant Operating State PPE Personal Protective Equipment PR Peer Review PRA Probabilistic Risk Assessment PRISM Plant Risk-Informed Systems Module PRM Plant Response Model PSA Probabilistic Safety Assessment PSF Performance Shaping Factor PVC Polyvinyl Chloride PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group QA Quality Assurance QNS Quantitative Screening QU Quantification RA Risk Assessment RAI Request for Additional Information RAW Risk Achievement Worth RC Reactor Coolant RCA Radiologically Controlled Area RCP Reactor Coolant Pump RCS Reactor Coolant System RES Nuclear Regulatory Commission – Office of Nuclear Regulatory Research RES Radiant Energy Shield RG U.S. NRC Regulatory Guide RH Residual Heat Removal RHR Residual Heat Removal RI-PB Risk-Informed, Performance-Based RIS Regulatory Issue Summary RPS Reactor Protection System RPV Reactor Pressure Vessel RRW Risk Reduction Worth RSP Remote Shutdown Panel RW River Water RWCU Reactor Water Cleanup RWST Refueling Water Storage Tank rx-yr Reactor year SAR Safety Analysis Report SBO Station Blackout SBDG Standby Diesel Generator SC Success Criteria SCBA Self-Contained Breathing Apparatus SCP Security Control Point
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SDC Shutdown Cooling SE Safety Evaluation SECY Commission Paper (NRC) SER Safety Evaluation Report SFP Spent Fuel Pool SFPE Society of Fire Protection Engineers SG Steam Generator SGTR Steam Generator Tube Rupture SI Safety Injection SLD Shutdown Logic Diagram SP Special Publication sq ft Square Feet SR Supporting Requirement SR Surveillance Requirement SRM Staff Requirements Memorandum SRV Safety Relief Valve SSA Safe Shutdown Analysis SSC Structures, Systems, and Components SSD Safe Shutdown SSE Safe Shutdown Earthquake SSEL Safe Shutdown Equipment List SSLD Safe Shutdown Logic Diagram SSO Single Spurious Operation STA Shift Technical Advisor SUT Startup Transformer SW Service Water SWGR Switchgear SWP Stairway Wet Pipe TB Turbine Building TBHX Thermal Barrier Heat Exchanger TD Turbine Driven TDAFP Turbine Driven Auxiliary Feedwater Pump TDAFW Turbine Driven Auxiliary Feedwater [Pump] T-H Thermal-Hydraulic TM Testing & Maintenance TSC Technical Support Center TS Technical Specification UAM Unreviewed Analysis Method (for Fire PRA) UFSAR Updated Final Safety Analysis Report UL Underwriters Laboratory USAR Updated Safety Analysis Report USC United States Code VAC Volts Alternating Current VC Chemical & Volume Control VCT Volume Control Tank V&V Verification and Validation
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VDC Volts Direct Current VFDR Variance From Deterministic Requirement WCAP Westinghouse Commercial Atomic Power WOG Westinghouse Owners Group WPS Wet Pipe Sprinkler yr Year ZOI Zone Of Influence
Northern States Power - Minnesota 1.0 Introduction
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1.0 INTRODUCTION
The Nuclear Regulatory Commission (NRC) has promulgated an alternative rule for fire protection requirements at nuclear power plants, 10 CFR 50.48(c), National Fire Protection Association Standard 805 (NFPA 805). NSPM is implementing the Nuclear Energy Institute methodology NEI 04-02, “Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program Under 10 CFR 50.48(c)” (NEI 04-02), to transition PINGP from its current fire protection licensing basis to the new requirements as outlined in NFPA 805. This report describes the transition methodology utilized and documents how PINGP complies with the new requirements.
Revision 1 to this Transition Report includes changes and corrections that NSPM determined were necessary after the NFPA 805 License Amendment Request (LAR) was submitted. The LAR was submitted for NRC review and approval on September 28, 2012 (ADAMS Accession No. ML12278A405). Additional information was submitted in a letter dated November 8, 2012 (ADAMS Accession No. ML12314A144), to provide final results of the Internal Flooding PRA peer review, and in a letter dated December 18, 2012 (ADAMS Accession No. ML12354A464), to address errors in the Fire PRA. The NRC subsequently accepted the LAR for review in a letter dated January 2, 2013 (ADAMS Accession No. ML13002A209).
During preparations for the initial NRC audit, NSPM determined that a revision to the Fire PRA was needed and in a letter dated May 3, 2013 (ADAMS Accession No. ML13126A115), committed to provide a supplement to the LAR by May 1, 2014. Revision 1 to this Transition Report includes the changes that are needed to satisfy this commitment. The revision to the Fire PRA resulted in changes to calculated risk values, fire modeling, recovery actions, and modifications, as provided in this LAR supplement. In addition, Revision 1 to this Transition Report includes changes to the Nuclear Safety Capability Assessment (NSCA) and Non-Power Operations (NPO) models that resulted from the replacement of PRISM computer software. The modifications and items for implementation were also revised based on additional fire modeling and engineering reviews.
Other reviews were determined to be necessary and are addressed in this supplement. An additional expert panel review was conducted for electrical systems to identify Multiple Spurious Operations (MSO) concerns. Also, a focused-scope peer review was performed for a hot gas layer (HGL) methodology not previously used.
This supplement also addresses generic industry issues that have been identified since submittal of the LAR. Issues identified as Generic Requests for Additional Information (RAIs) as of January 2014 are addressed. This resulted in a change to the License Condition and a number of other clarifications. Also, changes to the NEI 04-02 template subsequent to Revision 1L, which was the basis for the original LAR submittal, have been considered through Revision 1P.
To simplify the review effort, the entire PINGP NFPA 805 Transition Report document is included with this supplement. Sections and Attachments to the original 2012 Transition Report that have been changed are identified in the Revision Status table, and are briefly described in the Summary of Changes. Revision 1 is not a complete revision to the Transition Report and many attachments have not been changed. Where changes
Northern States Power - Minnesota 1.0 Introduction
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in text or content have been made, the entire section or attachment is marked “Revision 1” in the footer and the affected text is identified by a vertical bar in the right-hand margin. Exceptions include Attachment W, which has been completely revised and does not include any change markings, and editorial changes (e.g., spelling or grammar corrections, changes in format or pagination), which are not marked. Sections or attachments that have not been changed are marked “Revision 0.”
1.1 Background
1.1.1 NFPA 805 – Requirements and Guidance
On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes new Risk-Informed, Performance-Based (RI-PB) fire protection requirements. 10 CFR 50.48(c) incorporates by reference, with exceptions, the National Fire Protection Association’s NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants – 2001 Edition, as a voluntary alternative to 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.
As stated in 10 CFR 50.48(c)(3)(i), any licensee’s adoption of a RI-PB program that complies with the rule is voluntary. This rule may be adopted as an acceptable alternative method for complying with either 10 CFR 50.48(b), for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979, or 10 CFR 50.48(f), plants shutdown in accordance with 10 CFR 50.82(a)(1).
NEI developed NEI 04-02 to assist licensees in adopting NFPA 805 and making the transition from their current fire protection licensing basis to one based on NFPA 805. The NRC issued Regulatory Guide (RG) 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants, which endorses NEI 04-02, with exceptions, in December 2009.1
A depiction of the primary document relationships is shown in Figure 1-1:
1 Where referred to in this document NEI 04-02 is Revision 2 and RG 1.205 is Revision 1.
Northern States Power - Minnesota 1.0 Introduction
Nuclear Management Company (NMC) submitted a letter of intent to the NRC on November 30, 2005 (ADAMS Accession No. ML053460342) for PINGP to adopt NFPA 805 in accordance with 10 CFR 50.48(c). NSPM has subsequently assumed responsibility for actions and commitments previously submitted by NMC.
By letter dated September 7, 2006 (ADAMS Accession No. ML061500035), the NRC granted a three year enforcement discretion period. In accordance with NRC Enforcement Policy, the enforcement discretion period will continue until the NRC approval of the license amendment request (LAR) is completed.
In accordance with SECY-11-0061, in a letter dated June 22, 2011 (ADAMS Accession No. ML111740866), NSPM committed to submit a license amendment request no later than September 30, 2012, for PINGP to transition to 10 CFR 50.48(c). By letter dated July 29, 2011, (ADAMS Accession No. ML112010417), the NRC acknowledged the application date for PINGP and granted an extension of enforcement discretion.
Northern States Power - Minnesota 1.0 Introduction
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1.1.2.2 Transition Process
The transition to NFPA 805 includes the following high level activities:
• A new fire safe shutdown analysis;
• A new Fire Probabilistic Risk Assessment (PRA) using NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, as guidance and a revision to the Internal Events PRAs to support the Fire PRAs; and
• Completion of activities required to transition the pre-transition Licensing Basis to 10 CFR 50.48(c) as specified in NEI 04-02 and RG 1.205.
1.2 Purpose
The purpose of the Transition Report is as follows:
1) Describe the process implemented to transition the current fire protection program to compliance with the additional requirements of 10 CFR 50.48(c);
2) Summarize the results of the transition process;
3) Explain the bases for conclusions that the fire protection program complies with 10 CFR 50.48(c) requirements;
4) Describe the new fire protection licensing basis; and
5) Describe the configuration management processes used to manage post-transition changes to the plant and the Fire Protection Program, and resulting impact on the Licensing Basis.
Northern States Power - Minnesota 2.0 Overview of Existing Fire Protection Program
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2.0 OVERVIEW OF EXISTING FIRE PROTECTION PROGRAM
2.1 Current Fire Protection Licensing Basis
PINGP Unit 1 was licensed to operate on August 9, 1973 and Unit 2 was licensed to operate on October 29, 1974. As a result, the PINGP fire protection program is based on compliance with 10 CFR 50.48(a), 10 CFR 50.48(b), 10 CFR 50 Appendix R, and the following License Condition:
NSPM, PINGP Renewed Operating Licenses Nos. DPR-42 (Unit 1) and DPR-60 (Unit 2) both include License Condition 2.C.(4), “Fire Protection,” which states:
“NSPM shall implement and maintain in effect all provisions of the approved fire protection program as described and referenced in the Updated Safety Analysis Report for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in Safety Evaluation Reports dated February 14, 1978, September 6, 1979, April 21, 1980, December 29, 1980, July 28, 1981, October 27, 1989, and October 6, 1995, subject to the following provision:
NSPM may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.”
2.2 NRC Acceptance of the Fire Protection Licensing Basis
To conform to NRC guidelines issued prior to November 1980, Northern States Power performed a fire hazard analysis which analyzed the PINGP fire protection program against the guidance of Appendix A to Branch Technical Position (BTP) Auxiliary Power Conversion Systems Branch (APCSB) 9.5-1. The results of this analysis, in addition to proposed modifications and additions to the fire protection program, were communicated to the NRC by letters dated March 11, 1977, July 5, 1977, May 18, 1978, June 22, 1978, January 2, 1979, March 9, 1979, and May 2, 1979. Furthermore this analysis served as the basis for the Appendix A to BTP APCSB 9.5-1 safety evaluation dated September 6, 1979, and the associated License Amendment Nos. 39 (Unit 1) and 33 (Unit 2), which implemented fire protection technical specifications and added a license condition for the completion of fire protection modifications, submittal of additional information, and implementation of administrative controls.
The current Fire Protection License Condition quoted above identifies a number of NRC Safety Evaluation Reports (SERs) and approval letters which are briefly summarized below. Any provisions of these documents that are to be transitioned to the NFPA 805 fire protection program are identified in other sections and/or attachments to this Transition Report.
NRC letter dated February 14, 1978
In a letter dated February 14, 1978, the NRC issued License Amendment Nos. 26 (Unit 1) and 20 (Unit 2) which revised the PINGP Technical Specifications (TSs) to add Limiting Conditions for Operation (LCOs) and Surveillance Requirements (SRs) for fire protection equipment and instrumentation, and Administrative Controls related to fire protection.
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NRC SER dated September 6, 1979
In the SER for License Amendment Nos. 39 (Unit 1) and 33 (Unit 2) dated September 6, 1979, the NRC added a license condition relating to the completion of facility modifications and implementation of administrative controls, and approved TS changes regarding detector functional test frequency, valve position verification, fire brigade size, and administrative responsibilities for fire protection and training.
NRC Approval dated April 21, 1980
In a letter dated April 21, 1980, the NRC approved fourteen fire protection modifications that were previously described in the September 6, 1979 SER, and that had been designated as requiring additional information prior to implementation.
NRC Approval dated December 29, 1980
In a letter dated December 29, 1980, the NRC approved additional fire protection modifications that had previously been described in the September 6, 1979 SER, and that had been designated as requiring additional information prior to implementation. The December 29 NRC letter discussed fire barrier penetration seal upgrades, structural steel member coating in the vicinity of the lube oil reservoir, installation of fire dampers, a fire barrier enclosure for the motor driven fire pump, and fire detector response capabilities.
NRC SER dated July 28, 1981
In the SER for License Amendment Nos. 49 (Unit 1) and 43 (Unit 2) dated July 28, 1981, the NRC approved modifications to the TS LCOs and surveillance requirements for Fire Detection and Protection Systems, and accompanying Bases descriptions. These changes reflected modifications to fire protection equipment, structures, testing requirements, and administrative controls.
NRC SER dated October 27, 1989
In the SER for License Amendment Nos. 91 (Unit 1) and 84 (Unit 2) dated October 27, 1989, the NRC approved numerous changes throughout the TS in support of a human error reduction program. These changes included the reorganization and standardization of some TS sections, including fire protection program requirements, to achieve consistency and uniformity throughout the TS and minimize the potential for confusion.
NRC SER dated October 6, 1995
In the SER for License Amendment Nos. 120 (Unit 1) and 113 (Unit 2) dated October 6, 1995, the NRC approved TS changes to remove Fire Protection Program requirements from the TS. In accordance with Generic Letter 86-10, fire protection program elements were removed from the TS and the NRC-approved Fire Protection Program and major commitments, including the fire hazards analysis, were incorporated by reference into the USAR. A new fire protection license condition was added.
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Exemptions
The following is a list of the exemptions that have been granted by the NRC from the requirements of Appendix R to 10 CFR 50, Sections III.G, III.J, and III.O:
• An exemption from Section III.G.3.b for lack of a fixed fire suppression system in the Control Room, Units 1 and 2, Fire Area 13 (NRC SER dated February 2, 1983).
• An exemption from Subsection III.G.2 for lack of twenty feet of separation free of intervening combustibles or one hour fire rated barriers between redundant trains needed for safe shutdown in the “A” Train Hot Shutdown Panel, Instrument Air and Auxiliary Feedwater Pump Rooms, Units 1 and 2, Fire Area 31 (NRC SER dated May 4, 1983).
• An exemption from Subsection III.G.2 for lack of twenty feet of separation free of intervening combustibles or one hour fire rated barriers between redundant trains needed for safe shutdown in the “B” Train Hot Shutdown Panel, Instrument Air and Auxiliary Feedwater Pump Rooms, Units 1 and 2, Fire Area 32 (NRC SER dated May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Operating Level, Unit 1, Fire Area 60 (NRC SER dated May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Operating Level, Unit 2, Fire Area 75 (NRC SER dated May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Normal Switchgear Room, Unit 1, Fire Area 37 (NRC SER dated May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Ground Floor Level, Unit 1, Fire Area 58 (NRC SER dated January 9, 1984).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Ground Floor Level, Unit 2, Fire Area 73 (NRC SER dated January 9, 1984).
• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Mezzanine Level, Unit 1, Fire Area 59 (NRC SER dated January 9, 1984).
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• An exemption from Section III.G.2 for lack of an automatic fire suppression system in the Auxiliary Building Mezzanine Level, Unit 2, Fire Area 74 (NRC SER dated January 9, 1984).
• An exemption from Section III.G.2 for the lack of twenty feet of separation free of intervening combustibles between redundant trains needed for safe shutdown in the Containment, Units 1 and 2, Fire Areas 1 and 71, including a commitment to install a one-hour fire barrier to protect the cabling for one division of the pressurizer level transmitters in Unit 2 (NRC SER dated July 31, 1984). As described in Attachment K, different methods of protection for this cabling are provided in Unit 1 and Unit 2.
• An exemption from Section III.O for a reactor coolant pump lube oil collection system that does not drain to a vented closed container that can hold the entire lube oil system inventory, but instead is piped to a sump inside Containment and then is pumped to a closed vented container located in the Auxiliary Building; Units 1 and 2, Containment Fire Areas 1 and 71 (NRC SER dated July 31, 1984).
• An exemption from Section III.G.1 to allow operators to remove fuses from Power Operated Relief Valve (PORV) control circuits to preclude inadvertent valve operation in the event of a control room evacuation; this is considered a repair to ensure that one train of safe shutdown equipment remains operable which is contrary to Section III.G.1, Control Room, Units 1 and 2, Fire Area 13 (NRC SER dated February 21, 1995).
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3.0 TRANSITION PROCESS
3.1 Background
Section 4.0 of NEI 04-02 describes the process for transitioning from compliance with the current fire protection licensing basis to the new requirements of 10 CFR 50.48(c). NEI 04-02 contains the following steps:
1) Licensee determination to transition the licensing basis and devote the necessary resources to it;
2) Submit a Letter of Intent to the NRC stating the licensee’s intention to transition the licensing basis in accordance with a tentative schedule;
3) Conduct the transition process to determine the extent to which the current fire protection licensing basis supports compliance with the new requirements and the extent to which additional analyses, plant and program changes, and alternative methods and analytical approaches are needed;
4) Submit a LAR;
5) Complete transition activities that can be completed prior to the receipt of the License Amendment;
6) Receive a Safety Evaluation; and
7) Complete implementation of the new licensing basis, including completion of modifications identified in Attachment S.
3.2 NFPA 805 Process
Section 2.2 of NFPA 805 establishes the general process for demonstrating compliance with NFPA 805. This process is illustrated in Figure 3-1. It shows that except for the fundamental fire protection requirements, compliance can be achieved on a fire area basis either by deterministic or RI-PB methods. Consistent with the guidance in NEI 04-02, NSPM has implemented the NFPA 805 Section 2.2 process by first determining the extent to which its current fire protection program supports findings of deterministic compliance with the requirements in NFPA 805. RI-PB methods are being applied to the requirements for which deterministic compliance could not be shown.
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Figure 3-1 NFPA 805 Process [NEI 04-02 Figure 3-1 based on Figure 2-2 of NFPA 805]2
3.3 NEI 04-02 – NFPA 805 Transition Process
NFPA 805 contains technical processes and requirements for a RI-PB fire protection program. NEI 04-02 was developed to provide guidance on the overall process (programmatic, technical, and licensing) for transitioning from a traditional fire protection licensing basis to a new RI-PB method based upon NFPA 805, as shown in Figure 3-2.
2 Note: 10 CFR 50.48(c) does not incorporate by reference Life Safety and Plant Damage/Business
Interruption goals, objectives and criteria. See 10 CFR 50.48(c) for specific exceptions to the incorporation by reference of NFPA 805.
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Section 4.0 of NEI 04-02 describes the detailed process for assessing a fire protection program for compliance with NFPA 805, as shown in Figure 3-2.
Figure 3-2 Transition Process (Simplified) [based on NEI 04-02 Figure 4-1]
3.4 NFPA 805 Frequently Asked Questions (FAQs)
The NRC has worked with NEI and two Pilot Plants (Oconee Nuclear Station and Harris Nuclear Plant) to define the licensing process for transitioning to a new licensing basis under 10 CFR 50.48(c) and NFPA 805. Both the NRC and the industry recognized the need for additional clarifications to the guidance provided in RG 1.205, NEI 04-02, and the requirements of NFPA 805. The NFPA 805 FAQ process was jointly developed by NEI and NRC to facilitate timely clarifications of NRC positions. This process is described in a letter from the NRC dated July 12, 2006, to NEI (ML061660105) and in Regulatory Issues Summary (RIS) 2007-19, Process for Communicating Clarifications
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of Staff Positions Provided in RG 1.205 Concerning Issues Identified during the Pilot Application of NFPA Standard 805, dated August 20, 2007 (ML071590227).
Under the FAQ Process, transition issues are submitted to the NEI NFPA 805 Task Force for review, and subsequently presented to the NRC during public FAQ meetings. Once the NEI NFPA 805 Task Force and NRC reach agreement, the NRC issues a memorandum to indicate that the FAQ is acceptable. NEI 04-02 will be revised to incorporate the approved FAQs. This is an on-going revision process that will continue through the transition of NFPA 805 transition plants. Final closure of the FAQs will occur when future revisions of RG 1.205, endorsing the related revisions of NEI 04-02, are approved by the NRC. It is expected that additional FAQs will be written and existing FAQs will be revised as plants continue NFPA 805 transition after the Pilot Plant Safety Evaluations.
Attachment H contains the list of approved FAQs not yet incorporated into the endorsed revision of NEI 04-02. These FAQs have been used to clarify the guidance in RG 1.205, NEI 04-02, and the requirements of NFPA 805 and in the preparation of this LAR.
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4.0 COMPLIANCE WITH NFPA 805 REQUIREMENTS
4.1 Fundamental Fire Protection Program and Design Elements
The Fundamental Fire Protection Program and Design Elements are established in Chapter 3 of NFPA 805. Section 4.3.1 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis and plant configuration meets these criteria and for identifying the fire protection program changes that would be necessary for compliance with NFPA 805. NEI 04-02 Appendix B-1 provides guidance on documenting compliance with the program requirements of NFPA 805 Chapter 3.
4.1.1 Overview of Evaluation Process
The comparison of the PINGP Fire Protection Program to the requirements of NFPA 805 Chapter 3 was performed and documented in PINGP Engineering Evaluation EC 23309 entitled “NFPA 805 LAR Supplement Attachment A (Table B-1).” Engineering Evaluation EC 23309 used the guidance contained in NEI 04-02, Section 4.3.1 and Appendix B-1 (See Figure 4-1).
Each section and subsection of NFPA 805 Chapter 3 was reviewed against the current fire protection program. Upon completion of the activities associated with the review, the following compliance statement(s) was used:
• Complies - For those sections/subsections determined to meet the specific requirements of NFPA 805.
• Complies with Item for Implementation – For those sections/subsections determined to meet the requirements of NFPA 805 upon completion of an item as identified in Attachment S.
• Complies with Clarification - For those sections/subsections determined to meet the requirements of NFPA 805 with clarification.
• Complies by previous NRC approval - For those sections/subsections where the specific NFPA 805 Chapter 3 requirements are not met but previous NRC approval of the configuration exists.
• Complies with use of Existing Engineering Equivalency Evaluations (EEEEs) - For those sections/subsections determined to be equivalent to the NFPA 805 Chapter 3 requirements as documented by engineering analysis.
• Submit for NRC Approval - For those sections/subsections for which approval is sought in this LAR submittal in accordance with 10 CFR 50.48(c)(2)(vii). A summary of the bases of acceptability is provided (See Attachment L for details).
In some cases multiple compliance statements have been assigned to a specific NFPA 805 Chapter 3 section/subsection. Where this is the case, each compliance/compliance basis statement clearly references the corresponding requirement of NFPA 805 Chapter 3.
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Figure 4-1 - Fundamental Fire Protection Program and Design Elements Transition Process [Based on NEI 04-02 Figure 4-2]3
3 Figure 4-1 depicts the process used during the transition and therefore contains elements (i.e., open items) that represent interim resolutions. Additional detail
on the transition of EEEEs is included in Section 4.2.2.
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4.1.2 Results of the Evaluation Process
4.1.2.1 NFPA 805 Chapter 3 Requirements Met or Previously Approved by the NRC
Attachment A contains the NEI 04-02 Table B-1, Transition of Fundamental Fire Protection Program and Design Elements. This table provides the compliance basis for the requirements in NFPA 805 Chapter 3. Except as identified in Section 4.1.2.3, Attachment A demonstrates that the fire protection program at PINGP either:
• Complies directly with the requirements of NFPA 805 Chapter 3,
• Complies with clarification with the requirements of NFPA 805 Chapter 3,
• Complies through the use of existing engineering equivalency evaluations which are valid and of appropriate quality, or
• Complies with a previously NRC approved alternative to NFPA 805 Chapter 3 and therefore the specific requirement of NFPA 805 Chapter 3 is supplanted.
NFPA 805 Section 3.1 states in part, “Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein.” In some cases prior NRC approval of an NFPA 805 Chapter 3 program attribute may be unclear. NSPM requests that the NRC concur with their finding of prior approval for the following sections of NFPA 805 Chapter 3:
• None.
Although there are no NFPA 805 Chapter 3 requirements requiring clarification, Attachment T includes a request to accept clarification of prior NRC approval of an exemption to an NFPA 805 Chapter 4 requirement (NFPA 805 Section 4.2.3.1).
4.1.2.3 NFPA 805 Chapter 3 Requirements Not Met and Not Previously Approved by NRC
The following sections of NFPA 805 Chapter 3 are not specifically met nor do previous NRC approvals of alternatives exist:
• 3.5.16 – Approval is requested for the use of fire protection water for other purposes not related to fire protection.
The specific deviation and a discussion of how the alternative satisfies 10 CFR 50.48(c)(2)(vii) requirements is provided in Attachment L. NSPM requests NRC approval of these performance-based methods.
4.1.3 Definition of Power Block and Plant
Where used in NFPA 805 Chapter 3 the terms “Power Block” and “Plant” refer to structures that have equipment required for nuclear plant operations, such as Containment, Auxiliary Building, Service Building, Control Building, Fuel Building, Radioactive Waste, Water Treatment, Turbine Building, and intake structures or structures that are identified in the facility’s pre-transition licensing basis.
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All structures within the PINGP Owner Controlled Area were reviewed to determine the potential impact of fire on the nuclear safety and radioactive release criteria described in Section 1.5 of NFPA 805. This was accomplished by identifying the structures that contain either:
• Equipment that could affect
o Plant operation for power generation
o Equipment important to safety
o Ability to maintain nuclear safety performance criteria in the event of a fire
OR
• Radioactive materials that could potentially be released in the event of a fire.
The determination of structures defined as the power block was completed in PINGP Engineering Evaluation EC 23946, entitled “NFPA 805 LAR Supplement Attachment I – Power Block Definition.”
These structures are listed in Attachment I and define the “power block” and “plant.”
4.2 Nuclear Safety Performance Criteria
The Nuclear Safety Performance Criteria are established in Section 1.5 of NFPA 805. Chapter 4 of NFPA 805 provides the methodology to determine the fire protection systems and features required to achieve the performance criteria outlined in Section 1.5. Section 4.3.2 of NEI 04-02 provides a systematic process for determining the extent to which the pre-transition licensing basis meets these criteria and for identifying any necessary fire protection program changes. NEI 04-02, Appendix B-2 provides guidance on documenting the transition of Nuclear Safety Capability Assessment Methodology and the Fire Area compliance strategies.
The Nuclear Safety Capability Assessment (NSCA) Methodology review consists of four processes:
• Establishing compliance with NFPA 805 Section 2.4.2
• Establishing the Safe and Stable Conditions for the Plant
• Establishing Recovery Actions
• Evaluating Multiple Spurious Operations
The methodology for demonstrating reasonable assurance that a fire during non-power operational (NPO) modes will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition is an additional requirement of 10 CFR 50.48(c) and is addressed in Section 4.3.
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“The purpose of this section is to define the methodology for performing a nuclear safety capability assessment. The following steps shall be performed: (1) Selection of systems and equipment and their interrelationships necessary to
achieve the nuclear safety performance criteria in Chapter 1 (2) Selection of cables necessary to achieve the nuclear safety performance
criteria in Chapter 1 (3) Identification of the location of nuclear safety equipment and cables (4) Assessment of the ability to achieve the nuclear safety performance criteria
given a fire in each fire area”
The NSCA methodology review evaluated PINGP’s post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 1 (ML050310295), Chapter 3, “Deterministic Methodology,” as discussed in Appendix B-2 of NEI 04-02. The methodology is depicted in Figure 4-2 and consisted of the following activities:
• Each specific section of NFPA 805 2.4.2 was correlated to the corresponding section of Chapter 3 of NEI 00-01 Revision 1. Based upon the content of the NEI 00-01 methodology statements, a determination was made of the applicability of the section to the station.
• The plant-specific methodology was compared to applicable sections of NEI 00-01 and one of the following alignment statements and its associated basis were assigned to the section:
o Aligns
o Aligns with intent
o Not in Alignment
o Not in Alignment, but Prior NRC Approval
o Not in Alignment, but no adverse consequences
• For those sections that do not align, an assessment was made to determine if the failure to maintain strict alignment with the guidance in NEI 00-01 could have adverse consequences. Since NEI 00-01 is a guidance document, portions of its text could be interpreted as ‘good practice’ or intended as an example of an efficient means of performing the analyses. If the section has no adverse consequences, these sections of NEI 00-01 can be dispositioned without further review.
The comparison of the PINGP post-fire SSA to NEI 00-01, Revision 1, Chapter 3 (NEI 04-02 Table B-2) was performed and documented in PINGP Engineering Evaluation EC 23310, “NFPA 805 LAR Supplement Attachment B (Table B-2).”
In addition, a review of NEI-00-01, Revision 2 (ML091770265), Chapter 3, was conducted to identify the substantive changes from NEI 00-01, Revision 1 that are applicable to an NFPA 805 fire protection program. This review was performed and documented in PINGP Engineering Evaluation EC 23408, “NFPA 805 LAR Supplement NEI 00-01 Gap Analysis.”
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Results from Evaluation Process
The method used to perform the post-fire SSA with respect to selection of systems and equipment, selection of cables, and identification of the location of equipment and cables, either meets the NRC endorsed guidance from NEI 00-01, Revision 1, Chapter 3 (as supplemented by the gap analysis) directly or meets the intent of the endorsed guidance with adequate justification as documented in Attachment B with the following exceptions:
• Attachment B Section 3.1.1.4: As an exception to this section, the PINGP Fire Protection Program is transitioning an existing approved licensing action (exemption). This licensing action allows a “repair action” to assure isolation of pressurizer PORVs for a fire occurring in the control room or relay room (Fire Areas 13 and 18 respectively), that could cause spurious operation of the PORV isolation valves. Therefore, this section is “Not in Alignment, but Prior NRC Approval.” The details for this licensing action can be found in Attachments K and T.
• Attachment B Section 3.4.1.6: As an exception to this section, the PINGP Fire Protection Program is transitioning an existing approved licensing action (exemption) for oil collection system variances for Fire Areas 1 and 71 (containment). Therefore, this section is “Not in Alignment, but Prior NRC Approval.” The details for this existing licensing action can be found in Attachment K.
Figure 4-2 – Summary of Nuclear Safety Methodology Review Process (FAQ 07-0039)
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Comparison to NEI 00-01 Revision 2
An additional review was performed of NEI 00-01, Revision 2, Chapter 3, for specific substantive changes in the guidance from NEI 00-01, Revision 1 that are applicable to an NFPA 805 transition. The results of this review are summarized below:
• Post fire manual operation of rising stem valves in the fire area of concern (NEI 00-01 Section 3.2.1.2)
NSPM has performed reviews of all recovery actions, and has validated that there are no recovery actions requiring the use of rising stem valves where the credited valve has been exposed to the fire.
• Analysis of open circuits on high voltage (e.g., 4.16 kV) ammeter current transformers (NEI 00-01 Section 3.5.2.1)
NSPM Engineering Evaluation EC 20819, “Technical Evaluation Associated with Open Circuiting of Current Transformers Contained in 13.8 kV, 4.16 kV, and 480 V Switchgear,” demonstrated that although there are current transformers susceptible to fire-induced faults, including the highly unlikely open circuit fault, none of the current transformers are of concern for secondary fires.
• Analysis of control power for switchgear with respect to breaker coordination (NEI 00-01 Section 3.5.2.4)
DC control power required to maintain switchgear breaker coordination was analyzed in PINGP Engineering Evaluation EC 21264, “Loss of DC Control Power Common Enclosure Analysis for NFPA 805,” for both credited power supplies and non-credited power supplies. The analysis included both the common power supply as well as the common enclosure aspects of the loss of control power. Several modifications have been identified in Attachment S, Table S-2, to preclude coordination and circuit protection concerns resulting from this fire-induced failure sequence.
4.2.1.2 Safe and Stable Conditions for the Plant
Overview of Process
The nuclear safety goals, objectives and performance criteria of NFPA 805 allow more flexibility than the previous deterministic programs based on 10 CFR 50 Appendix R and NUREG-0800, Section 9.5-1 (and NEI 00-01, Chapter 3) since NFPA 805 only requires the licensee to maintain the fuel in a safe and stable condition rather than achieve and maintain cold shutdown.
NFPA 805, Section 1.6.56, defines Safe and Stable Conditions as follows
“For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling.”
The nuclear safety goal of NFPA 805 requires "...reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from
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achieving and maintaining the fuel in a safe and stable condition" without a specific reference to a mission time or event coping duration.
For the plant to be in a Safe and Stable condition, it may not be necessary to perform a transition to cold shutdown as currently required under 10 CFR 50, Appendix R. Therefore, the unit may remain at or below the temperature defined by a hot standby/hot shutdown plant operating state for the event.
Results
Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses.
• At-Power analysis, Mode 1 through Mode 3. This analysis is discussed in Section 4.2.4.
• Non-Power analysis, which includes Mode 4 and below. This analysis is discussed in Section 4.3.
The NFPA 805 licensing basis for PINGP for a Safe and Stable condition in the event of a fire starting with the reactor in at-power operating Modes 1, 2, or 3 (Power Operation, Startup, or Hot Standby, respectively) is to maintain Safe and Stable conditions in Hot Standby without Residual Heat Removal (RHR). PINGP will maintain Hot Standby conditions until a decision is made to either place the reactor in a non-power operating mode, i.e., Hot Shutdown (Mode 4) or Cold Shutdown (Mode 5), or to return to power operations. Determination of the final state will be based upon the extent of the fire damage, the inventory remaining in the Refueling Water Storage Tank (RWST), the ability to provide makeup water to the RWST, and the ability to re-establish inventory in the Condensate Storage Tank (CST) or realign Auxiliary Feedwater (AFW) to its alternate source (cooling water system).
Mission Time
A PINGP thermal-hydraulic analysis was performed for a mission time of 24 hours to assure that safe and stable conditions can be achieved within that time period. This mission time ensures that sufficient time is available for the Emergency Response Organization to respond to the event, assess the extent of fire damage, and assist the plant operating staff with maintaining Safe and Stable conditions or transitioning the plant to a non-power operating mode.
To sustain Safe and Stable conditions, Key Safety Functions are met as follows:
• Reactivity and Inventory Control
The reactor design ensures that Keff < 0.99 can be achieved by use of the control rods from any operating mode. Subsequent injection (using Charging or Safety Injection Pumps) of soluble poison can be used to assure continuation of Mode 3, Hot Standby, under all circumstances. The charging system and the Safety Injection system will remain available beyond the mission time for Safe and Stable. The RWST is the credited source of borated water and is capable of providing water for at least 38 hours, per PINGP Engineering Evaluation EC 20736, “Reactivity Control.” Operator actions to establish makeup sources of
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inventory to the RWST are described in existing plant procedure C12.5, “Boron Concentration Control.”
• Decay Heat Removal
One or both steam generators, as well as a motor driven or turbine driven AFW pump will remain available without additional actions to provide symmetrical or asymmetrical decay heat removal beyond the mission time for Safe and Stable. The CST is the initial source for the AFW pumps. Per Engineering Evaluation EC 20738, “Decay Heat Removal,” the CST will provide a continuous water supply for the AFW pumps for 20 hours. Beyond 20 hours, the CST can be refilled or the AFW pumps can be re-aligned to the cooling water system to provide an unlimited water source. This realignment is accomplished through existing plant procedures 1(2)E-1, “Loss of Reactor or Secondary Coolant,” and C28.1, AOP2, “Loss of Condensate Supply to Auxiliary Feedwater Pump Suction.”
• Vital Auxiliaries – Power and Support Systems
The Emergency Diesel Generators (EDGs) and Diesel Driven Cooling Water Pumps (DDCLPs) have an on-site fuel oil supply that will last for 14 days, assuming one EDG on each unit and one DDCLP, or 7 days if both EDGs are operating for each unit and both DDCLPs are operating. Offsite sources of fuel oil are available to replenish fuel oil levels if needed via established contracts.
Control room heating, ventilating, and air conditioning (HVAC) can be lost in several fire areas due to the loss of instrument air and other fire-induced HVAC component damage. EC 23925, “NFPA 805 LAR Supplement – Control Room HVAC Evaluation,” demonstrates that control room temperatures will remain below equipment limits for up to 36 hours with actions taken only within the control room itself. However, a portable fan may be installed in the control room prior to 36 hours to allow temperatures to remain below equipment limits indefinitely. If required, the portable fan will normally be powered by a designated welding receptacle or in cases where the welding receptacle power is lost due to the fire, by a 480 VAC portable generator located outside of the building (Reference Attachment S-3, Item 63, for additional details).
If conditions warrant placing the plant in Hot Shutdown (Mode 4) or Cold Shutdown (Mode 5), NSPM will initiate operation of the RHR System. Although the RHR system is not required for maintaining safe and stable conditions, the RHR system is included in the “at power” Nuclear Safety Capability Assessment (NSCA) Genesis model to demonstrate its availability for transition. Initiation of RHR system operations does not imply that the end state will be Cold Shutdown (Mode 5).
4.2.1.3 Establishing Recovery Actions
Overview of Process
NEI 04-02 and RG 1.205 suggest that a licensee submit a summary of its approach for addressing the transition of Operator Manual Actions (OMAs) as recovery actions in the LAR (Regulatory Position 2.2.1 and NEI-04-02, Section 4.6). As a minimum, NEI 04-02
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suggests that the assumptions, criteria, methodology, and overall results be included for the NRC to determine the acceptability of the licensee’s methodology.
The discussion below provides the methodology used to transition pre-transition OMAs and to determine the population of post-transition recovery actions. This process is based on FAQ 07-0030 (ML110070485) and consists of the following steps:
• Step 1: Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s) (Activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition.
• Step 2: Determine the population of recovery actions that are required to resolve variances from deterministic requirements (VFDRs) (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).
• Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path.
• Step 4: Evaluate the feasibility of the recovery actions.
• Step 5: Evaluate the reliability of the recovery actions.
Results
The review results are documented in PINGP Engineering Evaluation EC 23313, “NFPA 805 LAR Supplement Attachment G.” Refer to Attachment G for the detailed evaluation process and summary of the results from the process.
4.2.1.4 Evaluation of Multiple Spurious Operations
Overview of Process
NEI 04-02 suggests that a licensee submit a summary of its approach for addressing potential fire-induced Multiple Spurious Operations (MSOs) for NRC review and approval. As a minimum, NEI 04-02 suggests that the summary contain sufficient information relevant to methods, tools, and acceptance criteria used to enable the NRC to determine the acceptability of the licensee’s methodology. The methodology utilized to address MSOs for PINGP is summarized below.
As part of the NFPA 805 transition project, a review and evaluation of PINGP susceptibility to fire-induced MSOs was performed. The process was conducted in accordance with NEI 04-02 and RG 1.205, as supplemented by FAQ 07-0038 Revision 3 (ML110140242). The PWR Generic MSO list dated May 2009 was utilized.
The approach outlined in Figure 4-3 (based on Figure XX from FAQ 07-0038) is one acceptable method to address fire-induced MSOs. This method used insights from the PRA developed in support of transition to NFPA 805 and consists of the following:
• Identifying potential MSOs of concern.
• Conducting an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1, Section F.4.2).
• Updating the Fire PRA model and NSCA to include the MSOs of concern.
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• Evaluating for NFPA 805 Compliance.
• Documenting Results.
This process is intended to support the transition to a new licensing basis. Post-transition changes would use the RI-PB change process. The post-transition change process for the assessment of a specific MSO would be a simplified version of this process, and may not need the level of detail shown in the following section (e.g., An expert panel may not be necessary to identify and assess a new potential MSO. Identification of new potential MSOs may be part of the plant change review process and/or inspection process).
Figure 4-3 – Multiple Spurious Operations – Transition Resolution Process (Based on FAQ 07-0038)
Results
Refer to Attachment F for a description of the process used at PINGP, which is based on the approach outlined in Figure 4-3, and the results from the process.
Identify Potential MSOs of Concern
SSA
Generic List of MSOs
Self Assessments
PRA Insights
Operating Experience
Update PRA model & NSCA (as
appropriate) to include MSOs of
concern
ID equipment
ID logical relationships
ID cables
ID cable routing
Expert Panel
Identify and Document MSOs of
Concern
Evaluate for NFPA 805
Compliance
Compliant with
NFPA 805?
Document Results
Step 1
Step 2
Step 3
Step 4 Pursue other resolution options
Yes
No
Step 5
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The EEEEs that support compliance with NFPA 805 Chapter 3 or Chapter 4 (both those that existed prior to the transition and those that were created during the transition) were reviewed using the methodology contained in NEI 04-02. The methodology for performing the EEEE review included the following determinations:
• The EEEE is not based solely on quantitative risk evaluations,
• The EEEE is an appropriate use of an engineering equivalency evaluation,
• The EEEE is of appropriate quality,
• The standard license condition is met,
• The EEEE is technically adequate,
• The EEEE reflects the plant as-built condition, and
• The basis for acceptability of the EEEE remains valid.
In accordance with the guidance in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs that demonstrate that a fire protection system or feature is “adequate for the hazard” are summarized in the LAR as follows:
• If not requesting specific approval for “adequate for the hazard” EEEEs, then the EEEE was referenced where required and a brief description of the evaluated condition was provided.
• If requesting specific NRC approval for “adequate for the hazard” EEEEs, then EEEE was referenced where required to demonstrate compliance and was included in Attachment L for NRC review and approval.
In all cases, the reliance on EEEEs to demonstrate compliance with NFPA 805 requirements was documented in the LAR.
Results
The review results for EEEEs are documented in PINGP Engineering Evaluation EC 20386, “NFPA 805 Existing Engineering Equivalency Evaluation Review Report.”
In accordance with the guidance provided in RG 1.205, Regulatory Position 2.3.2, and NEI 04-02, as clarified by FAQ 07-0054, Demonstrating Compliance with Chapter 4 of NFPA 805, EEEEs used to demonstrate compliance with Chapters 3 and 4 of NFPA 805 are referenced in Attachments A and C as appropriate.
The EEEE discussions in Attachment C include evaluations of conformance to applicable NFPA codes for the PINGP fire suppression and fire detection system installations. Where detection and suppression systems are required to satisfy NFPA 805 Chapter 4 requirements, any modifications to resolve code deviations identified within the evaluations are included in Attachment S, Table S-2. In cases where systems are not required to satisfy NFPA 805 Chapter 4 requirements, code deviations are resolved through the PINGP corrective action program.
In addition, none of the transitioning EEEEs require NRC approval.
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4.2.3 Licensing Action Transition
Overview of Evaluation Process
The existing licensing actions (exemptions) review was performed in accordance with NEI 04-02. The methodology for the licensing action review included the following:
• Determination of the bases for acceptability of the licensing action.
• Determination that these bases for acceptability are still valid and required for NFPA 805.
• In addition, variances from the deterministic requirements were identified in the NEI 04-02 Table B-3 (See Attachment C). Some of these variances were subsequently dispositioned via the use of the performance-based approach. A licensing action summary was completed for each fire area using the performance-based approach.
Results
Attachment K contains the detailed results of the Licensing Action Review. Where NRC clarification is needed for the continued acceptability of the exemption, the appropriate request for clarification is included in Attachment T.
The following licensing actions will be transitioned into the NFPA 805 fire protection program as previously approved (NFPA 805 Section 2.2.7). These licensing actions are considered compliant under 10 CFR 50.48(c).
• Appendix R Exemption, RCP Oil Collection, RCP oil collection system not in strict compliance (III.O criteria), Units 1 and 2, Fire Areas 1 and 71 (NRC SER dated July 31, 1984).
• Appendix R Exemption, Control Room, Repair action to remove fuses from PORV control circuits in event of control room evacuation, Units 1 and 2, Fire Area 13 (NRC SER dated February 21, 1995), subject to clarification requested in Attachment T.
The following licensing actions are no longer necessary and will not be transitioned into the NFPA 805 fire protection program:
• Appendix R Exemption, Control Room, Lack of automatic fixed suppression system (III.G.3 criteria), Units 1 and 2, Fire Area 13 (NRC SER dated February 2, 1983).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit a fixed suppression system in the control room.
• Appendix R Exemption, Train "A" Hot Shutdown Panel; Instrument Air Room and Auxiliary Feedwater Pump Room, Lack of 20' separation free of intervening combustibles or lack of a 1-hour fire barrier (III.G.2 criteria), Units 1 and 2, Fire
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Area 31 (NRC SER dated May 4, 1983).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit 20’ separation with no intervening combustibles.
• Appendix R Exemption, Train "B" Hot Shutdown Panel; Instrument Air Room and Auxiliary Feedwater Pump Room, Lack of 20' separation free of intervening combustibles or lack of a 1-hr fire barrier (III.G.2 criteria), Units 1 and 2, Fire Area 32 (NRC SER dated May 4, 1983).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit 20’ separation with no intervening combustibles.
• Appendix R Exemption, Normal Switchgear Room, Lack of automatic fixed suppression system (III.G.2 criteria), Unit 1, Fire Area 37 (NRC SER dated May 4, 1983).
This exemption was previously withdrawn and is no longer required because a facility modification relocated redundant cables outside the fire area.
• Appendix R Exemption, Auxiliary Building, Operating Level, Lack of automatic suppression system (III.G.2 criteria), Unit 1, Fire Area 60 (NRC SER dated May 4, 1983).
This exemption is no longer required because redundant equipment required for safe shutdown is no longer located in this fire area, due to a facility modification that changed power supplies for steam supply valves.
• Appendix R Exemption, Auxiliary Building, Operating Level, Lack of area wide suppression (III.G.2 criteria), Unit 2, Fire Area 75 (NRC SER dated May 4, 1983).
This exemption is no longer required because redundant equipment required for safe shutdown is no longer located in this fire area, due to a facility modification that changed power supplies for steam supply valves.
• Appendix R Exemption, Auxiliary Building, Ground Level, Lack of automatic fire suppression system (III.G.2 criteria), Unit 1, Fire Area 58 (NRC SER dated January 9, 1984).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit an automatic fire suppression system in this fire area.
• Appendix R Exemption, Auxiliary Building, Ground Level, Lack of automatic fire suppression system (III.G.2 criteria), Unit 2, Fire Area 73 (NRC SER dated January 9, 1984).
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This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit an automatic fire suppression system in this fire area.
• Appendix R Exemption, Auxiliary Building, Mezzanine Level, Lack of automatic fixed suppression (III.G.2 criteria), Unit 1, Fire Area 59 (NRC SER dated January 9, 1984).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit an automatic fixed suppression system in this fire area.
• Appendix R Exemption, Auxiliary Building, Mezzanine Level, Lack of automatic fixed suppression (III.G.2 criteria), Unit 2, Fire Area 74 (NRC SER dated January 9, 1984).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit an automatic fixed suppression system in this fire area.
• Appendix R Exemption, Containment, Intervening combustibles between redundant shutdown divisions, Units 1 and 2, Fire Areas 1 and 71 (NRC SER dated July 31, 1984).
This exemption is no longer required because the PINGP NFPA 805 transition compliance strategy is in accordance with NFPA 805 Section 4.2.4, and uses a performance based approach that does not credit 20’ separation with no intervening combustibles and takes into account the different fire protection features installed to protect pressurizer level transmitter cables in Unit 1 and Unit 2.
Since the exemptions are either compliant with 10 CFR 50.48(c) or no longer necessary, in accordance with the requirements of 10 CFR 50.48(c)(3)(i), PINGP requests that the exemptions listed in Attachment K be rescinded as part of the LAR process. It is NSPM’s understanding that implicit in the superseding of the current license condition, all prior fire protection program Safety Evaluation Reports and commitments will be superseded in their entirety. See Attachment O, Orders and Exemptions.
4.2.4 Fire Area Transition
Overview of Evaluation Process
The Fire Area Transition (NEI 04-02 Table B-3) was performed using the methodology contained NEI 04-02 and FAQ 07-0054. The methodology for performing the Fire Area Transition, depicted in Figure 4-4, is outlined as follows:
Step 1 - Assembled documentation. Gathered industry and plant-specific fire area analyses and licensing basis documents.
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Step 2 – Documented fulfillment of nuclear safety performance criteria.
• Assessed accomplishment of nuclear safety performance goals. Documented the method of accomplishment, in summary level form, for the fire area.
• Documented evaluation of effects of fire suppression activities. Documented the evaluation of the effects of fire suppression activities on the ability to achieve the nuclear safety performance criteria.
• Performed licensing action reviews. Performed a review of the licensing aspects of the selected fire area and documented the results of the review. See Section 4.2.3.
• Performed existing engineering equivalency evaluation reviews. Performed a review of existing engineering equivalency evaluations (or created new evaluations) documenting the basis for acceptability. See Section 4.2.2.
• Pre-transition OMA reviews. Performed a review of pre-transition OMAs to determine those actions taking place outside of the main control room or outside of the primary control station(s). See Section 4.2.1.3.
Step 3 – VFDR Identification and characterization and resolution considerations. Identified variances from the deterministic requirements of NFPA 805, Section 4.2.3. Documented variances as either a separation issue or a degraded fire protection system or feature. Developed VFDR problem statements to support resolution and selected an approach in accordance with NFPA 805 Chapter 4.
Step 4 – Performance-Based evaluations (Fire Modeling or Fire Risk Evaluations) See Section 4.5.2 for additional information. Alternatively, as shown in Figure 4-4, the VFDR condition was brought into compliance with Section 4.2.3 of NFPA 805.
Step 5 – Final Disposition.
• Documented final disposition of the VFDRs in Attachment C (NEI 04-02 Table B-3).
• For recovery action compliance strategies, ensured the manual action feasibility analysis of the required recovery actions was completed. Note: if a recovery action cannot meet the feasibility requirements established per NEI 04-02, then alternate means of compliance was considered.
• Documented the post transition NFPA 805 Chapter 4 compliance basis.
Step 6 – Documented required fire protection systems and features. Reviewed the NFPA 805 Section 4.2.3 compliance strategies (including fire area licensing actions and engineering evaluations) and the NFPA 805 Section 4.2.4 compliance strategies (including simplifying deterministic assumptions) to determine the scope of fire protection systems and features ‘required’ by NFPA 805 Chapter 4. The ‘required’ fire protection systems and features are subject to the applicable requirements of NFPA 805 Chapter 3.
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Figure 4-4 – Summary of Fire Area Review [Based on FAQ 07-0054 Revision 1]
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Results of the Evaluation Process
Attachment C contains the results of the Fire Area Transition review (NEI 04-02 Table B-3). On a fire area basis, Attachment C summarizes compliance with Chapter 4 of NFPA 805.
NEI 04-02 Table B-3 includes the following summary level information for each fire area:
• Performance Goal Summary – An overview of the method of accomplishment of each of the performance criteria in NFPA 805 Section 1.5 is provided.
• Reference Documents – Specific references to Nuclear Safety Capability Assessment Documents are provided.
• Licensing Actions – Specific references to exemptions that will remain part of the post-transition licensing basis are provided. A brief description of the condition and the basis for acceptability of the licensing action is provided. In addition summaries of Fire Risk Evaluations performed for variances from the deterministic requirements are also provided. Attachment T contains items for which NSPM is requesting concurrence of prior approval.
• EEEE – Specific references to EEEE that rely on determinations of “adequate for the hazard” that will remain part of the post-transition licensing basis are provided. A brief description of the condition and the basis for acceptability is provided.
• VFDRs – Specific variances from the deterministic requirements of NFPA 805 Section 4.2.3 are identified. Refer to Section 4.5.2 for a discussion of the performance-based approach.
4.3 Non-Power Operational Modes
4.3.1 Overview of Evaluation Process
NSPM implemented the process outlined in NEI 04-02 and FAQ 07-0040, Clarification on Non-Power Operations. The goal (as depicted in Figure 4-5) is to ensure that contingency plans are established when the plant is in a Non-Power Operational (NPO) mode where the risk is intrinsically high. During low risk periods, normal risk management controls and fire prevention/protection processes and procedures will be utilized.
The process to demonstrate that the nuclear safety performance criteria are met during NPO modes involved the following steps:
• Reviewed the existing Outage Management Processes.
• Identified Equipment/Cables:
o Reviewed plant systems to determine success paths that support each of the defense-in-depth Key Safety Functions (KSFs), and
o Identified cables required for the selected components and determined their routing.
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• Performed Fire Area Assessments (identify pinch points – plant locations where a single fire may damage all success paths of a KSF).
• Managed pinch-points associated with fire-induced vulnerabilities during the outage.
The process is depicted in Figures 4-5 and 4-6. The results are presented in Section 4.3.2.
Figure 4-5 Review POSs, KSFs, Equipment, and Cables, and Identify Pinch Points
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Figure 4-6 Manage Pinch Points
No
Yes
KSFEquipmentAvailability Changed?
KSFLost?
Determine
Fire Area Impact based on
NPO Fire Area Assessments
Implement
Contingency Plan for
Specific KSF
EquipmentOut of Service
(OOS)
No
Fire Protection Defense-in-
DepthActions
Higher Risk Evolution as Defined by Plant Specific Outage Risk Criteria for example1) Time to Boil
2) Reactor Coolant System and Fuel Pool Inventory3) Decay Heat Removal
Fire Protection Defense-in-
DepthActions
Higher Risk
Evolution?
Yes
Yes
No
Fire Protection Defense-in-
DepthActions
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4.3.2 Results of the Evaluation Process
Based on FAQ 07-0040 Revision 4, the Plant Operating States (POS) considered for equipment and cable selection are defined in PINGP Engineering Evaluation EC 23507, “NFPA 805 LAR Supplement Attachment D.” Components were identified to support the KSFs of Reactivity, Core Decay Heat Removal, Containment, Inventory, and associated support functions. A model was developed in the NFPA 805 Analysis Database (Genesis Solution Suite, SAFE Module). Equipment was logically tied to the supported KSF. Power supplies, interlocks, and supporting equipment were logically tied to their parent component. For those components which had not been previously analyzed in support of the at-power analysis or whose functional requirements may have been different for the NPO analysis, cable selection was performed in accordance with approved project procedures. Cables necessary to support the selected function of a component were selected and analyzed for fire impact. PINGP Engineering Evaluation, EC 23507, “NFPA 805 LAR Supplement Attachment D,” references the fire area assessment, the identified pinch points, and general recommendations for administrative controls to reduce fire risk as well as a proposed strategy for recovering the KSF should a fire occur. In accordance with FAQ 07-0040 Revision 4, any area experiencing fire damage which eliminates all success paths for a KSF (without recovery actions outside the main control room) is considered a pinch point. Fire modeling was not used to eliminate any fire area from being a pinch point. The list of generic recommendations specified in PINGP Engineering Evaluation EC 23507, “NFPA 805 LAR Supplement Attachment D,” considers the following actions from FAQ 07-0040 Revision 4:
• Prohibition or limitation of hot work in fire areas during periods of increased vulnerability.
• Verification of operable detection and/or suppression in the vulnerable areas.
• Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability.
• Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position).
• Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability.
• Use of recovery actions to mitigate potential losses of KSFs.
• Identification and monitoring in‐situ ignition sources for “fire precursors” (e.g., equipment temperatures).
• Reschedule the work to a period with lower risk or higher Defense-In-Depth (DID).
Refer to Attachment D for more complete details. Based on consideration of the vulnerable areas and incorporation of generic recommendations from FAQ 07-0040 Revision 4 into appropriate plant procedures and practices, prior to implementation of NFPA 805, the performance goals (KSFs) for NPO will be fulfilled and the requirements of NFPA 805 will be met. Implementation of the NPO fire area assessment results into the Prairie Island Nuclear Generating Plant outage management processes will be completed as part of LAR implementation. (See Attachment S).
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4.4 Radioactive Release Performance Criteria
4.4.1 Overview of Evaluation Process
The review of the fire protection program against NFPA 805 requirements for fire suppression related radioactive release was performed using the methodology contained in Nuclear Energy Institute (NEI) 04-02, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c) and NFPA 805 FAQ 09-0056. The methodology consisted of the following:
• A review of fire pre-plans and fire brigade training materials to identify fire protection program elements (e.g., systems / components / procedural control actions / flow paths, etc.) that are being credited to meet the radioactive release goals, objectives, and performance criteria during all plant operating modes, including full power and non-power conditions.
• A review of engineering controls to ensure containment of gaseous and liquid effluents (e.g., smoke and fire fighting agents). This review included all plant operating modes (including full power and non-power conditions). Otherwise, provided a bounding analysis, quantitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of radioactive effluents specified in the unit’s Technical Specifications or Offsite Dose Calculation Manual are met.
4.4.2 Results of the Evaluation Process
The PINGP fire strategies are developed based on the fire detection zone alarms received in the control room and can cover one or more fire areas. The fire strategies were reviewed to screen them for applicability by fire area based on their potential to contain radioactive or contaminated materials. PINGP Engineering Evaluation EC 23734, “NFPA 805 LAR Supplement Attachment E,” contains detailed evaluation bases and results regarding when a fire area is screened in (affects radioactive release) or screened out (cannot affect radioactive release). The radioactive release review determined the fire protection program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the implementation items identified in Attachment S.
The review determined that radioactive release to any unrestricted area due to the direct effects of fire suppression activities (but not involving fuel damage) would be as low as reasonably achievable. The site specific review of associated fire event and fire suppression related radioactive release is summarized in Attachment E of this document, which is based on the NEI 04-02 Table G-1.
As described in Attachment S (Table S-3, item 10), the Fire Strategies (PINGP-version of Pre-Fire Plans) will be revised to identify potential cross-contamination issues for each applicable fire area and fire detection zone. Information will be provided on cross contamination concerns to assist the fire brigade leader and Control Room personnel in determining the best available methods for minimizing cross contamination and radiological release based on the location of the fire.
As described in Attachment S (Table S-3, items 7, 8, 9, and 11), fire fighting instructions and brigade lesson plans will be revised to provide additional instructions on the control
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of the spread of contamination as a result of fire fighting activities. The responsibilities of each brigade member relative to limiting the spread of cross contamination when fighting fires in radiologically controlled areas will be incorporated into the instructions and lesson plans. Training materials for radiation protection will be revised as applicable to address their role in controlling the spread of contamination (Table S-3, item 12).
As described in Attachment S (Table S-3, item 13), ventilation and runoff will be addressed with respect to the impact on the spread of contamination to adjacent radiologically controlled areas, radiologically clean areas and release to the exterior. For those fire areas without installed ventilation controls (and for those time periods where existing ventilation controls are not available), mitigative actions will be taken to utilize a combination of exhausting potentially contaminated smoke through adjacent areas with filtered ventilation or the use of portable filtered ventilation equipment. The mitigative actions will be based on the radiological conditions as monitored by radiation protection personnel and communicated to the fire brigade leader during the event. Potentially contaminated water will be controlled by the use of booms to limit the spread of contamination to adjacent radiologically controlled areas, radiologically clean areas and to the exterior.
As described in Attachment S (Table S-3, items 15 and 16), a combination of containerization and administrative controls will be used limit the amount of exposed contaminated combustible materials in areas without filtered ventilation or where the spread of contaminated water to adjacent radiologically controlled areas, radiologically clean areas or to the exterior are potential concerns.
4.5 Fire PRA and Performance-Based Approaches
RI-PB evaluations are an integral element of an NFPA 805 fire protection program. Key parts of RI-PB evaluations include:
• A Fire PRA (discussed in Section 4.5.1 and Attachments U, V, and W).
• NFPA 805 Performance-Based Approaches (discussed in Section 4.5.2).
4.5.1 Fire PRA Development and Assessment
In accordance with the guidance in RG 1.205, a Fire PRA model was developed for PINGP in compliance with the requirements of Part 4 “Requirements for Fires At Power PRA,” of the ASME and ANS combined PRA Standard, ASME/ANS RA-Sa-2009, “Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application,” (hereafter referred to as Fire PRA Standard). NSPM conducted a peer review by independent industry analysts in accordance with RG 1.200 prior to a risk-informed submittal. The resulting fire risk assessment model is used as the analytical tool to perform Fire Risk Evaluations during the transition process.
Section 4.5.1.1 describes the Internal Events PRA model. Section 4.5.1.2 describes the Fire PRA model. Section 4.5.1.3 describes the results and resolution of the peer review of the Fire PRA, and Section 4.5.1.4 describes insights gained from the Fire PRA.
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4.5.1.1 Internal Events PRA
The PINGP base internal events PRA, Revision 3.1, was the starting point for the Fire PRA. Previously in 2006, the PINGP PRA underwent a gap assessment against the Capability Category II requirements of ASME RA-S-2002, with ASME RA-Sa-2003 and ASME RA-Sb-2005 Addenda, ASME, 2005. To update the PINGP PRA to Capability Category II of the Standard, a large-effort PRA upgrade was planned and initiated in 2007. At the conclusion of the PRA upgrade, NSPM completed the Revision 3.0 PRA model. An independent peer review team evaluated the PINGP Revision 3.0 PRA model against the most current combined PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2. With the exception of the Internal Flooding analysis which was performed separately as described later, the PINGP Revision 3.0 Level 1 analysis evaluated core damage frequency (CDF) from all internal events and large early release frequency (LERF) utilizing the Westinghouse Owner’s Group Simplified Level 2 Analysis Approach, WCAP-16341. Facts and Observations (F&Os) were issued by the peer team as an output of the peer review. NSPM subsequently evaluated these for inclusion into the next revision of the PRA, Revision 3.1, which was used as the starting point for the Fire PRA. Attachment U discusses the peer review findings to illustrate the technical adequacy of the Internal Events PRA supporting the Fire PRA. In addition, NSPM updated the Internal Flooding analysis. The Internal Flooding analysis was amended to the Revision 3.1 PRA model in a sensitivity analysis to determine the risk contribution of Internal Flooding to overall risk. The analysis was peer reviewed against the most current combined PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2, during a review performed in September 2012. Attachment U discusses the peer review findings to illustrate the technical adequacy of the Internal Flooding Analysis and provides a disposition of each finding as to its impact on the Fire PRA.
4.5.1.2 Fire PRA
The fire PRA was developed using the internal events PRA as a starting point. The internal events PRA was modified to capture the effects of fire both as an initiator of an event and as a potential failure mode of affected circuits and individual targets. The Fire PRA is a unit-specific analysis that takes into account inter-unit dependencies. The Fire PRA was quantified using the CAFTA and PRAQuant software. A Fire PRA model was developed for NSPM using the guidance provided in NUREG/CR-6850/EPRI TR-1011989, Supplement 1 to NUREG/CR-6850/EPRI 1019259, and draft NUREG-1921.
A Peer Review of the PINGP Fire PRA against the requirements of Section 4 of the combined PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2, was conducted the week of May 7 through May 11, 2012. There were not any previously un-reviewed methods used to complete the PINGP Fire PRA. The Fire PRA quality and insights are discussed in Attachments V and W, respectively. After the initial submittal
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of the PINGP NFPA 805 LAR, NSPM revised the Fire PRA. The results of this Fire PRA revision are discussed in Attachment W, Revision 1.
Fire Model Utilization in the Application
Fire modeling was performed as part of the Fire PRA development (NFPA 805 Section 4.2.4.2). RG 1.205, Regulatory Position 4.2 and Section 5.1.2 of NEI 04-02, provide guidance to identify fire models that are acceptable to the NRC for plants implementing a risk-informed, performance-based licensing basis.
The following fire models were used:
• Flame Height (Method of Heskestad)
• Plume Centerline Temperature (Method of Heskestad)
• Radiant Heat Flux (Point Source Method)
• Hot Gas Layer (Method of MQH)
• Hot Gas Layer (Method of Beyler)
• Hot Gas Layer (Method of Foote, Pagni, and Alvares [FPA])
• Ceiling Jet Temperature (Method of Alpert)
• Hot Gas Layer Calculations using Fire Dynamics Simulator (Version 5)
• Hot Gas Layer Calculations using CFAST (Version 6)
• Corner and Wall HRR
• Correlation for Heat Release Rates of Cables (Method of Lee)
• FLASH-CAT
The acceptability of the use of these fire models is included in Attachment J.
4.5.1.3 Results of Fire PRA Peer Review
The PINGP Fire PRA, Revision 0, was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, Part 4, and in accordance with the peer review guidelines of NEI 07-12. This peer review was conducted the week of May 7 through May 11, 2012.
Section 4 of the ASME/ANS PRA Standard contains a total of 183 Supporting Requirements (SRs) under 13 technical elements, and configuration control from Section 1.5. Of these 183 SRs, eighteen (18) were determined to be not applicable to the PINGP Fire PRA either due to the fact that the requirements were not applicable to the PINGP approach, or the technical element was not used for the PINGP analysis (i.e., quantitative screening, QNS). For the PINGP Fire PRA about 92% of the SRs were assessed at Capability Category II or higher, including 5% of the SRs being assessed at Capability Category III. The PINGP Fire PRA had an additional 3% of the applicable SRs assessed at the CC-I level. The PINGP Fire PRA does not meet 5% of the applicable SRs. There were no SRs “Not Reviewed” by the Peer Review Team. There were also no “Unreviewed Analysis Methods” identified by the Team. The Peer Review also noted a total of 56 Facts and Observations (F&Os). These included fifteen (15) “Suggestions,” forty (40) “Findings” and one (1) “Best Practice.”
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The Finding F&Os and their disposition with respect to the NFPA 805 License Amendment Request are provided in Table V-1, organized by Technical Element and Supporting Requirement. The Finding F&Os covered a variety of topics, but many dealt with the need to incorporate additional detailed analyses to develop results that are more realistic rather than bounding. Several others were on the need to better identify assumptions and discuss their impact on overall results. Suggestion F&Os largely involve optional clarifications or improvements. The Best Practice F&O was issued for the Seismic Fire Interaction Technical Element.
Subsequent to the May 2012 peer review, updates were made to the Fire PRA. An additional focused peer review was completed in December 2013 to review fire modeling methods that were used to support Hot Gas Layer (HGL) calculations. The analysis was peer reviewed against the most current combined PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2.
The Focused Peer Review noted a total of five (5) F&Os. These included four (4) “Suggestions,” and one (1) “Finding.” The Finding F&O and its disposition with respect to the NFPA 805 License Amendment Request are provided in Table V-1. One of the four Suggestions provided additional information and was appended to an F&O that had been previously identified during the initial Fire PRA Peer Review.
Attachment V contains a summary of the FPRA peer review F&Os and their disposition by NSPM.
4.5.1.4 Risk Insights
Risk insights were documented as part of the development of the Fire PRA. The total plant fire CDF/LERF was derived using the NUREG/CR-6850 methodology for Fire PRA development and is useful in identifying the areas of the plant where fire risk is greatest. A review of the fire scenarios that individually contribute more than 0.5% of calculated fire risk are included as Attachment W.
4.5.2 Performance-Based Approaches
NFPA 805 outlines the approaches for performing performance-based analyses. As specified in Section 4.2.4, there are generally two types of analyses performed for the performance-based approach:
• Fire Modeling (NFPA 805 Section 4.2.4.1).
• Fire Risk Evaluation (FRE, NFPA 805 Section 4.2.4.2).
The PINGP NFPA 805 transition implemented the FRE approach per NFPA 805 Section 4.2.4.2 to evaluate the risk significance and acceptability of the VFDRs.
4.5.2.1 Fire Modeling Approach
The fire modeling approach per NFPA 805 Section 4.2.4.1 was not utilized for the PINGP NFPA 805 transition.
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4.5.2.2 Fire Risk Approach
Overview of Evaluation Process
The Fire Risk Evaluations were completed as part of the PINGP NFPA 805 transition. These Fire Risk Evaluations were developed using the process described below. This methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These are summarized in Table 4-1.
Table 4-1 Fire Risk Evaluation Guidance Summary Table
Document Section(s) Topic
NFPA 805 2.2(h), 4.2.4, A.2.2(h), A.2.4.4, D.5 Change Evaluation (2.2(h), 2.2.9, 2.4.4 A.2.2(h), A.2.4.4, D.5) Risk of Recovery Actions (4.2.4) Use of Fire Risk Evaluation (4.2.4.2)
NEI 04-02 Revision 2 4.4, 5.3, Appendix B, Appendix I, Appendix J
Change Evaluation, Change Evaluation Forms (App. I), No specific discussion of Fire Risk Evaluation
During the transition to NFPA 805, variances from the deterministic approach in Section 4.2.3 of NFPA 805 were evaluated using a Fire Risk Evaluation per Section 4.2.4.2 of NFPA 805. A Fire Risk Evaluation was performed for each fire area containing variances from the deterministic requirements of Section 4.2.3 of NFPA 805 (VFDRs).
If the Fire Risk Evaluation meets the acceptance criteria, this is confirmation that a success path effectively remains free of fire damage and that the performance-based approach is acceptable per Section 4.2.4.2 of NFPA 805.
The Fire Risk Evaluation process consists of the following steps (Figure 4-7 depicts the Fire Risk Evaluation process used during transition. This is generally based on FAQ 07-0054 Revision 1 (ML110140183):
Step 1 – Preparation for the Fire Risk Evaluation.
• Definition of the Variances from the Deterministic Requirements. The definition of the VFDR includes a description of problem statement and the section of NFPA 805 that is not met, type of VFDR (e.g., separation issue or degraded fire protection system), and proposed evaluation per applicable NFPA 805 section.
• Preparatory Evaluation – Fire Risk Evaluation Team Review. Using the information obtained during the development of NEI 04-02, Table B-3 and the Fire PRA, a team review of the VFDR was performed. The FRE review team included a Safe Shutdown/NSCA Engineer, a Fire Protection Engineer, and a Fire PRA Engineer. The purpose and objective of this team review was to address the following:
o Review of the Fire PRA modeling treatment of VFDR
o Ensure discrepancies were captured and resolved
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Step 2 – Performed the Fire Risk Evaluation
• The Evaluator coordinated as necessary with the Safe Shutdown/NSCA Engineer, Fire Protection Engineer and Fire PRA Engineer to assess the VFDR using the Fire Risk Evaluation process to perform the following:
o Change in Risk Calculation with consideration for additional risk of recovery actions and required fire protection systems and features due to fire risk.
o Fire area change in risk summary.
Step 3 – Reviewed the Acceptance Criteria
• The acceptance criteria for the Fire Risk Evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are ∆CDF and ∆LERF. The qualitative factors are defense-in-depth and safety margin.
o Risk Acceptance Criteria. The transition risk evaluation was measured quantitatively for acceptability using the ∆CDF and ∆LERF criteria from RG 1.174, as clarified in RG 1.205 Regulatory Position 2.2.4.
o Defense-in-Depth. A review of the impact of the change on defense-in-depth was performed, using the guidance from NEI 04-02. NFPA 805 defines defense-in-depth as:
- Preventing fires from starting
- Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage
- Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.
In general, the defense-in-depth requirement was considered to be satisfied if the proposed change does not result in a substantial imbalance among these elements (or echelons).
The review of defense-in-depth was qualitative and addressed each of the elements with respect to the proposed change. Defense-in-depth was performed on a fire area basis.
Fire protection features and systems relied upon to ensure defense-in-depth were identified as a result of the assessment of defense-in-depth.
o Safety Margin Assessment. A review of the impact of the change on safety margin was performed. An acceptable set of guidelines for making that assessment is summarized below. Other equivalent acceptance guidelines may also be used.
- Codes and standards or their alternatives accepted for use by the NRC are met, and
- Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.
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The requirements related to safety margins for the change analysis are described for each of the specific analysis types used in support of the FRE.
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Figure 4-7 – Fire Risk Evaluation Process (NFPA 805 Transition) [Based on FAQ 07-0054 Revision 1]
Identification of VFDRs
(From B-3 Tables)
Determine How to Model
the VFDR in the Fire PRA
Discuss and Document in
Fire PRA and Fire Risk
Evaluation Documentation
Prepare for Fire Risk
Evaluation
Perform Fire Risk
Evaluation
Evaluate the Maintenance
of
Defense-In-Depth
And
Safety Margin
Discuss and Document in
Fire Risk Evaluation
Calculation
Review of Acceptance
Criteria
Evaluate
Delta CDF
And
Delta LERF
Calculate VFDR
Delta CDF
And
Delta LERF
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Results of Evaluation Process
Disposition of VFDRs
The PINGP existing post-fire SSA / NSCA and the NFPA 805 transition project activities have identified a number of variances from the deterministic requirements of NFPA 805 Section 4.2.3. These variances were dispositioned using the fire risk evaluation process.
Each variance dispositioned using a Fire Risk Evaluation was assessed against the Fire Risk Evaluation acceptance criteria of ∆CDF and ∆LERF; and maintenance of defense-in-depth and safety margin criteria from Section 5.3.5 of NEI 04-02 and RG 1.205. The results of these calculations are summarized in Attachment C.
Following completion of transition activities and planned modifications and program changes, the plant will be compliant with 10 CFR 50.48(c).
Risk Change Due to NFPA 805 Transition
In accordance with the guidance in RG 1.205, Section C.2.2.4, Risk Evaluations, risk increases or decreases for each fire area using Fire Risk Evaluations and the overall plant should be provided. Note that the risk increase due to the use of recovery actions was included in the risk change for transition for each fire area.
RG 1.205, Section C.2.2.4.2 states in part
“The total increase or decrease in risk associated with the implementation of NFPA 805 for the overall plant should be calculated by summing the risk increases and decreases for each fire area (including any risk increases resulting from previously approved recovery actions). The total risk increase should be consistent with the acceptance guidelines in Regulatory Guide 1.174. Note that the acceptance guidelines of Regulatory Guide 1.174 may require the total CDF, LERF, or both, to evaluate changes where the risk impact exceeds specific guidelines. If the additional risk associated with previously approved recovery actions is greater than the acceptance guidelines in Regulatory Guide 1.174, then the net change in total plant risk incurred by any proposed alternatives to the deterministic criteria in NFPA 805, Chapter 4 (other than the previously approved recovery actions), should be risk neutral or represent a risk decrease.”
The risk increases and decreases are provided in Attachment W.
4.6 Monitoring Program
4.6.1 Overview of NFPA 805 Requirements for the NFPA 805 Monitoring Program
Section 2.6 of NFPA 805 states:
“A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid.”
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The intent of the monitoring review is to confirm the adequacy of the existing surveillance, inspection, testing, compensatory measures, and oversight processes for transition to NFPA 805. This review considers the following:
• The adequacy of the scope of structures, systems and components within existing plant programs.
• The performance criteria for the availability and reliability of the required structures, systems and components.
• The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence.
4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program
This section provides an overview of the post-transition NFPA 805 Monitoring Program process. The Monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring process is comprised of four phases:
• Phase 1 – Scoping
• Phase 2 – Screening Using Risk Criteria
• Phase 3 – Risk Target Value Determination
• Phase 4 – Monitoring Implementation
Figure 4-8 provides detail on the Phase 1 and 2 processes.
Phase 1 – Scoping
In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program:
• Structures, Systems, and Components required to comply with NFPA 805, specifically:
o Fire protection systems and features - Required by the Nuclear Safety Capability Assessment - Modeled in the Fire PRA - Required by Chapter 3 of NFPA 805
o Nuclear Safety Capability Assessment equipment4 - Nuclear safety equipment - Fire PRA equipment - NPO equipment
o SSCs relied upon to meet radioactive release criteria
• Fire Protection Programmatic Elements
• Radioactive Release Engineered Systems and Features
4 For the purpose of the NFPA 805 Monitoring, “NSCA equipment” is intended to include Nuclear Safety
Equipment, Fire PRA equipment, and NPO equipment.
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Phase 2 – Screening Using Risk Criteria
The equipment from Phase 1 scoping will be screened to determine the appropriate level of NFPA 805 monitoring. As a minimum, the SSCs identified in Phase 1 should be part of an inspection and test program and a system/program health program. If not in the current program, the SSCs will be added in order to assure that the criteria can be met reliably.
The following screening process will be used to determine those SSCs that may require additional monitoring beyond normal surveillance activities.
1. Fire Protection Systems and Features
Those fire protection systems and features identified in Phase 1 would be candidates for additional monitoring in the NFPA 805 program commensurate with risk significance.
Risk significance may be accomplished at the component, programmatic element, and/or functional level. Since risk is evaluated at the compartment level or fire area level, criteria must be developed to determine those analysis units for which the fire protection SSCs contained within the area are considered risk significant. Screening compartments and fire areas will also include considerations for design/operation/ maintenance limitations. For instance, fire detection should not subdivide systems beyond the system/train/channel level used in normal operation/maintenance.
The Fire PRA is the primary tool used to establish the risk significance criteria and performance bounding guidelines. The screening thresholds used to determine risk significant analysis units are those that meet the following criteria:
Risk Achievement Worth (RAW) of the monitored parameter ≥ 2.0
(AND) either
Core Damage Frequency (CDF) x (RAW) ≥ 1.0E-7 per year
(OR)
Large Early Release Frequency (LERF) x (RAW) ≥ 1.0E-8 per year
CDF, LERF, and RAW(monitored parameter) are calculated for each fire area. The ‘monitored parameter’ will be established by licensee at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration). If compartments are used that are smaller than fire areas, sufficient basis will be documented.
The monitoring program will include the appropriate fire protection program SSCs based on the criteria above. Additional fire protection program SSCs may also be screened in based on plant-specific considerations.
2. Nuclear Safety Capability Assessment Equipment
NSCA equipment may already be appropriately monitored by the Maintenance Rule. A comparison of NSCA equipment to the SSCs that are monitored in the Maintenance Rule program will be performed to determine what equipment may require additional NFPA 805 Monitoring. For NSCAs SSCs not monitored by the Maintenance Rule, the
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basis for inclusion or exclusion of the SSCs in the NFPA 805 monitoring program will be documented, and the process used to make this determination will be fully described.
The Fire PRA will be used to identify high-safety-significant (HSS) NSCA SSCs that require monitoring. The Maintenance Rule guidelines differentiating HSS from low-safety-significant (LSS) SSCs will be used. High-safety-significant NSCA SSCs not currently monitored in Maintenance Rule will be included in the PINGP Maintenance Rule program. Revisions to PINGP Procedure FP-E-MR-01, “Maintenance Rule Process,” will be completed as an Implementation Item described in Attachment S, Table S-3. All NSCA SSCs that are not HSS should be considered LSS and need not be included in the monitoring program.
For fires originating during non-power operational modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement. Therefore, fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire prevention programs. Additional monitoring beyond inspection and test programs and system/program health programs is not considered necessary.
3. SSCs Relied upon for Radioactive Release Criteria
The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSCs relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently low risk. Therefore, additional monitoring beyond inspection and test programs and system/program health programs is not considered necessary.
4. Monitoring of Fire Protection Programmatic Elements
Monitoring of programmatic elements is required in order to “assess the performance of the fire protection program in meeting the performance criteria.” Programmatic aspects include:
• Transient Combustible Control; Transient Exclusion Zones
• Hot Work Control; Administrative Controls
• Fire Watch Programs; Program compliance and effectiveness
• Fire Brigade Effectiveness
Fire protection health reports, self-assessments, regulator and insurance company reports provide inputs to the monitoring program. The monitoring of programmatic elements and program effectiveness will be performed as part of the management of engineering programs. This monitoring is more qualitative in nature since the programs do not lend themselves to the numerical methods of reliability and availability. These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements.
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Phase 3 – Risk Target Value Determination
Phase 3 consists of using the Fire PRA, or other processes as appropriate, to determine target values of reliability and availability for the HSS fire protection/NSCA SSCs and programmatic elements established in Phase 2 as requiring additional monitoring beyond inspection and test programs, and system/program health programs.
Failure criteria are established by an expert panel or evaluation based on the required fire protection and nuclear safety capability SSCs and programmatic elements assumed level of performance in the supporting analyses. Action levels are established for the SSCs at the component level, program level, or functionally through the use of the pseudo system or ‘performance monitoring group’ concept. An action level will be developed for the NSCA SSCs that are included in a monitoring program.
Since the HSS SSCs have been identified using the Maintenance Rule guidelines, the associated SSC specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (~2-3 operating cycles). Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels. The Monitoring Program failure criteria and action level targets will be documented.
Phase 4 – Monitoring Implementation
Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established. Monitoring consists of periodically gathering, trending, and evaluating information pertinent to the performance, and/or availability of the SSCs and comparing the results with the established goals and performance criteria to verify that the goals are being met. Results of monitoring activities will be analyzed in a timely manner to assure that appropriate action is taken. The corrective action process will be used to address performance of fire protection and nuclear safety SSCs that do not meet performance criteria.
For fire protection and NSCA SSCs that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, corrective action will be initiated to identify the negative trend. A corrective action plan will then be developed using the appropriate licensee process. Once the plan has been implemented, improved performance should return the SSC back to below the established action level.
A periodic assessment will be performed (e.g., at a frequency of approximately every two to three operating cycles), taking into account, where practical, industry wide operating experience. This will be conducted as part of other established assessment activities. Issues that will be addressed include:
• Review systems with performance criteria. Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the fire protection and nuclear safety capability assessment systems?
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• Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection and nuclear safety capability assessment SSCs, programmatic elements and/ or functions need to be in scope?
• Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?
Function currently in
Maintenance Rule?
Component currently
in FPRA?
Fire Protection Systems
and FeaturesNSEL Components
Rad Release
Engineered Systems
and Features
No
High Safety
Significance of feature
by compartment?
NFPA 805 Specific
Monitoring Process
Establish targets for
reliability/unavailability in
Phase 3
Use Maintenance Rule
for Monitoring
Yes
Yes
Normal System &
Program Health
Monitoring Process or
Outage Risk
Management for NPO
Include in
Maintenance Rule?
High Risk
Significance?
Yes
No
Yes
No
Fire Protection
Programmatic
Elements
Yes
No
NPO Components FPRA Components
NSCA
No
Phase 1 - Scoping
Phase 2 - Screening
*Fully describe process used*
Figure 4-8 – NFPA 805 Monitoring – Scoping and Screening
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4.7 Program Documentation, Configuration Control, and Quality Assurance
4.7.1 Compliance with Documentation Requirements in Section 2.7.1 of NFPA 805
In accordance with the requirements and guidance in NFPA 805 Section 2.7.1 and NEI 04-02, PINGP has documented analyses to support compliance with 10 CFR 50.48(c). The analyses are performed in accordance with NSPM’s processes for ensuring assumptions are clearly defined, that results are easily understood, that results are clearly and consistently described, and that sufficient detail is provided to allow future review of the entire analyses.
Analyses, as defined by NFPA 805 Section 2.4, performed to demonstrate compliance with 10 CFR 50.48(c) will be maintained for the life of the plant and organized to facilitate review for accuracy and adequacy. Note these analyses do not include items such as periodic tests, hot work permits, fire impairments, etc.
The Fire Protection Design Basis Document described in Section 2.7.1.2 of NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 will be created as part of transition to 10 CFR 50.48(c) to ensure program implementation following receipt of the safety evaluation. Appropriate cross references will be established to supporting documents as required by NSPM processes. Figure 4-9 depicts the planned post-transition documentation and relationships.
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Figure 4-9 – NFPA 805 Planned Post-Transition Documents and Relationships
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4.7.2 Compliance with Configuration Control Requirements in Section 2.7.2 and 2.2.9 of NFPA 805
Program documentation established, revised, or utilized in support of compliance with 10 CFR 50.48(c) is subject to NSPM configuration control processes that meet the requirements of Section 2.7.2 of NFPA 805. This includes the appropriate procedures and configuration control processes for ensuring that changes impacting the fire protection program are reviewed appropriately. The RI-PB post transition change process methodology is based upon the requirements of NFPA 805, industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-2.
The Plant Change Evaluation Process consists of the following 4 steps and is depicted in Figure 4-10:
• Defining the Change
• Performing the Preliminary Risk Screening
• Performing the Risk Evaluation
• Evaluating the Acceptance Criteria
Change Definition
The Change Evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).
1. The baseline is defined as that plant condition or configuration that is consistent with the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).
2. The changed or altered condition or configuration that is not consistent with the Design Basis and Licensing Basis is defined as the proposed alternative.
Preliminary Risk Review
Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-
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03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).
The characteristics of an acceptable screening process that meets the “assessment of the acceptability of risk” requirement of Section 2.4.4 of NFPA 805 are:
• The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk.
• The screening process must be documented and be available for inspection by the NRC.
• The screening process does not pose undue evaluation or maintenance burden.
If any of the above is not met, proceed to the Risk Evaluation step.
Risk Evaluation
The screening is followed by engineering evaluations that may include fire modeling and risk assessment techniques. The results of these evaluations are then compared to the acceptance criteria. Changes that satisfy the acceptance criteria of NFPA 805, Section 2.4.4, and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition. The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature.
The risk evaluation involves the application of fire modeling analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below.
Acceptability Determination
The Change Evaluations are assessed for acceptability using the ∆CDF (change in core damage frequency) and ∆LERF (change in large early release frequency) criteria from the license condition. The proposed changes are also assessed to ensure they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.
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Figure 4-10 Plant Change Evaluation [NEI 04-02 Figure 5-1] Note references in Figure refer to NEI 04-02 Sections
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The PINGP Fire Protection Program configuration is defined by the program documentation. To the greatest extent possible, the existing configuration control processes for modifications, calculations and analyses, and Fire Protection Program License Basis Reviews will be utilized to maintain configuration control of the Fire Protection program documents. The configuration control procedures which govern the various PINGP documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.
Several NFPA 805 document types such as: NSCA Supporting Information, Non-Power Mode NSCA Treatment, etc., generally require new control procedures and processes to be developed since they are new documents and databases created as a result of the transition to NFPA 805. The new procedures will be modeled after the existing processes for similar types of documents and databases. System level design basis documents will be revised to reflect the NFPA 805 role that the system components now play.
The process for capturing the impact of proposed changes to the plant on the Fire Protection Program will continue to be a multiple step review. The first step of the review is an initial screening for process users to determine if there is a potential to impact the Fire Protection program as defined under NFPA 805 through a series of screening questions/checklists contained in one or more procedures depending upon the configuration control process being used. Reviews that identify potential Fire Protection program impacts will be sent to qualified individuals (Fire Protection, Safe Shutdown/NSCA, Fire PRA) to ascertain the program impacts, if any. If Fire Protection program impacts are determined to exist as a result of the proposed change, the issue would be resolved by one of the following:
• Deterministic Approach: Comply with NFPA 805 Chapter 3 and 4.2.3 requirements.
• Performance-Based Approach: Utilize the NFPA 805 change process developed in accordance with NEI 04-02, RG 1.205, and the PINGP NFPA 805 fire protection license condition to assess the acceptability of the proposed change. This process would be used to determine if the proposed change could be implemented "as-is" or whether prior NRC approval of the proposed change is required.
This process follows the requirements in NFPA 805 and the guidance outlined in RG 1.174 which requires the use of qualified individuals, procedures that require calculations be subject to independent review and verification, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Plant procedures will be developed (or existing procedures revised, as appropriate) to govern the configuration control processes required by NFPA 805. See Attachment S, Table S-3, for associated implementation actions.
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4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805
Fire Protection Program Quality
NSPM will continue to maintain the existing Fire Protection Quality Assurance program.
During the transition to 10 CFR 50.48(c), NSPM performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.
Fire PRA Quality
Configuration control of the Fire PRA model will be maintained by integrating the Fire PRA model into the existing processes used to ensure configuration control of the internal events PRA model. This process complies with Section 5 of the ASME Standard for PRA Quality and ensures that NSPM maintains an as-built, as-operated PRA model of the plant. The process has been peer reviewed. Quality assurance of the Fire PRA is assured via the same processes applied to the internal events model.
This process follows the guidance outlined in RG 1.174 which requires the use of qualified individuals, procedures that require calculations be subject to independent review, verification, or checking, record retention, peer review, and a corrective action program that ensures appropriate actions are taken when errors are discovered. Although the entire scope of the formal 10 CFR 50 Appendix B program is not applied to the PRA models or processes in general, often parts of the program are applied as a convenient method of complying with the requirements of RG 1.174. For instance, the procedures which address software control and the corrective action program for the PINGP 10 CFR 50 Appendix B program are applied to the PRA program.
With respect to Quality Assurance Program requirements for independent reviews of calculations and evaluations, those existing requirements for Fire Protection Program documents will remain unchanged. NSPM specifically requires that the calculations and evaluations in support of the NFPA 805 LAR, exclusive of the Fire PRA, be performed within the scope of the QA program which requires independent review as defined by NSPM procedures. As recommended by NUREG/CR-6850, the sources of uncertainty in the Fire PRA were identified and specific parameters were analyzed in support of the NFPA 805 Fire Risk Evaluation process.
Specifically with regard to uncertainty, uncertainty associated with Fire PRA parameters was qualitatively addressed in Fire PRA Uncertainty Analysis Notebook.
While the removal of conservatism inherent in the Fire PRA is a long-term goal, the Fire PRA results were deemed sufficient for evaluating the risk associated with this application. While NSPM continues to strive toward a more "realistic" estimate of fire risk, use of mean values continues to be the best estimate of fire risk. During the Fire Risk Evaluation process, the uncertainty and sensitivity associated with specific Fire PRA parameters were considerations in the evaluation of the change in risk relative to the applicable acceptance thresholds.
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Specific Requirements of NFPA 805 Section 2.7.3
NFPA 805 Section 2.7.3.1 – Review
Analyses, calculations, and evaluations performed in support of compliance with 10 CFR 50.48(c) are performed in accordance with NSPM procedures that require independent review.
NFPA 805 Section 2.7.3.2 – Verification and Validation
Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805.
NFPA 805 Section 2.7.3.3 – Limitations of Use
Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) are used and were used appropriately as required by Section 2.7.3.3 of NFPA 805.
NFPA 805 Section 2.7.3.4 – Qualification of Users
Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.
For personnel performing fire modeling or Fire PRA development and evaluation, NSPM will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work. These implementation items are identified in Attachment S-3.
NFPA 805 Section 2.7.3.5 – Uncertainty Analysis
Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and Fire PRA development.
4.8 Summary of Results
4.8.1 Results of the Fire Area Review
A summary of the NFPA 805 compliance basis and the required fire protection systems and features is provided in Attachment C, NEI 04-02, Table B-3, “Fire Area Review.” The table provides the following information from NEI 04-02, Table B-3:
• Fire Area / Fire Zone: Fire Area/Zone Identifier.
• Description: Fire Area/Zone Description.
• NFPA 805 Regulatory Basis: Post-transition NFPA 805 Chapter 4 compliance basis (Note: Compliance is determined on a Fire Area basis therefore a compliance basis is not provided for individual fire zones.)
• Required Fire Protection System / Feature: Detection / suppression required in the Fire Area based on NFPA 805 Chapter 4 compliance. Other Required
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Features may include Electrical Raceway Fire Barrier Systems, fire barriers, etc. The documentation of required fire protection systems and features does not include the documentation of the fire area boundaries. Fire area boundaries are required and documentation of the fire area boundaries has been performed as part of reviews of engineering evaluations, licensing actions, or as part of the reviews of the NEI 04-02 Table B-1 process. The basis for the requirement of the fire protection system / feature is designated as follows:
o S – Separation Criteria: Systems/Features required for Chapter 4 Separation Criteria in Section 4.2.3.
o E – EEEE Systems/Features required for acceptability of Existing Engineering Equivalency Evaluations (Section 2.2.7).
o L – Licensing Action Criteria – Systems/Features required for acceptability of NRC approved Licensing Action (i.e., Exemptions) (Section 2.2.7).
o R – Risk Criteria: Systems/Features required to meet the Risk Criteria for the Performance-Based Approach (Section 4.2.4).
o D – Defense-in-depth Criteria: Systems/Features required to maintain adequate balance of Defense-in-Depth for a Performance-Based Approach (Section 4.2.4).
Attachment W contains the results of the Fire Risk Evaluations, additional risk of recovery actions, and the change in risk on a fire area basis.
4.8.2 Plant Modifications and Items to be Completed During the Implementation Phase
Planned modifications, studies, and evaluations to comply with NFPA 805 are described in Attachment S. These actions were identified in PINGP Engineering Evaluation EC 23430, “NFPA 805 LAR Supplement Attachment S, Planned Modifications.” Table S-1 identifies plant modifications associated with the transition to NFPA 805 that have been completed. Table S-2 summarizes plant modifications that are committed for implementation. Table S-3 provides a list of those items (e.g., procedure changes, process updates, and training of affected plant personnel) that will be completed prior to the implementation of the new NFPA 805 Fire Protection Program at PINGP.
The Fire PRA model represents the as-built, as-operated and maintained plant as it will be configured at the completion of the transition to NFPA 805. The Fire PRA model includes credit for the planned implementation of the modifications listed in Attachment S. Following completion of the implementation items listed in Attachment S, such as further development of procedure changes and training, additional refinements may need to be incorporated into the Fire PRA based on industry initiatives. No other significant plant changes are outstanding with respect to their inclusion in the Fire PRA model. See Implementation Item in Attachment S, Table S-3.
4.8.3 Supplemental Information –Other Licensee Specific Issues
None.
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5.0 REGULATORY EVALUATION
5.1 Introduction – 10 CFR 50.48
On July 16, 2004 the NRC amended 10 CFR 50.48, Fire Protection, to add a new subsection, 10 CFR 50.48(c), which establishes alternative fire protection requirements. 10 CFR 50.48 endorses, with exceptions, NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants – 2001 Edition (NFPA 805), as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b), Appendix R, and Section (f), Decommissioning.
The voluntary adoption of 10 CFR 50.48(c) by PINGP does not eliminate the need to comply with 10 CFR 50.48(a) and 10 CFR 50, Appendix A, GDC 3, Fire Protection. The NRC addressed the overall adequacy of the regulations during the promulgation of 10 CFR 50.48(c) (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086).
“NFPA 805 does not supersede the requirements of GDC 3, 10 CFR 50.48(a), or 10 CFR 50.48(f). Those regulatory requirements continue to apply to licensees that adopt NFPA 805. However, under NFPA 805, the means by which GDC 3 or 10 CFR 50.48(a) requirements may be met is different than under 10 CFR 50.48(b). Specifically, whereas GDC 3 refers to SSCs important to safety, NFPA 805 identifies fire protection systems and features required to meet the Chapter 1 performance criteria through the methodology in Chapter 4 of NFPA 805. Also, under NFPA 805, the 10 CFR 50.48(a)(2)(iii) requirement to limit fire damage to SSCs important to safety so that the capability to safely shut down the plant is ensured is satisfied by meeting the performance criteria in Section 1.5.1 of NFPA 805. The Section 1.5.1 criteria include provisions for ensuring that reactivity control, inventory and pressure control, decay heat removal, vital auxiliaries, and process monitoring are achieved and maintained.
This methodology specifies a process to identify the fire protection systems and features required to achieve the nuclear safety performance criteria in Section 1.5 of NFPA 805. Once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design must meet any applicable requirements of NFPA 805, Chapter 3. Having identified the required fire protection systems and features, the licensee selects either a deterministic or performance-based approach to demonstrate that the performance criteria are satisfied. This process satisfies the GDC 3 requirement to design and locate SSCs important to safety to minimize the probability and effects of fires and explosions.” (Reference FR Notice 69 FR 33536 dated June 16, 2004, ML041340086)
The new rule provides actions that may be taken to establish compliance with 10 CFR 50.48(a), which requires each operating nuclear power plant to have a fire protection program plan that satisfies GDC 3, as well as specific requirements in that section. The transition process described in 10 CFR 50.48(c)(3)(ii) provides, in pertinent parts, that a licensee intending to adopt the new rule must, among other things, “modify the fire protection plan required by paragraph (a) of that section to reflect the licensee’s decision to comply with NFPA 805.” Therefore, to the extent that the
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contents of the existing fire protection program plan required by 10 CFR 50.48(a) are inconsistent with NFPA 805, the fire protection program plan must be modified to achieve compliance with the requirements in NFPA 805. All other requirements of 10 CFR 50.48 (a) and GDC 3 have corresponding requirements in NFPA 805.
A comparison of the current requirements in Appendix R with the comparable requirements in Section 3 of NFPA 805 shows that the two sets of requirements are consistent in many respects. This was further clarified in FAQ 07-0032, 10 CFR 50.48(a) and GDC 3 clarification (ML081400292). The following tables provide a cross reference of fire protection regulations associated with the post-transition PINGP fire protection program and applicable industry and PINGP documents that address the topic.
(1) Each holder of an operating license issued under this part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part. This fire protection plan must:
See below
(i) Describe the overall fire protection program for the facility;
NFPA 805 Section 3.2 NEI 04-02 Table B-1
(ii) Identify the various positions within the licensee's organization that are responsible for the program;
NFPA 805 Section 3.2.2 NEI 04-02 Table B-1
(iii) State the authorities that are delegated to each of these positions to implement those responsibilities; and
NFPA 805 Section 3.2.2 NEI 04-02 Table B-1
(iv) Outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage.
NFPA 805 Section 2.7 and Chapters 3 and 4 NEI 04-02 B-1 and B-3 Tables
(2) The plan must also describe specific features necessary to implement the program described in paragraph (a)(1) of this section such as:
See below
(i) Administrative controls and personnel requirements for fire prevention and manual fire suppression activities;
NFPA 805 Sections 3.3.1 and 3.4 NEI 04-02 Table B-1
(ii) Automatic and manually operated fire detection and suppression systems; and
NFPA 805 Sections 3.5 through 3.10 and Chapter 4 NEI 04-02 B-1 and B-3 Tables
(iii) The means to limit fire damage to structures, systems, or components important to safety so that the capability to shut down the plant safely is ensured.
NFPA 805 Section 3.3 and Chapter 4 NEI 04-02 B-3 Table
(3) The licensee shall retain the fire protection plan and each change to the plan as a record until the Commission terminates the reactor license. The licensee shall retain each superseded revision of the procedures for 3 years from the date it was superseded.
NFPA 805 Section 2.7.1.1 requires that documentation (Analyses, as defined by NFPA 805 2.4, performed to demonstrate compliance with this standard) be maintained for the life of the plant. NSPM, “Records Management” procedure (FG-NP-RM-10) and “Records Retention Schedule” (RM-0044).
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(4) Each applicant for a design approval, design certification, or manufacturing license under part 52 of this chapter must have a description and analysis of the fire protection design features for the standard plant necessary to demonstrate compliance with Criterion 3 of appendix A to this part.
Not applicable. PINGP is licensed under 10 CFR 50.
General Design Criterion 3
The PINGP fire protection system was originally designed and constructed in accordance with General Design Criteria 3 as proposed by the Atomic Energy Commission (AEC) and as published in the Federal Register on July 11, 1967. AEC GDC 3 states the following:
“The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features.”
Since the construction of the plant was significantly completed prior to the issuance of the February 20, 1971, 10 CFR 50, Appendix A General Design Criteria, the plant was not reanalyzed and the FSAR was not revised to reflect these later criteria. However, the AEC Safety Evaluation Report acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and “... are satisfied that the plant design generally conforms to the intent of these criteria.” When 10 CFR 50.48 became effective, the NRC’s basic criterion for fire protection as set forth in GDC 3, Appendix A to 10 CFR 50 became applicable to PINGP on October 29, 1980. The applicability of GDC 3 in 10 CFR 50 Appendix A to the PINGP NFPA 805 fire protection program is addressed as follows:
GDC 3, Fire Protection, Statement Applicability/Compliance Reference
Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
NFPA 805 Chapters 3 and 4 NEI 04-02 B-1 and B-3 Tables
Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room.
GDC 3, Fire Protection, Statement Applicability/Compliance Reference
Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety.
NFPA 805 Chapters 3 and 4 NEI 04-02 B-1 and B-3 Tables
Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components
NFPA 805 Sections 3.4 through 3.10 and 4.2.1 NEI 04-02 Table B-3
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(1) Approval of incorporation by reference. National Fire Protection Association (NFPA) Standard 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition” (NFPA 805), which is referenced in this section, was approved for incorporation by reference by the Director of the Federal Register pursuant to 5 U.S.C. 552(a) and 1 CFR part 51.
General Information. NFPA 805 (2001 edition) is the edition used.
(2) Exceptions, modifications, and supplementation of NFPA 805. As used in this section, references to NFPA 805 are to the 2001 Edition, with the following exceptions, modifications, and supplementation:
General Information. NFPA 805 (2001 edition) is the edition used.
(i) Life Safety Goal, Objectives, and Criteria. The Life Safety Goal, Objectives, and Criteria of Chapter 1 are not endorsed.
The Life Safety Goal, Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.
(ii) Plant Damage/Business Interruption Goal, Objectives, and Criteria. The Plant Damage/Business Interruption Goal, Objectives, and Criteria of Chapter 1 are not endorsed.
The Plant Damage/Business Interruption Goal, Objectives, and Criteria of Chapter 1 of NFPA 805 are not part of the LAR.
(iii) Use of feed-and-bleed. In demonstrating compliance with the performance criteria of Sections 1.5.1(b) and (c), a high-pressure charging/injection pump coupled with the pressurizer power-operated relief valves (PORVs) as the sole fire-protected safe shutdown path for maintaining reactor coolant inventory, pressure control, and decay heat removal capability (i.e., feed-and-bleed) for pressurized-water reactors (PWRs) is not permitted.
Feed and bleed is not utilized as the sole fire-protected safe shutdown methodology.
(iv) Uncertainty analysis. An uncertainty analysis performed in accordance with Section 2.7.3.5 is not required to support deterministic approach calculations.
Uncertainty analysis was not performed for deterministic methodology.
(v) Existing cables. In lieu of installing cables meeting flame propagation tests as required by Section 3.3.5.3, a flame-retardant coating may be applied to the electric cables, or an automatic fixed fire suppression system may be installed to provide an equivalent level of protection. In addition, the italicized exception to Section 3.3.5.3 is not endorsed.
Electrical cable construction complies with a flame propagation test that was found acceptable to the NRC as documented in NEI 04-02 Table B-1.
(vi) Water supply and distribution. The italicized exception to Section 3.6.4 is not endorsed. Licensees who wish to use the exception to Section 3.6.4 must submit a request for a license amendment in accordance with paragraph (c)(2)(vii) of this section.
PINGP complies by previous approval as documented in Attachment A, Table B-1.
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(vii) Performance-based methods. Notwithstanding the prohibition in Section 3.1 against the use of performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach; (A) Satisfies the performance goals, performance objectives, and
performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
(B) Maintains safety margins; and (C) Maintains fire protection defense-in-depth (fire prevention, fire detection,
fire suppression, mitigation, and post-fire safe shutdown capability).
The use of performance-based methods for NFPA 805 Chapter 3 is requested. See Attachment L.
(3) Compliance with NFPA 805. See below
(i) A licensee may maintain a fire protection program that complies with NFPA 805 as an alternative to complying with paragraph (b) of this section for plants licensed to operate before January 1, 1979, or the fire protection license conditions for plants licensed to operate after January 1, 1979. The licensee shall submit a request to comply with NFPA 805 in the form of an application for license amendment under § 50.90. The application must identify any orders and license conditions that must be revised or superseded, and contain any necessary revisions to the plant’s technical specifications and the bases thereof. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the licensee has identified orders, license conditions, and the technical specifications that must be revised or superseded, and that any necessary revisions are adequate. Any approval by the Director or the designee must be in the form of a license amendment approving the use of NFPA 805 together with any necessary revisions to the technical specifications.
The LAR was submitted in accordance with 10 CFR 50.90. The LAR included applicable license conditions, orders, technical specifications/bases that needed to be revised and/or superseded.
(ii) The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee’s decision to comply with NFPA 805, before changing its fire protection program or nuclear power plant as permitted by NFPA 805.
The LAR and transition report summarize the evaluations and analyses performed in accordance with Chapter 2 of NFPA 805.
(4) Risk-informed or performance-based alternatives to compliance with NFPA 805. A licensee may submit a request to use risk-informed or performance-based alternatives to compliance with NFPA 805. The request must be in the form of an application for license amendment under § 50.90 of this chapter. The Director of the Office of Nuclear Reactor Regulation, or designee of the Director, may approve the application if the Director or designee determines that the proposed alternatives: (i) Satisfy the performance goals, performance objectives, and
performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
(ii) Maintain safety margins; and (iii) Maintain fire protection defense-in-depth (fire prevention, fire detection,
fire suppression, mitigation, and post-fire safe shutdown capability).
No risk-informed or performance-based alternatives to compliance with NFPA 805 (per 10 CFR 50.48(c)(4)) were utilized. See Attachment P.
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5.2 Regulatory Topics
5.2.1 License Condition Changes
The current PINGP fire protection license condition 2.C.(4) is being replaced with the standard license condition based upon Regulatory Position 3.1 of RG 1.205, modified as shown in Attachment M.
5.2.2 Technical Specifications
NSPM conducted a review of the Technical Specifications to determine which Technical Specifications are required to be revised, deleted, or superseded. NSPM determined that the changes to the Technical Specifications and applicable justification listed in Attachment N are adequate for the PINGP adoption of the new fire protection licensing basis.
5.2.3 Orders and Exemptions
A review was conducted of the PINGP docketed correspondence to determine if there were any orders or exemptions that needed to be superseded or revised. A review was also performed to ensure that compliance with the physical protection requirements, security orders, and adherence to those commitments applicable to the plant are maintained. A discussion of affected orders and exemptions is included in Attachment O.
5.3 Regulatory Evaluations
5.3.1 No Significant Hazards Consideration
A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92. According to 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
• Involve a significant increase in the probability or consequences of an accident previously evaluated; or
• Create the possibility of a new or different kind of accident from any accident previously evaluated; or
• Involve a significant reduction in a margin of safety.
This evaluation is contained in Attachment Q.
Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission’s regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. NSPM has evaluated the proposed amendment and determined that it involves no significant hazards consideration.
5.3.2 Environmental Consideration
Pursuant to 10 CFR 51.22(b), an evaluation of the LAR has been performed to determine whether it meets the criteria for categorical exclusion set forth in 10 CFR
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51.22(c). That evaluation is discussed in Attachment R. The evaluation confirms that this LAR meets the criteria set forth in 10 CFR 51.22(c)(9) for categorical exclusion from the need for an environmental impact assessment or statement.
5.4 Revision to USAR
After the approval of the LAR, the PINGP Updated Safety Analysis Report (USAR) will be revised in accordance with 10 CFR 50.71(e) except as follows. The revised USAR will be submitted in accordance with an exemption from 10 CFR 50.71(e)(4) dated May 22, 2006, which allows periodic updates of the PINGP USAR to be submitted within 6 months after the completion of each Unit 2 refueling outage, not to exceed 24 months from the previous submittal. The format and content will be consistent with FAQ 12-0062.
5.5 Transition Implementation Schedule
The following schedule for transitioning PINGP to the new fire protection licensing basis requires NRC approval of the LAR in accordance with the following schedule:
• Implementation of new NFPA 805 fire protection program will include procedure changes, process updates, and training of affected plant personnel. Implementation will occur within the later of six months after NRC approval, or six months after a refueling outage if in progress at the time of approval. See Attachment S, Table S-3. Note that Implementation Item 20 in Attachment S, Table S-3, is associated with modifications described in Table S-2 and will be completed 180 days after modifications are complete.
• Attachment S, Table S-2, provides a listing of modifications associated with the transition to NFPA 805. NSPM will complete implementation of these modifications at PINGP before the end of the second full operating cycle for each unit after approval of the LAR. Appropriate compensatory measures will be maintained until modifications are complete.
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6.0 REFERENCES
The following references were used in the development of the TR. Additional references are in the NEI 04-02 Tables in the various Attachments.
6.1 NFPA 805, “Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,” 2001 Edition.
6.2 10 CFR 50.48, Fire Protection.
(a) Fire Protection Plans.
(b) Appendix R.
(c) “National Fire Protection Association Standard NFPA 805.”
(f) Decommissioning.
6.3 Federal Register Notice 69 FR 33536, June 16, 2004 (ADAMS Accession Number ML041340086).
6.4 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants.
6.5 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
6.6 10 CFR 50, Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979.
6.7 10 CFR 20, Standards for Protection Against Radiation.
6.8 10 CFR 50.71, Maintenance of Records, Making of Reports.
6.9 10 CFR 50.82, Termination of License.
6.10 10 CFR 50.92, Issuance of Amendment.
6.11 10 CFR 51.22, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review.
6.12 NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c),” Revision 2, April 2008.
6.13 NEI 00-01, “Guidance for Post-Fire Safe Shutdown Circuit Analysis,” Revision 1, January 2005.
6.14 NEI 02-03, “Guidance for Performing a Regulatory Review of Proposed Changes to the Approved Fire Protection Program,” Revision 1, June 2004.
6.15 Branch Technical Position (BTP), Auxiliary Power Conversion Systems Branch (APSCB) 9.5-1, Fire protection.
6.16 SECY 11-0061, A Request to Revise the Interim Enforcement Policy for Fire Protection Issues on 10 CFR 50.48(C) to Allow Licensees to Submit License Amendment Requests in a Staggered Approach (RIN 3150-AG48), June 10, 2011.
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6.17 Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.”
6.18 Regulatory Guide 1.200, “An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities.”
6.19 Regulatory Guide 1.205, “Risk-Informed, Performance-Based Fire Protection for Existing Light Water Nuclear Power Plants,” Revision 1.
6.20 NUREG/CR-6850, “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities.”
6.21 Renewed Operating License, Prairie Island Nuclear Generating Plant, Unit 1, DPR-42, Docket No.50-282.
6.22 Renewed Operating License, Prairie Island Nuclear Generating Plant, Unit 2, DPR-60, Docket No. 50-306.
6.23 Letter from M.A. Schimmel (NSPM) to NRC Document Control Desk, “Request for Extension of Enforcement Discretion and Commitment to Submittal Date for 10 CFR 50.48(c) License Amendment Request, June 22, 2011, ADAMS Accession Number ML111740866.
6.24 Letter from J.G. Giitter (NRC) to M.A. Schimmel (NSPM), “Commitment to Submit a License Amendment Request to Transition to 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, and Request to Extend Enforcement Discretion – Prairie Island Nuclear Generating Plant, Units 1 and 2 (TAC Nos. ME6675 and ME6676),” July 29, 2011, ADAMS Accession Number ML112010417.
6.25 Letter from D. Malone (NMC) to NRC Document Control Desk, “Letter of Intent to Transition to 10 CFR 50.48(c) – National Fire Protection Association Standard NFPA 805, ‘Performance-based Standards for Fire Protection for Light Water Reactor Electric Generating Plants,’ 2001 Edition,” November 30, 2005.
6.26 Letter from C. Haney (NRC) to M. Redderman (NMC),”Letter of Intent to Adopt Title 10 of the Code of Federal Regulations, Part 50.48(c) for Monticello Nuclear Generating Plant, Palisades Nuclear Plant, Point Beach Nuclear Plant, Units 1 & 2, and Prairie Island Nuclear generating Plant, Units 1 and 2 (TAC Nos. MC9289 through MC9294),” September 7, 2006.
6.27 Letter from D.K. Davis (NRC) to L.O. Mayer (NSP), Issuance of License Amendment Nos. 26 (Unit 1) and 20 (Unit 2), Fire Protection Technical Specifications, February 14, 1978.
6.28 Letter from A Schwencer (NRC) to L.O. Mayer (NSP), Issuance of License Amendment Nos. 39 (DPR-42, Unit 1) and 33 (DPR-60, Unit 2), and related Fire Protection Safety Evaluation Report, dated September 6, 1979.
6.29 Letter from A. Schwencer (NRC) to L.O. Mayer (NSPM), Approval of Fire Protection Modifications, April 21, 1980.
6.30 Letter, from R.A. Clark (NRC) to L.O. Mayer (NSP), Approval of Fire Protection Modifications, December 29, 1980.
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6.31 Letter from R.A. Clark (NRC) to L.O. Mayer (NSP), Issuance of License Amendment Nos. 49 (DPR-42, Unit 1) and 43 (DPR-60, Unit 2), and related Safety Evaluation, July 28, 1981.
6.32 Letter from D.C. Dilanni (NRC) to T.M. Parker (NSP), Amendments Nos. 91 and 84 to Facility Operating Licenses Nos. DPR-42 [Unit 1] and DPR-60 [Unit 2]: Technical Specification (TS) Upgrade (TAC Nos. 61081 and 61082), October 27, 1989.
6.33 Letter from B.A. Wetzel (NRC) to R.O. Anderson (NSP), Issuance of Amendments [No. 120, Unit 1, and No. 113, Unit 2] Re: Fire Protection and Detection Systems – Limiting Conditions for Operation (TAC Nos. M89962 and M89963), October 6, 1995.
6.34 Letter from L. Mayer (MSP) to V. Stello (NRC), “Fire Hazard Analysis Report,” March 11, 1977.
6.35 Letter from L. Mayer (NSP) to V. Stello (NRC), “Completion of Fire Protection Review,” July 5, 1977.
6.36 Letter from L. Mayer (NSP) to Director of NRR, “Nuclear Plant Fire Protection Functional Responsibilities, Administrative controls, and Quality Assurance,” May 18, 1978.
6.37 Letter from L. Mayer (NSP) to Director of NRR, “Nuclear Plant Fire Brigade Requirements,” June 22, 1978.
6.38 Letter from L. Mayer (NSP) to Director of NRR, “NRC Staff Evaluation of Fire Protection Program,” January 2, 1979.
6.39 Letter from L. Mayer (NSP) to Director of NRR, “NRC Staff Evaluation of Fire Protection Program,” March 9, 1979.
6.40 Letter from L. Mayer (NSP) to Director of NRR, “NRC Plant Fire Protection Functional Responsibilities, Administrative Controls, and Quality Assurance,” May 2, 1979.
6.41 Letter from, R.A. Clark (NRC) to D.M. Musolf (NSP), Fire Protection - Request for Exemption from a Requirement of Appendix R to 10 CFR Part 50, Section III. G, February 2, 1983.
6.42 Letter from R.A. Clark (NRC) to D.M. Musolf (NSP), Exemption from Appendix R Subsection III.G.2 for Lack of Separation Between Redundant Safe Shutdown Equipment, Fire Area 31, May 4, 1983.
6.43 Letter from J.R. Miller (NRC) to D.M. Musolf (NSP), Exemption from Appendix R Subsection III.G.2 for Lack of Automatic Fire Suppression System, Unit 1, Fire Area 59, January 9, 1984.
6.44 Letter from J.R. Miller (NRC) to D.M. Musolf (NSP), Exemption from Appendix R Subsection III.G.2 for Lack of Separation Between Redundant Safe Shutdown Equipment, and Subsection III.O for Lube Oil Collection Piping, Fire Areas 1 and 71, July 31, 1984.
Northern States Power – Minnesota 6.0 References
PINGP Page 69 – Revision 1
6.45 Letter from J.R. Miller (NRC) to D.M. Musolf (NSP), Issuance of Exemption Re: Certain Technical Requirements of Appendix R to 10 CFR Part 50 (TAC Nos. M89461 and M89462), February 21, 1995.
6.46 Letter from S. Weerakkody (NRC) to A. Marion (NEI), Process for Frequently Asked Questions for Title 10 of the Code of Federal Regulations, Part 50.48(c) Transitions, July 12, 2006 (ADAMS Accession No. ML061660105).
6.47 Regulatory Issues Summary (RIS) 2007-19, Process for Communicating
Clarifications of Staff Positions Provided in RG 1.205 Concerning Issues Identified during the Pilot Application of NFPA Standard 805, August 20, 2007 (ADAMS Accession Number ML071590227).
6.48 PINGP Engineering Evaluation EC 23309, “NFPA 805 LAR Supplement
Attachment A (Table B-1).” [Fundamental Fire Protection Program and Design Elements Review].
6.49 PINGP Engineering Evaluation EC 23946, “NFPA 805 LAR Supplement
6.62 NSPM Procedure FG-NP-RM-10, Records Management.
6.63 NSPM Procedure RM-0044, Records Retention Schedule.
Northern States Power – Minnesota 6.0 References
PINGP Page 70 – Revision 1
6.64 PINGP Fire PRA Uncertainty Notebook, FPRA-PI-UNC.
6.65 PINGP Fire PRA Quantification Notebook, FPRA-PI-FQ.
6.66 PRA Standard, ASME/ANS RA-Sa-2009, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Application.
6.73 NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, Draft report (final report was published July 2012, after the FPRA Peer Review in June 2012).
6.74 “Fire PRA Peer Review of the PINGP Fire Probabilistic Risk Assessment Against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard,” June 2012, attachment to Westinghouse letter to Xcel Energy, LTR-RAM-12-07.
6.75 NRC letter, “Point Beach Nuclear Plant, Units 1 and 2, and Prairie Island Nuclear Generating Plant, Units 1 and 2 - Exemption to 10 CFR 50.71(e)(4) (TAC Nos. MC8654, MC8655, MC8656, and MC8657),” dated May 22, 2006 (ADAMS Accession Number ML061110032).
6.76 Letter from J.P. Sorensen (NSPM) to Document Control Desk (NRC), “License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors,” L-PI-12-089, September 28, 2012 (ADAMS Accession Number ML12278A405).
6.77 Letter from J.E. Lynch (NSPM) to Document Control Desk (NRC), “Supplement to License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors – Submittal of Internal Flooding Peer Review Final Results and Revised Total Plant Risk Values,” L-PI-12-102, November 8, 2012 (ADAMS Accession Number ML12314A144).
6.78 Letter from J.E. Lynch (NSPM) to Document Control Desk (NRC), “Supplement to License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors – Response to Acceptance Review Questions (TAC Nos. ME9734 and ME9735),” L-PI-12-117, December 18, 2012 (ADAMS Accession No. ML12354A464).
6.79 Letter from T.J. Wengert (NRC) to J.E. Lynch (NSPM), “Prairie Island Nuclear Generating Plant, Units 1 and 2 – Acceptance for Review of License Amendment Request for National Fire Protection Association Standard NFPA 805 (TAC Nos.
Northern States Power – Minnesota 6.0 References
PINGP Page 71 – Revision 1
ME9734 and ME9735),” January 2, 2013 (ADAMS Accession No. ML13002A209).
6.80 Letter from J.E. Lynch (NSPM) to Document Control Desk (NRC), “Commitment to Submit a Supplement to the PINGP NFPA 805 License Amendment Request (TAC Nos. ME9734 and ME9735),” L-PI-13-044, May 3, 2013 (ADAMS Accession No. ML13126A115).
6.81 “Focused-Scope Fire PRA Peer Review For The Prairie Island Nuclear Generating Plant,” February 2014, attached to Westinghouse letter LTR-RAM-II-13-117.
6.82 PINGP Engineering Evaluation EC 20819, “Technical Evaluation Associated with Open Circuiting of Current Transformers Contained in 13.8 kV, 4.16 kV, and 480 V Switchgear.”
6.83 PINGP Engineering Evaluation EC 21264, “Loss of DC Control Power Common Enclosure Analysis for NFPA 805.”
6.84 PINGP Engineering Evaluation EC 23408, “NFPA 805 LAR Supplement NEI 00-01 Gap Analysis.”
Northern States Power – Minnesota Attachments
PINGP Page 72 – Revision 0
ATTACHMENTS
PINGP Page A-1 – Revision 1
Northern States Power - Minnesota Attachment A – NEI 04-02 Table B-1 – Transition of Fundamental Fire Protection Program & Design Elements
A. NEI 04-02 Table B-1 – Transition of Fundamental Fire Protection Program & Design Elements
166 Pages Attached
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-2 – Revision 1
NFPA 805 Section #
3.1
Plant Documentation
None
Subsection Title
General
Requirement/Guidance
This chapter contains the fundamental elements of the fire protection program and specifies the minimum design requirements for fire protection program elements and minimum design requirements shall not be subject to the performance-based methods permitted elsewhere in this standard. Previously approved alternatives from the fundamental protection program attributes of this chapter by the AHJ take precedence over the requirements contained herein.
Compliance Statement
N/A
Compliance Basis
10 CFR 50.48(c)(2)(vii) states, "Notwithstanding the prohibition in Section 3.1 against the use of performance-based methods, the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard. Licensees who wish to use performance-based methods for these fire protection program elements and minimum design requirements shall submit a request in the form of an application for license amendment under § 50.90. The Director of the Office of Nuclear Reactor Regulation, or a designee of the Director, may approve the application if the Director or designee determines that the performance-based approach:
(A) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
(B) Maintains safety margins; and
(C) Maintains fire protection defense-in-depth (fire
prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability)."
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-3 – Revision 1
NFPA 805 Section #
3.2.1
Plant Documentation
Subsection Title
Intent
Requirement/Guidance
A site-wide fire protection plan shall be established. This plan shall document management policy and program direction and shall define the responsibilities of those individuals responsible for the plan's implementation. This section establishes the criteria for an integrated combination of components, procedures, and personnel to implement all fire protection program activities.
Compliance Statement
Complies
Compliance Basis
Procedure 5AWI 3.13.0, "Fire Protection Program," establishes a fire protection program for the plant. The procedure defines responsibilities of plant personnel, administrative controls, and implementing documents and procedures in place relative to the program.
Per Section 1.0, "The Fire Protection Program at the Prairie Island Nuclear Generating Plant (PINGP) has been established to protect the health and safety of the public and site personnel, to minimize radioactive release to the environment, minimize property loss, and assure the capability to achieve and maintain safe shutdown conditions in the event of a fire. The Fire Protection Program is an integrated process involving design features, systems, trained personnel, equipment and procedures to provide a defense-in-depth approach to fire protection."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-4 – Revision 1
NFPA 805 Section #
3.2.2
Subsection Title
Management
Policy Direction and Responsibility
Requirement/Guidance
A policy document shall
be prepared that defines management authority and responsibilities and establishes the general policy for the site fire protection program.
Compliance Statement
Complies
Compliance Basis
Per Section 3.0 of Procedure 5AWI 3.13.0, "Fire Protection Program," "This Instruction identifies...The licensing basis associated with the various elements of the PINGP Fire Protection Program...Organizational responsibilities required to maintain the Program in accordance with regulatory requirements and plant commitments…[and] Implementing documents associated with program elements.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-5 – Revision 1
NFPA 805 Section #
3.2.2.1
Plant Documentation
Subsection Title
Management
Policy Direction and Responsibility
Requirement/Guidance
The policy document
shall designate the senior management position with immediate authority and responsibility for the fire protection program.
Compliance Statement
Complies
Compliance Basis
Per Section 7.1.1 of Procedure 5AWI 3.13.0, "Fire Protection Program," The Site Vice President is responsible for, "Overall implementation of the Fire Protection Program in accordance with licensing commitments."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-6 – Revision 1
NFPA 805 Section #
3.2.2.2
Plant Documentation
Subsection Title
Management
Policy Direction and Responsibility
Requirement/Guidance
The policy document
shall designate a position responsible for the daily administration and coordination of the fire protection program and its implementation.
Compliance Statement
Complies
Compliance Basis
Per Section 7.4 of Procedure 5AWI 3.13.0, "Fire Protection Program," "The Fire Protection Program Engineer is the overall single point of contact for the site Fire Protection Program. The Fire Protection Program Engineer has primary responsibility for the program."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-7 – Revision 1
NFPA 805 Section #
3.2.2.3
Plant Documentation
Subsection Title
Management
Policy Direction and Responsibility
Requirement/Guidance
The policy document
shall define the fire protection interfaces with other organizations and assign responsibilities for the coordination of activities. In addition, this policy document shall identify the various plant positions having the authority for implementing the various areas of the fire protection program.
Compliance Statement
Complies
Compliance Basis
Section 7.0 of Procedure 5AWI 3.13.0, "Fire Protection Program," outlines the fire protection organization, which includes 19 positions: Site Vice President, Engineering Director, Engineering Programs Supervisor, Fire Protection Program Engineer, Fire Protection Coordinator, Detection and Alarm Engineer, System Engineer, Design Engineer, Instrument and Control Systems Technician, Appendix R Engineer, Operations Manager, Shift Manager, Shift Supervisors, Work Supervisors, Construction Superintendents, Fire Brigade, Production Planning and Scheduling, Regulatory Compliance, and Training Manager. These positions are responsible for assuring adequate implementation of the various areas of
Section 7.4.5 states that the Fire Protection Program Engineer is responsible for "Interfacing with the respective industry organizations concerning Fire Protection Program issues and
Identifier
None
Items for Implementation None
EEEE Description Summary operating experience."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-8 – Revision 1
NFPA 805 Section #
3.2.2.4
Plant Documentation
Subsection Title
Management
Policy Direction and Responsibility
Requirement/Guidance
The policy document
shall identify the appropriate AHJ for the various areas of the fire protection program.
Compliance Statement
Complies
Compliance Basis
Per Section 5.5 of Procedure 5AWI 3.13.0, "Fire Protection Program," "For interpretation and implementation of NFPA Codes and Standards, both the NRC and NEIL may be considered as the AHJ, dependent upon the situation to which the code is applied."
5.5.1 For application of all Codes of Record to Safety-related areas or other areas under the NRC jurisdiction (or covered under the Fire Protection Program and various commitments), the NRC is considered the primary AHJ. In
Identifier
None
Items for Implementation None
EEEE Description Summary accordance with the fire protection license condition, AHJ authority is delegated to the site when evaluation shows a change or condition has no adverse affect on safe shutdown ability.
5.5.2 For areas where NRC and NEIL share jurisdiction, either or both may be considered the AHJ, dependent on which organization is enforcing the code in any particular instance. The engineer should be cognizant of any regulatory impacts which may occur as a byproduct of a NEIL-related change or evaluation.
5.5.3 For areas outside of NRC jurisdiction, but within the auspices of NEIL, NEIL is considered to be the AHJ.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-9 – Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Subsection 3.2.3(1)
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(1) Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program.
Compliance Statement
Complies
Compliance Basis
Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," establishes requirements for surveillance and testing of fire protection equipment. Section 1.0, Rev. 15 states, "This procedure provides a system overview, functional, requirements, compensatory actions, surveillance requirements, and reporting requirements of fire protection systems."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-10 - Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Subsection 3.2.3(2)
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(2) Compensatory actions implemented when fire protection systems and other systems credited by the fire protection program and this standard cannot perform their intended function and limits on impairment duration.
Compliance Statement
Complies
Compliance Basis
Section 6.1 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," establishes provisions for compensatory measures when fire protection systems and equipment are impaired. Specific compensatory actions are identified for each specific system discussed within the procedure.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-11 - Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Subsection 3.2.3(3)
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(3) Reviews of fire protection program - related performance and trends.
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Per Section 16.0 of Procedure 5AWI 3.13.0, "Fire Protection Program," "Fire Protection Program health is monitored using objective criteria related to the program health criteria including, when practical, objective measures for the program attributes, leading indicators which are identified and emphasized wherever possible to promote a proactive approach to program improvement. Objective criteria indicating health degradation include: Systems performance; Failed PM component tracking; The performance indicator trends are often just as important as the indicator values in assessing program health, as they may represent potential future vulnerabilities; Industry- identified precursors to declining program performance are also potential sources for leading indicators...Detailed Health Report parameters are
included in FP-PE-PHS-01, Program Health Process."
Per Section 5.2.5.5.e of Fleet Procedure FP-PE- PHS-01, "Program Health Process," "For programs that receive regular major NRC inspections (App R and Fire Protection), Focused Self-Assessments are conducted per FP-PA-SA- 01, Focused Self-Assessment Planning, Conduct
Identifier
None
Items for Implementation
EEEE Description Summary and Reporting" at least every 36 months, unless approved by Programs Engineering Director."
The monitoring program required by NFPA 805 Section 2.6 will be implemented after the LAR approval as part of the FPP transition to NFPA 805, in accordance with NFPA 805 FAQ 10- 0059, and will include a process that reviews the FPP performance and trends in performance. Refer to Attachment S.
Per Section 1.0 of Fleet Procedure FP-PA-SA-01, "Focused Self-Assessment Planning, Conduct and Reporting," "This procedure provides the process for planning, conducting and reporting the results of a Focused Self-Assessment (FSA)...The objective of a FSA is to verify compliance, improve performance and achieve excellence."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-12 - Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Subsection 3.2.3(4)
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(4) Reviews of physical plant modifications and procedure changes for impact on the fire protection program.
Compliance Statement
Complies
Compliance Basis
Procedure 5AWI 3.13.1, "Fire Protection Review of Plant Modifications," states, “This Instruction establishes the process for reviewing modifications (engineering changes, equivalency changes, design changes, temporary modifications, etc.) at PINGP. The review ensures that the fire protection requirements are included. It also ensures adequate evaluation and documentation of the type and quantity of combustible loading in each fire area.”
Section 1.0 of Fleet Procedure FP-G-DOC-04, "Procedure Processing," states, "The purpose of this procedure is to establish a common process for initiation, revision, review, and approval of the following document types (listed according to Document Hierarchy): Corporate Directives; Fleet Program/Process Descriptions, Codes of Conduct;
Fleet Procedures; Centralized Department Procedures; Corporate Office Procedures; Site Procedures; Forms (that are independent of procedures)."
Section 1.1 of Fleet Procedure FP-E-MOD-02, "Engineering Change Control," states, "This procedure provides instruction for the initiation, classification and overall control of all
Identifier
None
Items for Implementation None
EEEE Description Summary modifications at facilities owned and operated by Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy, hereinafter NSPM. This procedure also includes instructions and reference information for initiation and control of other types of engineering changes."
Section 1.1 of Fleet Procedure FP-E-MOD-04, "Design Inputs," states, "This procedure controls the identification, documentation, and revision of design inputs throughout the modification design process."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-13 - Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(5) Long-term maintenance and configuration of the fire protection program.
Compliance Statement
Complies
Compliance Basis
Section 16.0 of 5AWI-3.13.0, "Fire Protection Program" states, "It is important that the Fire Protection Program Engineer includes a strategy for updating the program in anticipation of changing requirements and industry standards, or significant business needs, such as plant license renewal. The strategy should be supported by a long-term PI plan that ensures resources, cost, and upgrades are planned far enough in advance to minimize the impact on routine operations and program implementation. The Fire Protection Program should be designed such that the results of health monitoring, benchmarking, self- assessment, license renewal, and management oversight are also reviewed for potential updates to the long-term plans."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-14 - Revision 1
NFPA 805 Section #
3.2.3
Plant Documentation
Subsection Title
Procedures
Subsection 3.2.3(6)
Requirement/Guidance
Procedures shall be
established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:
(6) Emergency response procedures for the plant industrial fire brigade
Compliance Statement
Complies
Compliance Basis
Procedure F5, "Fire Fighting," establishes a procedure for emergency response of the fire brigade.
Section 1.0 states, "The purpose of this section is to provide specific instructions on the organization of fire brigades, individual responsibilities in regard to fires, and procedures for extinguishing fires."
Procedure F5 Appendix A, "Fire Strategies" describes the preplanned actions for fighting fires in each fire area.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-15 - Revision 1
NFPA 805 Section #
3.3
Plant Documentation
Subsection Title
Prevention
Requirement/Guidance
A fire prevention program with the goal of preventing a fire from starting shall be established, documented, and implemented as part of the fire protection program. The two basic components of the fire prevention program shall consist of both of the following:
(1) Prevention of fires and fire spread by controls on operational activities
(2) Design controls that restrict the use of combustible materials The design control requirement listed in the remainder of this section shall be provided as described.
Compliance Statement
Complies
Compliance Basis
Procedure FP-PE-CC-01, "Combustible Control" establishes specific requirements for the fire prevention program.
Section 1.0 states, "This instruction establishes combustible controls requirements that are consistent with regulatory commitments and acceptable industry measures to reduce the potential for a fire that involves combustible materials."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-16 - Revision 1
NFPA 805 Section #
3.3.1
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities
Requirement/Guidance
The fire prevention
program activities shall consist of the necessary elements to address the control of ignition sources and the use of transient combustible materials during all aspects of plant operations. The fire prevention program shall focus on the human and programmatic elements necessary to prevent fires from starting or, should a fire start, to keep the fire as small as possible.
Compliance Statement
Complies
Compliance Basis
Procedure 5AWI 3.13.3, "Hot Work," establishes administrative, procedural, and conditional requirements for hot work and associated fire watches.
Section 1.0 states, "This instruction describes requirements for performing hot work activities during power operations and outages."
Section 2.1 of Procedure FP-PE-CC-01, "Combustible Control" applies to the use, staging and storage of all combustible materials at MNGP (Monticello Nuclear Generating Plant) and PINGP (Prairie Island Nuclear Generating Plant) only as identified in Section 2.2.
Section 10.0 of Instruction 5AWI 3.13.0, "Fire Protection Program," establishes controls for limiting the spread of fire.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-17 - Revision 1
NFPA 805 Section #
3.3.1.1
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities General Fire Prevention Activities, Subsection 3.3.1.1(1)
Requirement/Guidance
The fire prevention
activities shall include but not be limited to the following program elements:
(1) Training on fire safety information for all employees and contractors including, as a minimum, familiarization with plant fire prevention procedures, fire reporting, and plant emergency alarms
Compliance Statement
Complies
Compliance Basis
Procedure 5AWI 3.13.0, "Fire Protection Program," includes but is not limited to training.
Per Section 8.10 of Procedure 5AWI 3.13.0 "Individuals with unescorted access to the plant shall be instructed on how to identify adverse conditions and report them to supervisory personnel. These instructions shall include: Housekeeping and cleanliness criteria...Keeping access to fire extinguishers and hose stations unobstructed...[and] Work management process overview."
Per Section 8.11, Level I (A) fire protection training shall be general training given to operations and mechanical maintenance personnel. The training shall include: Basic
principles of fire chemistry and physics...Fire hazards...Fire detection systems...Types of extinguishing systems...Special fire hazards associated with nuclear power…[and] Emergency planning with emphasis on fire emergency. "Following initial training, Level 1 topics shall be
Identifier
None
Items for Implementation None
EEEE Description Summary reviewed annually with required personnel."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-18 - Revision 1
NFPA 805 Section #
3.3.1.1
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities General Fire Prevention Activities, Subsection 3.3.1.1(2)
Requirement/Guidance
The fire prevention
activities shall include but not be limited to the following program elements:
(2) Documented plant inspections including provisions for corrective actions for conditions where unanalyzed fire hazards are identified.
Compliance Statement
Complies
Compliance Basis
Fleet Procedure FP-PA-ARP-01, "CAP Action Request Process," and Procedure 5AWI 8.5.0, "Housekeeping and Materiel Condition," include but are not limited to processes for documenting plant inspections and corrective actions.
Section 2.1 of Fleet Procedure FP-PA-ARP-01 states, "This procedure establishes the process for documenting and tracking the resolution of issues at each site. It provides the framework to ensure that deviations from performance expectations, including conditions adverse to quality, employee concerns, operability issues,
functionality issues, and reportability issues are promptly identified, evaluated, and corrected as appropriate."
Per Section 1.0 of Procedure 5AWI 8.5.0, "This Instruction establishes Housekeeping and Materiel Condition requirements for the control of work activities, conditions and environment that
Identifier
None
Items for Implementation None
EEEE Description Summary could affect quality. The objective of this program is to encompass all activities related to the control of cleanliness of facilities, cleanliness of material and equipment and protection of equipment."
Per Section 6.14.2, "The area owner SHALL walk down each area monthly. Use the normal site processes to document and correct deficiencies; primarily: • Use a Work Request to initiate action on items requiring maintenance attention; • Use email or face-to-face communication to initiate general housekeeping improvements requiring Nuclear Plant Service Attendant attention; and • Use the Corrective Process to address more significant issues, or those worth trending."
5AWI 3.13.0 Fire Protection Program Section 7.11 of procedure 5AWI 3.13.0 states, “Operations Manager SHALL be responsible for:
7.11.10 Ensuring required daily fire hazard housekeeping inspections are conducted and documented.”
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-19 - Revision 1
NFPA 805 Section #
3.3.1.1
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities General Fire Prevention Activities, Subsection 3.3.1.1(3)
Requirement/Guidance
The fire prevention
activities shall include but not be limited to the following program elements:
(3) Administrative controls addressing the review of plant modifications and maintenance to ensure that both fire hazards and the impact on plant fire protection systems and features are minimized.
Compliance Statement
Complies
Compliance Basis
Procedure 5AWI 3.13.1, "Fire Protection Review of Plant Modifications" includes but is not limited to administrative controls addressing the review of plant modifications and maintenance.
Per Section 1.0 of Procedure 5AWI 3.13.1, “This Instruction establishes the process for reviewing modifications (engineering changes, equivalency changes, design changes, temporary modifications, etc.) at PINGP. The review ensures that the fire protection requirements are included. It also ensures adequate evaluation and documentation of the type and quantity of combustible loading in each fire area.”
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-20 - Revision 1
NFPA 805 Section #
3.3.1.2
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(1)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(1) Wood used within the power block shall be listed pressure- impregnated or coated with a listed fire- retardant application.
Exception: Cribbing timbers 6 in. by 6 in. (15.2 cm by 15.2 cm) or larger shall not be required to be fire- retardant treated.
Compliance Statement
Complies
Compliance Basis
Procedure FP-PE-CC-01, "Combustible Control" includes but is not limited to procedures on the control of wood within the plant. The procedure specifically requires that any wood in the plant shall be coated, treated or covered with a fire retardant material and listed on a Combustible Control Permit.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-21 - Revision 1
NFPA 805 Section #
3.3.1.2
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(2)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(2) Plastic sheeting materials used in the power block shall be fire- retardant types that have passed NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, large-scale tests, or equivalent.
Compliance Statement
Complies
Compliance Basis
Procedure FP-PE-CC-01, "Combustible Control" includes but is not limited to procedures on the control of plastic sheeting.
Per Section 5.1.6 of Procedure FP-PE-CC-01, "Combustible Control", "Plastic sheeting procured for and used in the plant shall be fire retardant unless evaluated by the Fire Marshal (MNGP)/Fire Protection Coordinator (PINGP) and documented in a CCP."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-22 - Revision 1
NFPA 805 Section #
3.3.1.2
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(3)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(3) Waste, debris, scrap, packing materials, or other combustibles shall be removed from an area immediately following the completion of work or at the end of the shift, whatever comes first.
Compliance Statement
Complies
Compliance Basis
Procedure FP-PE-CC-01, "Combustible Control" includes but is not limited to procedures on the control of waste, debris, scrap, packing materials, or other combustibles resulting from an activity.
Per Section 5.1.10 of Procedure FP-PE-CC-01, "Combustible Control", "Combustible packing material or shipping containers SHALL be removed from the area immediately following the unpacking."
Per Section 3.8.5 of Procedure FP-PE-CC-01, "Combustible Control", "Maintain good housekeeping and properly dispose of combustible materials immediately after use or at the end of each shift that are not identified on an approved CCP."
Per Section 6.6.3 of Procedure 5AWI 8.2.0, "Material Identification and Inventory Control," "Housekeeping shall be per 5AWI 8.5.0. Cleanliness and good housekeeping practices
shall be enforced at all times in the storage areas. Storage areas shall be cleaned as required to avoid the accumulation of trash, discarded packaging materials and other detrimental soil."
Identifier
None
Items for Implementation None
EEEE Description Summary
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PINGP Page A-23 - Revision 1
NFPA 805 Section #
3.3.1.2
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(4)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(4) Combustible storage or staging areas shall be designated, and limits shall be established on the types and quantities of stored materials.
Compliance Statement
Complies
Compliance Basis
Procedure FP-PE-CC-01, "Combustible Control" includes but is not limited to procedures on the control of combustible storage or staging areas. Section 5.6 of Procedure FP-PE-CC-01, "Combustible Control" establishes requirements for the storage of combustibles in designated areas.
Per Section 6.6.5 of Procedure 5AWI 8.2.0, "Material Identification and Material Control," "Fire protection for stored materials shall be per 5AWI 3.13.2."
Per Section 6.8.1 of Procedure 5AWI 8.5.0, "Housekeeping and Materiel Condition," "Combustible material (including gases and liquids, high efficiency particulate air and charcoal filters, dry ion exchange resins, or other combustible supplies) SHALL be stored in approved cabinets
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NFPA 805 Section #
3.3.1.2
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(5)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(5) Controls on use and storage of flammable and combustible liquids shall be in accordance with NFPA 30, Flammable and Combustible Liquids Code, or other applicable NFPA standards.
Compliance Statement
Complies
Compliance Basis
Section 5.3 of Procedure FP-PE-CC-01, "Combustible Control" establishes requirements for the storage of flammable and combustible liquids.
The storage of flammable and combustible liquids has been reviewed against the requirements of NFPA 30, as detailed in the NFPA 30–1969 and NFPA 30–1987 code review checklists.
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NFPA 805 Section #
3.3.1.2
Plant Documentation
Subsection Title
Fire Prevention for
Operational Activities Control of Combustible Materials, Subsection 3.3.1.2(6)
Requirement/Guidance
Procedures for the
control of general housekeeping practices and the control of transient combustibles shall be developed and implemented. These procedures shall include but not be limited to the following program elements:
(6) Controls on use and storage of flammable gases shall be in accordance with applicable NFPA standards.
Compliance Statement
Complies
Compliance Basis
Fire Protection Engineering Evaluation FPEE-11- 018 documents the code compliance review for National Fire Protection Association 55, (NFPA) – 2005, Standard for the Storage, Use, and Handling of Compressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks. This review concluded that the use and storage of flammable gases is compliant with the applicable code requirements.
Fire Protection Engineering Evaluation FPEE-11-018, National Fire Protection Association 55, (NFPA) – 2005, Standard for the Storage, Use, and Handling of Compressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks
Industry-Related References
FAQ 06-0020, "Identification of 'Applicable NFPA Standards'," Rev. 1, dated 2/16/07 NFPA 55, "Standard for the Storage, Use, and Handling of Compressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks," 2005 Edition
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NFPA 805 Section #
3.3.1.3.1
Plant Documentation
Subsection Title
Control of Ignition
Sources
Requirement/Guidance
A hot work safety
procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Procedure 5AWI 3.13.3, "Hot Work," establishes guidance and requirements for those performing hot work. Section 1.0 states, "This Instruction describes requirements for performing hot work activities during power operations and outages."
Hot work safety procedures have been reviewed against the requirements of NFPA 51B and NFPA 241, as detailed in the NFPA 51B–1999 code review checklist and FPEE-11-020.
Procedure 5AWI 3.13.3, "Hot Work," Rev 3, dated 1/27/12 FPEE-11-020, "NFPA Codes Referenced in NFPA 805 not addressed by separate code reviews"
Industry-Related References
NFPA 51B, "Standard for Fire Prevention during Welding, Cutting, and other Hot Work," 1999 Edition
FPEE-11-020, NFPA Codes Referenced in NFPA 805 not addressed by separate code reviews
EEEE Description
The purpose of this analysis is to document the review of the NFPA codes referenced in NFPA 805 that are not addressed in separate NFPA code compliance reviews
Summary
The following sections of NFPA 241 are not applicable to PINGP, with the bases identified:
Section 5.1.2: PINGP does not have gas operated welding and cutting equipment that uses multiple oxygen and fuel gas cylinders.
Section 5.1.3.2 (partial): Per discussions with station personnel, and based on review of recent roofing modifications, torch applied roof coverings have not been used at PINGP.
Section 5.1.4: Thermit welding is not conducted at PINGP.
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Chapter 9: Per discussions with station personnel, and based on review of recent roofing modifications, torch applied roof coverings have not been used at PINGP.
Section 5.1.1 of NFPA 241 identifies that hot work shall be in accordance with NFPA 51B. In addition, the fire watch requirement of Section 5.1.3.1 and the partial criteria of Section 5.1.3.2 regarding posting of a fire watch for the duration of the work are also covered in NFPA 51B.
The code compliance review of NFPA 51B is addressed in FPEE-11-021. Only one deviation is identified. Section 5.1 of FPEE-11-021 documents the acceptability of not performing a final check after completion of the work in Designated Hot Work Areas that do not require a fire watch. As such, NFPA 51B adequately addresses all hot work criteria of NFPA 241 except for torch applied roofing.
Items for Implementation As described in Table S-3, Item #2, revise procedure 5 AWI 3.13.3, “Hot Work,” to address the following: - There is no requirement in plant documents that specifically address the requirements for hot tapping. (NFPA 51B–1999, Section 3-5) - Plant documents do not address the requirements for a fire watch where torch-applied roofing hot work operations are in effect. (NFPA 241–1999, Section 5.1.3.2)
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NFPA 805 Section #
3.3.1.3.2
Plant Documentation
Subsection Title
Control of Ignition
Sources
Requirement/Guidance
Smoking and other
possible sources of ignition shall be restricted to properly designated and supervised safe areas of the plant.
Compliance Statement
Complies
Compliance Basis
Per Section 3.15.9 of Procedure CD 5.13, “Fire Protection Program Standard," - "Designate smoking areas and areas where smoking is prohibited such as in areas containing safety related/safe shutdown equipment.”
Procedure CD 5.13, “Fire Protection Program Standard”, Rev, dated 01/17/14
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NFPA 805 Section #
3.3.1.3.3
Plant Documentation
Subsection Title
Control of Ignition
Sources
Requirement/Guidance
Open flames or
combustion-generated smoke shall not be permitted for leak or air flow testing.
Compliance Statement
Complies with item for implementation
Compliance Basis
Per Item 73 of the table attached to Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76, "Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1," "Leak testing is done using only smoke sticks or generators."
Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76 Procedure 5AWI 3.13.2, "Fire Prevention," Rev 19, dated 01/05/12
As described in Table S-3, item #2, Procedure 5AWI 3.13.3, "Hot Work," will be revised
to require that Open flames or combustion-generated smoke shall not be permitted for leak
or air flow testing
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NFPA 805 Section #
3.3.1.3.4
Plant Documentation
Subsection Title
Control of Ignition
Sources
Requirement/Guidance
Plant administrative
procedure shall control the use of portable electrical heaters in the plant. Portable fuel-fired heaters shall not be permitted in plant areas containing equipment important to nuclear safety or where there is a potential for radiological releases resulting from a fire.
Compliance Statement
Complies
Compliance Basis
Section 6.8 of Procedure 5AWI 3.13.3, "Hot Work," provides the requirements that shall apply to portable space heaters.
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NFPA 805 Section #
3.3.2
Plant Documentation
Subsection Title
Structural
Requirement/Guidance
Walls, floors, and
components required to maintain structural integrity shall be of noncombustible construction, as defined in NFPA 220, Standard on Types of Building Construction.
Compliance Statement
Complies
Compliance Basis
Per Section 2.6 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Materials of construction are rated 'Noncombustible' or 'Fire-Resistive' construction in accordance with National Fire Protection Association (NFPA) No. 220—1961. Coating systems applied to structures exhibit flame spread characteristics which are within the limits of the definition for 'Noncombustible' contained in NFPA No. 220."
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NFPA 805 Section #
3.3.3
Plant Documentation
Subsection Title
Interior Finishes
Requirement/Guidance
Interior wall or ceiling
finish classification shall be in accordance with NFPA 101®, Life Safety Code®, requirements for Class A materials. Interior floor finishes shall be in accordance with NFPA 101 requirements for Class 1 interior floor finishes.
Compliance Statement
Complies
Compliance Basis
Per Section 4.2 of Procedure D71.1, "General Wall Painting, Concrete Wall, Blockwalls and Floor Surfaces," "All coatings shall meet the requirements of 'Class A Interior Wall and Ceiling Finish' as defined in the National Fire Code, Section 101, 6-5.3.1, 1994 Edition."
Per Section 2.6 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Coating systems applied to structures exhibit flame spread characteristics which are within the limits of the definition for 'Noncombustible' contained in NFPA No. 220.
Coating systems in the Containment, Safety Injection, Containment Spray, and Residual Heat Removal pump rooms use Carboline #2 primer and phenoline 305 which have a flame spread rating of 0-25."
Per Section 4.1 of Maintenance Procedure D71 "Nuclear Coating Applications," "In containment, use only those coating systems qualified according to ANSI 101.2 and referenced in section 2.8."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per Section 2.8 of Procedure D71, "Nuclear Coating Application," Qualified Coatings for use on the interior of reactor buildings are Carbo Zinc 11 SG, Carboguard 890N (Formerly called Carboline 890 and Carboguard 890); Phenoline 368 WG; and Carboguard 2011S top coated with Carboguard 890N.
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NFPA 805 Section #
3.3.4
Plant Documentation
Subsection Title
Insulation Materials
Requirement/Guidance
Thermal insulation
materials, radiation shielding materials, ventilation duct materials, and soundproofing materials shall be noncombustible or limited combustible.
Compliance Statement
Complies
Compliance Basis
Per Section 1.1 of Engineering Manual 3.2.1.7, "Specification for Thermal Insulation," "The scope of this specification shall be to provide the required design information as needed for the replacement, modification, and/or addition of thermal insulation on previously existing plant piping, valves, fittings, and equipment."
Per Section 1.5, "Insulation materials including
Engineering Manual 3.2.1.7, "Engineering Design Standard for Specification for Thermal Insulation," Rev. 1, dated 10/31/2009. Ductwork Construction of Pioneer Services &Engineering Co, Std Spec –Mech (SS - 613 (NSP) 9-70, “Standard Specification for Sheet Metal Ductwork Low Pressure and Medium Pressure”, SS 613, Low and Medium Pressure Ductwork, SS 614 (HVAC for Turb, Aux, Screenhouse, Containment and Shield Building), and SS 619 for High Pressure Sheet Metal Ductwork. FPEE (AR# 01163081, Rev. 1) documents the adequacy of the soundproofing material installed in Access Control, the Control Room, and the Cable Spreading Room computer area. Nuclear Power Outfitters Lead Blanket Specifications
Industry-Related References California Department of Forestry and Fire Protection, Flame Retardant Fabrics and Chemicals Program NFPA 701, Standard Methods of Fire Tests for Flame Propagation of Textiles and Films, 2010 Edition CPAI-84, Industrial Fabrics Association International (formerly the Canvas Products Association), Specification For Flame Resistant Materials Used in Camping Tentage, 1995 Edition
adhesives, shall be non-combustible and of a fire resistant application. This application shall be U.L. rated with a normal maximum U.L. rating as follows: Flame spread 25, smoke developed 50, fuel contribution 25."
Per Section 4.0, Ductwork Construction of Pioneer Services & Engineering Co, Std Spec –Mech (SS - 613 (NSP) 9-70), “Standard Specification for Sheet Metal Ductwork Low Pressure and Medium Pressure”, “All ductwork shall be constructed prime quality galvanized material except as called out in TS - M 614, Technical Specification for Ventilating Systems. Ductwork shall be inside smooth, air – tight substantially supported and stayed sufficiently to avoid reverberations or flutter.”
The basis for the limited combustible soundproofing material in Access Control is provided in FPEE (AR#01163081). “Based on the definition of “non-combustible/limited combustible” the fire test performance of the material used for sound proofing in the Access Control Center is acceptable and will not impact the ability to achieve and maintain safe shutdown.”
Per the Manufacturer (Nuclear Power Outfitters), the Radiation Shielding material is flame retardant and meets California Fire Marshall requirements, NFPA 701 and CPAI 84.
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NFPA 805 Section #
3.3.5.1
Plant Documentation
Subsection Title
Electrical
Requirement/Guidance
Wiring above suspended
ceiling shall be kept to a minimum. Where installed, electrical wiring shall be listed for plenum use, routed in armored cable, routed in metallic conduit, or routed in cable trays with solid metal top and bottom covers.
Compliance Statement
Complies
Compliance Basis
12/8/1976 Letter, L.O. Mayer (NSP) to V. Stello (NRC), Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1, APCSB 9.5-1 IV.B.1 (f ) states "Room or areas containing safeguard equipment do not have suspended ceilings. The control room has a suspended ceiling of “aluminum crate" design".
Updated Safety Analysis Report (USAR) Section 7, Rev. 31, states "...The control room suspended ceiling consists of aluminum louver grids with fluorescent light fixtures mounted above the grids..”
Updated Safety Analysis Report (USAR) Section 7, Rev. 31 12/8/1976 Letter, L.O. Mayer (NSP) to V. Stello (NRC) Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1
Based on this information, there is no wiring located above solid suspended ceilings.
Identifier
None
Items for Implementation None
EEEE Description Summary
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NFPA 805 Section #
3.3.5.2
Plant Documentation
Subsection Title
Electrical
Requirement/Guidance
Only metal tray and
metal conduits shall be used for electrical raceways. Thin wall metallic tubing shall not be used for power, instrumentation, or control cables. Flexible metallic conduits shall only be used in short lengths to connect components.
Compliance Statement
Complies
Compliance Basis
Per Sheet 1 of the Unit 1 & 2 Project Design Manual, dated 4/1967, Section 4, Index 324.53, "Electrical Cable Tray Systems," "The cable tray system shall be fabricated from aluminum alloys throughout, except in the containment vessel. The use of aluminum is prohibited in the containment vessel, and only galvanized steel materials for this cable tray system is allowed." Per Section 9.2.2 of Procedure SWI CON-3, "Raceway Installation," cable trays are pan or ladder style, and conduits are steel or aluminum.
Per Section 4.1.1 of Engineering Manual 4.3.1- C.4, "Engineering Design, Fabrication and Installation Summary for Conduit," "Conduit shall be RMC [Rigid Metal Conduit] steel. EMT [Electrical Metallic Tubing] may be used only for lighting or telephones. Aluminum conduit may be specified by the Engineer for special applications." Per Section 4.1.2, "Steel RMC shall be rigid conduit conforming to ANSI Standard C80.1-1966
Identifier
None
Items for Implementation None
EEEE Description Summary (Rev. 1971) rigid steel, zinc-coated." Per Section 4.1.3, "Aluminum conduit is prohibited for use inside containment." Per Section 4.1.4, "All flexible metal conduits for outdoor and high moisture installations shall be liquid tight galvanized steel flexible conduits covered with a synthetic outer jacket such as Anaconda Metal Hose Co., "Sealtight." Sizes 1/2" through 4" will be Anaconda Type "UA"; sizes 5" and 6" shall be Anaconda Type "EF" or approved equal." Per Section 5.1.4, "Rigid conduits may terminate directly on equipment only when both conduit and equipment are supported on the same wall, column, ceiling, floor, or within five (5) feet of adjoining (corner) walls. Flexible conduit, or armored cable/jacketed cable (with appropriate connector) shall be utilized in all other cases unless specified otherwise. "Flexible conduit is required, allowing for two (2) inches of movement, in conduit runs between the Reactor Shield Building and Auxiliary Building Structures."
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Per Section 5.3.1, "Flexible metal conduit shall be of sufficient length to provide adequate movement due to seismic, vibration, and thermal motions.
The sketches in Engineering Manual 4.3.1-C.4 Sections 4.3.1-C.4.2 through 4.3.1-C.4.7 identify the maximum allowable lengths of flexible conduits for various configurations. These allowable lengths minimize the use of flexible conduits.
Per Section 7.8.4 of the Updated Safety Analysis Report (USAR), "To prevent the spread of fire from the electrical rooms beneath the control room, the following provisions are made: a. Cables used throughout the relay room have an exterior jacket that meets the insulated Power Cable Engineer's Association (IPCEA) test requirements. All non-metallic materials in the cable construction and accessory devices have been chosen so that they will not support combustion. Power cables for the 480 volt system are three conductor, insulated with ethylene propylene rubber, with a neoprene-like, or an asbestos braid impregnated with flame retardant compound jacket and interlocked aluminum armor overall. b. Cabling and wiring in the relay control room are installed in trays or in metallic conduit."
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NFPA 805 Section #
3.3.5.3
Plant Documentation
Subsection Title
Electrical
Requirement/Guidance
Electric cable
construction shall comply with a flame propagation test as acceptable to the AHJ.
Compliance Statement
Complies
Compliance Basis
Per Section 5.2.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "The majority of cable used in the plant was purchased from Kerite Company. The cable is ethylene propylene rubber insulated with a chlorosulfated polyethylene jacket and is commonly known as EPR-Hypalon.
Laboratory tests showed this cable to be highly resistive to auto-ignition and flame propagation. This cable has subsequently been qualified in accordance with IEEE 383-1974 flame test. A small amount of cable purchased from Boston Insulated Wire and Cable company was tested in accordance with the Philadelphia flame test which
Identifier
None
Items for Implementation None
EEEE Description Summary is a predecessor to the oil soaked rag test of IEEE 383-1974. The cables used in the D5/D6 addition was qualified to IEEE 383-1974 as above."
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NFPA 805 Section #
3.3.6
Plant Documentation
Subsection Title
Roofs
Requirement/Guidance
Metal roof deck
construction shall be designed and installed so the roofing system will not sustain a self- propagating fire on the underside of the deck when the deck is heated by a fire inside the building. Roof coverings shall be Class A as determined by tests described in NFPA 256, Standard Methods of Fire Tests of Roof Coverings.
Compliance Statement
Complies with
Clarification
Compliance Basis
NFPA 256 was withdrawn at the Annual 2008 Meeting with no bases for where equivalent criteria are addressed. The following documents demonstrate compliance that the roofing system will not sustain a self-propagating fire on the underside of the deck when the deck is heated by a fire inside the building.
Per Section 4, Account No. 321.244 of the PINGP Project Design Manual, dated April 1967, original specifications for roofing materials identified as "Insulated flat steel deck with built up roofing meeting NBFU requirements for 'Roof Deck Construction No. 1'."
Per Letter from Severson (Howard R. Green Company) to Samson (NSP) dated 5/13/99,
PINGP Project Design Manual, dated April 1967, Account No. 321.244 Design Change 98BM01, "Rad Waste & Service Building Roofs," Rev. 0 Drawing NF-38385, "Reactor Building Unit 1 Concrete Wall Detail and Dome Plan, Sections & Details," Rev. M Drawing NF-38513, "Architectural Plant Roof Plan," Rev. T, dated 1/21/05 Drawing NF-116987, "D5/D6 Bldg. - Concrete Roof Plan at El. 755'-0"," Rev. A Letter from Severson (Howard R. Green Company) to Samson (NSP) dated 5/13/99 Modification 96BM01, "Ancillary Roof Replacement," Rev. 0, Project Description Modification 97BM01, "Turbine Roof Replacement," Rev. 0, Project Description Modification 97BM02, "1998 Reroofing," Rev. 0, Project Description
Industry-Related References
NFPA 256, "Standard Methods of Fire Tests of Roof Coverings," 1998 Edition Underwriters' Laboratories Fire Resistance Directory
"According to the Uniform Building Code (UBC), the existing gravel surfaced built-up roof membrane is considered a Class A roofing system. The new proposed system will also be Class A…According to the 'UL Fire Resistance Directory', the new built-up roof system meets assembly P#508. This indicates that the new system combined with the existing structure has a proper fire rating."
Per Section 6 of Design Change 98BM01, "Rad Waste & Service Building Roofs," "According to the UL Fire Resistance Directory, the new built up roof system meets Assembly No. P508 and No. J928. This indicates that the new system, combined with the existing structure, has a proper
Identifier
None
Items for Implementation None
EEEE Description Summary fire rating."
Per Page 5 of 6 of the Project Description for Modification 97BM01, "Turbine Roof Replacement," "According to the UL Fire Resistance Directory, the new built up roof system meets Assembly No. P508. This indicates that the new system, combined with the existing structure, has a proper fire rating."
Per Pages 1 of 7 and 2 of 7 of the Project Description for Modification 97BM02, "1998 Reroofing," "The purpose of this modification is to remove the existing Aux Building high roof, Old Admin Bldg and Rad Monitoring Station roofing and replace it with new, in accordance with the
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original design specification…According to the UL Fire Resistance Directory, the new built up roof system meets Assembly No. P508 and J928. This indicates that the new systems, combined with the existing structure, have an appropriate fire rating."
Per Pages 1 of 7 and 5 of 7 of the Project Description for Modification 96BM01, "Ancillary Roof Replacement," "The scope of the project includes removal and replacement of the following roof and roof systems: High roof of the old Screenhouse; Trash Basket roof; Chlorine House roofs; Maintenance Building Roof; Unit 1 and Unit 2 Condensate Polishing roof…The new proposed roof system, consisting of a ballasted 60 mil EPDM single ply membrane, expanded polystyrene and rigid insulation, is also considered to be a Class A roofing system…According to the UL Fire Resistance Directory, a ballasted single ply assembly system by Carlisle Syntec Systems meets Assembly Number P213. This indicates that the new system, combined with the existing structure, has a proper fire rating."
Per Drawing NF-116987, "D5/D6 Bldg. - Concrete Roof Plan at El. 755'-0"," the D5/D6 Building roof is 18" concrete slab over metal decking.
Per Drawing NF-38385, "Reactor Building Unit 1 Concrete Wall Detail and Dome Plan, Sections & Details," the Shield Building dome roof is of 24" reinforced concrete.
Per Drawing NF-38513, "Architectural Plant Roof Plan," "All roofs as shown & noted shall carry a 20 year bond & shall be UL rated Class A Type III construction, unless otherwise noted."
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NFPA 805 Section #
3.3.7
Plant Documentation
Subsection Title
Bulk Flammable
Gas Storage
Requirement/Guidance
Bulk compressed or
cryogenic flammable gas storage shall not be permitted inside structures housing systems, equipment, or components important to nuclear safety.
Compliance Statement
Complies
Compliance Basis Per Section 6.2 of Procedure 5AWI 3.13.1, "Fire Protection Review of Modifications," the plant fire protection engineer is responsible to ensure that plant modifications, including storage of flammable gases, are in compliance with the plant's fire protection program. There is no bulk storage of flammable gases inside structures with SSCs important to nuclear safety.
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NFPA 805 Section #
3.3.7.1
Plant Documentation
Subsection Title
Bulk Flammable
Gas Storage
Requirement/Guidance
Storage of flammable
gas shall be located outdoors, or in separate detached buildings, so that a fire or explosion will not adversely impact systems, equipment, or components important to nuclear safety. NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be followed for hydrogen storage.
Compliance Statement
Complies with
Clarification
Compliance Basis
Section 3.1.6 of Attachment to Letter from Schwencer (NRC) to Mayer (NSP), dated 4/21/80, states, "The licensee has shown to our satisfaction that the combustible gas cylinders will be stored in locations separated from safety related areas by three hour fire rated barriers. The piping and cylinder installation will be in accordance with NFPA No. 58. Based on our review, we find the licensee's method of storing combustible gas cylinders acceptable." The requirement guidance cites NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites; however this standard has been withdrawn and not superseded. NFPA 55, 2005,
Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80 Drawing NF-38225, “Turbine Room – Concrete Hydrogen House,” Rev. F, dated 12/26/74 Drawing NF-38504, “Architectural Power Plant Sections,” Rev. F, dated 12/10/81 Procedure C40.1, “H2/O2 Gas Systems,” Rev. 9, dated 8/31/09 Procedure F5 Appendix A, “Fire Strategies,” Rev. 27, dated 11/17/11 System Description B40, “Miscellaneous Gas Systems,” Rev. 7, dated 5/27/09 PINGP Unit 1 & 2 Project Desigccn Manual, dated 4/1967 FPEE-11-018, NFPA 55 - 2005, Standard for the Storage, Use, and Handling ff Com Pressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks Contract No. 00027187 between PINGP and Praxair Distribution, Inc.
Industry-Related References
NFPA 55 Standard for the Storage, Use, and Handling of Compressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks–2005
Standard for the Storage, Use, and Handling of Compressed Gases and Cryogenic Fluids in Portable and Stationary Containers, Cylinders, and Tanks (Code of Record) has been used as the design basis for the bulk hydrogen storage system. Hydrogen storage has been reviewed against the requirements of NFPA 55, as detailed in the NFPA 55–2005 code compliance report, FPEE-11-018. Contract No. 00027187 between PINGP and Praxair Distribution includes an annual system inspection.
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-44 - Revision 1
NFPA 805 Section #
3.3.7.2
Plant Documentation
Subsection Title
Bulk Flammable
Gas Storage
Requirement/Guidance
Outdoor high-pressure
flammable gas storage containers shall be located so that the long axis is not pointed at buildings.
Compliance Statement
Complies
Compliance Basis
Per Engineering Change 12191 close out package (EC-0441), The three tube banks will be positioned side by side with the valving facing east. In the event a nozzle break occurs, the projectile generated would be propelled to the east, away from the power block. The individual tubes were not considered as missiles because
Engineering Change 12191 close out package (EC-0441)
Industry-Related References
NFPA 55, Compressed Gases and Cryogenic Fluids Code
they are fastened to their support structure and are considered too heavy to act as missiles. Additionally, there are no safety related SSCs in the travel path for the tube path. These tubes also would be unable to achieve the same level of penetration compared to smaller missiles
Identifier
None
Items for Implementation None
EEEE Description Summary considered in the USAR due to their large weight and size.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-45 - Revision 1
NFPA 805 Section #
3.3.7.3
Plant Documentation
Subsection Title
Bulk Flammable
Gas Storage
Requirement/Guidance
Flammable gas storage
cylinders not required for normal operation shall be isolated from the system.
Compliance Statement
Complies
Compliance Basis
Per Table 2 of Procedure 5AWI 8.5.0, "Housekeeping and Material Condition," it is expected that "All gas cylinders are secured appropriately."
Per Section 6.8.1, "Combustible material
Procedure 5AWI3.13.3, "Hot Work," Rev 3, dated 1/27/12 Procedure 5AWI 8.5.0, "Housekeeping and Material Condition," Rev. 10, 2/25/09 System Description B40, "Miscellaneous Gas Systems," Rev. 7, dated 5/27/09
(including gases and liquids…) shall be stored in approved cabinets and containers or in posted areas.
Per Section 10.0 of Procedure 5AWI 3.13.3, "Hot Work," "Cylinder use and storage shall be in accordance with applicable site safety policies."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per the Nuclear Plant Pocket Safety Guide, compressed gas cylinder valve protection caps shall be in place at all times except when regulators are attached (when cylinders are in use).
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-46 - Revision 1
NFPA 805 Section #
3.3.8
Plant Documentation NRC SER dated 9/6/79
Subsection Title Bulk Storage of Flammable and Combustible Liquids
Requirement/Guidance Bulk storage of flammable and combustible liquids shall not be permitted inside structures containing systems, equipment, or components important to nuclear safety. As a minimum, storage and use shall comply with NFPA 30, Flammable and Combustible Liquids Code.
Compliance Statement Complies
Compliance Basis Section 5.1.2 (b) of Procedure FP-PE-CC-01, "Combustible Control" establishes requirements for the storage of flammable and combustible liquids.
Bulk storage of flammable and combustible liquids has been reviewed against the requirements of NFPA 30, as detailed in the NFPA 30–1969 and NFPA 30–1987 code review checklists.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-47 - Revision 1
NFPA 805 Section #
3.3.9
Plant Documentation
Subsection Title Transformers
Requirement/Guidance Where provided, transformer oil collection basins and drain paths shall be periodically inspected to ensure that they are free of debris and capable of performing their design function.
Compliance Statement Complies
Compliance Basis Per Table 1 of Procedure 5AWI 8.5.0, "Housekeeping and Materiel Condition," all plant transformers are included in the plant's inspection zone program. Per Table 2, this inspection program includes ensuring that drains are not clogged. Per Section 6.14.2, these inspections are performed monthly.
Per Section 10.2 of Procedure D14.6, "Storm Water Pollution Prevention Plan":
Procedure 5AWI 8.5.0, "Housekeeping and Materiel Condition," Rev. 10, dated 2/25/10 Procedure D14.6, "Storm Water Pollution Prevention Plan," Rev. 5, dated 3/22/06 NF-39320-1, Transformer Oil Drain Piping Unit 1 NF-39321-1, Transformer Oil Drain Piping Unit 2
"The preventative maintenance program consists of various actions taken on a routine basis, i.e., checking transformers, changing oil absorbents, and pumping berms, inspections by the environmental group.
10.2.1 The oil demister (roof of Auxiliary Building) has its’ oil absorbents changed on a computer generated schedule.
Identifier
None
Items for Implementation None
EEEE Description Summary
10.2.2 The berms surrounding the above ground storage tanks are inspected weekly. A written log entry is completed which documents that, if the berm contained water, it was inspected for oil residue before it was pumped or drained. This includes the large berm pits for the plant transformers which are also inspected for oil residue before pumping.
10.2.3 The plant transformers have a continuing leakage problem due to their large size. The various types of oil absorbents placed under them are routinely inspected and changed as necessary. This minimizes the chance of oil reaching the berm pits."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-48 - Revision 1
NFPA 805 Section #
3.3.10
Plant Documentation
Subsection Title
Hot Pipes and
Surfaces
Requirement/Guidance
Combustible liquids,
including high flashpoint lubricating oils, shall be kept from coming in contact with hot pipes and surfaces, including insulated pipes and surfaces. Administrative controls shall require the prompt cleanup of oil on insulation.
Compliance Statement
Complies
Compliance Basis
Per Section 5.3.6 of Procedure FP-PE-CC-01, "Combustible Control", "Liquids, including oils, SHALL be cleaned in a timely manner when spilled onto insulation."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-49 - Revision 1
NFPA 805 Section #
3.3.11
Plant Documentation
Subsection Title
Electrical
Equipment
Requirement/Guidance
Adequate clearance, free of combustible material, shall be maintained around energized electrical equipment.
Compliance Statement
Complies
Compliance Basis
Per Section 5.1.12 of Procedure FP-PE-CC-01, "Combustible Control" - "Combustible materials SHALL NOT be stored within three (3) feet of energized electrical cabinets."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-50 - Revision 1
NFPA 805 Section #
3.3.12(1)
Plant Documentation
Subsection Title
Reactor Coolant
Pumps
Requirement/Guidance
For facilities with non-
inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.
(1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oil from all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil system.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix E, "The Reactor Coolant Pump Lube Oil Collection System is piped to sump “A” inside containment. The contents of the sump can be pumped to closed vented tank(s) inside the Auxiliary Building via two (2) alternative flow paths…The sump in the basement of containment is a concrete pit having a capacity of 990 gallons. This is more than the capacity needed to contain the total inventory of lube oil (410 gallons) for the two (2) reactor coolant pumps for each unit…The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the Control Room if this level is exceeded. The operator can initiate manual control of the sump pump(s) at any time by overriding the automatic control of sump level."
Per Page 2 of Attachment 1 of Letter from Musolf (NSP) to Director (NRC) dated 4/5/84, "In support of the system as is currently installed it should be noted…the lube oil collection system is seismically designed." Per Page 2 of Attachment 1 of Letter from NSP to NRC dated 4/5/84, "In summary, Northern States Power has made an extensive effort to comply with the requirements of Appendix R. In comparing the lube oil collection system to the requirements of Section III.0 [of Appendix R], concerns were voiced over the use
Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84 Letter from Musolf (NSP) to Director (NRC) dated 4/5/84 Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," Rev. 14, dated 4/29/10
of a closed vented container inside containment because of the need for it to also act as a collection point for seal leakage. Northern States Power believes that the existing configuration meets the intent of Appendix R in that all lube oil is collected to a common point which will prevent its contact with hot piping in the area and is isolated from electrical power cable which might cause ignition."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per Page 5 of Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84, "the pipe from the sump to the vented container in the auxiliary building has been designed to seismic category Class III which meets the requirement of Regulatory Guide 1.29, paragraph C-2. If failure of this pipe were to occur during a seismic event, the functions of plant features described in paragraph 1 (a through q) of Regulatory Guide 1.29 will not be affected and the plant can be brought to cold shutdown." Per Page
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-51 - Revision 1
6 of Letter from NRC to NSP dated 7/31/84, "We agree with the licensee that, although lube oil leakage is collected in the sump before it is pumped to a vented container, the sump design at this plant assures us that oil collected there will not lead to fire during normal or design basis accident conditions. The capacity of the sump and the vented containers is adequate to safely contain any anticipated lube oil leakage and the controls provide reasonable assurance that any lube oil collected in the sump can be safely pumped to the vented container in the auxiliary building. Based on our evaluation, the existing lube oil collection system for reactor coolant pumps provides a level of protection equivalent to the requirements specified in Subsection III.0 of Appendix R. Therefore, the exemption from the requirements specified in Subsection III.0 for the lube oil collection system is granted."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-52 - Revision 1
NFPA 805 Section #
3.3.12(2)
Plant Documentation
Subsection Title
Reactor Coolant
Pumps
Requirement/Guidance
For facilities with non-
inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system shall be designed and installed such that leakage from the oil systems is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.
(2) Leakage shall be collected and drained to a vented closed container that can hold the inventory of the reactor coolant pump lubricating oil system.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix E, "The Reactor Coolant Pump Lube Oil Collection System is piped to sump “A” inside containment. The contents of the sump can be pumped to closed vented tank(s) inside the Auxiliary Building via two (2) alternative flow paths…The sump in the basement of containment is a concrete pit having a capacity of 990 gallons. This is more than the capacity needed to contain the total inventory of lube oil (410 gallons) for the two (2) reactor coolant pumps for each unit…The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the Control Room if this level is exceeded. The operator can initiate manual control of the sump pump(s) at any time by overriding the automatic control of sump level."
Per Page 2 of Attachment 1 of Letter from Musolf (NSP) to Director (NRC) dated 4/5/84, "In support of the system as is currently installed it should be noted…the lube oil collection system is seismically designed." Per Page 2 of Attachment 1 of Letter from NSP to NRC dated 4/5/84, "In summary, Northern States Power has made an
Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84 Letter from Musolf (NSP) to Director (NRC) dated 4/5/84 Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," Rev. 14, dated 4/29/10
extensive effort to comply with the requirements of Appendix R. In comparing the lube oil collection system to the requirements of Section III.0 [of Appendix R], concerns were voiced over the use of a closed vented container inside containment because of the need for it to also act as a collection point for seal leakage. Northern States Power believes that the existing configuration meets the intent of Appendix R in that all lube oil
Identifier
None
Items for Implementation None
EEEE Description Summary is collected to a common point which will prevent its contact with hot piping in the area and is isolated from electrical power cable which might cause ignition."
Per Page 5 of Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84, "the pipe from the sump to the vented container in the auxiliary building has been designed to seismic category Class III which meets the requirement of Regulatory Guide 1.29, paragraph C-2. If failure of this pipe were to occur during a seismic event, the functions of plant features described in paragraph 1 (a through q) of Regulatory Guide 1.29 will not be affected and the plant can be brought to cold shutdown."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-53 - Revision 1
Per Page 6 of Letter from NRC to NSP dated 7/31/84, "We agree with the licensee that, although lube oil leakage is collected in the sump before it is pumped to a vented container, the sump design at this plant assures us that oil collected there will not lead to fire during normal or design basis accident conditions. The capacity of the sump and the vented containers is adequate to safely contain any anticipated lube oil leakage and the controls provide reasonable assurance that any lube oil collected in the sump can be safely pumped to the vented container in the auxiliary building. Based on our evaluation, the existing lube oil collection system for reactor coolant pumps provides a level of protection equivalent to the requirements specified in Subsection III.0 of Appendix R. Therefore, the exemption from the requirements specified in Subsection III.0 for the lube oil collection system is granted."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-54 - Revision 1
NFPA 805 Section #
3.3.12(3)
Plant Documentation
Subsection Title
Reactor Coolant
Pumps
Requirement/Guidance
For facilities with non-
inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.
(3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazard of a fire flashback.
Compliance Statement
Complies
Compliance Basis
Per Procedure D18, Equipment Lubrication, the oil in the Reactor Coolant Pumps is Mobil SHC 824. This oil has a flash point of 248°C (478°F), therefore, a flame arrestor is not needed for this system.
Procedure D18, "Equipment Lubrication", Rev. 84, dated 12/21/11 Mobil Oil specification sheet, Mobil SHC 824 (http://www.mobil.com/USA- English/Lubes/PDS/GLXXENINDMOMobil_SHC_800.aspx)
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-55 - Revision 1
NFPA 805 Section #
3.3.12(4)
Plant Documentation
Subsection Title
Reactor Coolant
Pumps
Requirement/Guidance
For facilities with non-
inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal conditions such as accident conditions or earthquakes. All of the following shall apply.
(4) Leakage points on a reactor coolant pump motor to be protected shall include but not be limited to the lift pump and piping, overflow lines, oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and the oil reservoirs, where such features exist on the reactor coolant pumps.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix E, "The Reactor Coolant Pump Lube Oil Collection System is piped to sump “A” inside containment. The contents of the sump can be pumped to closed vented tank(s) inside the Auxiliary Building via two (2) alternative flow paths…The sump in the basement of containment is a concrete pit having a capacity of 990 gallons. This is more than the capacity needed to contain the total inventory of lube oil (410 gallons) for the two (2) reactor coolant pumps for each unit…The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the Control Room if this level is exceeded. The operator can initiate manual control of the sump pump(s) at any time by overriding the automatic control of sump level."
Per Page 2 of Attachment 1 of Letter from Musolf (NSP) to Director (NRC) dated 4/5/84, "In support of the system as is currently installed it should be noted…the lube oil collection system is seismically designed." Per Page 2 of Attachment 1 of Letter from NSP to NRC dated 4/5/84, "In summary, Northern States Power has made an extensive effort to comply with the requirements of Appendix R. In comparing the lube oil collection system to the requirements of Section III.0 [of Appendix R], concerns were voiced over the use of a closed vented container inside containment because of the need for it to also act as a collection point for seal leakage. Northern States Power believes that the existing configuration
Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84 Letter from Musolf (NSP) to Director (NRC) dated 4/5/84 Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," Rev. 14, dated 4/29/10
meets the intent of Appendix R in that all lube oil is collected to a common point which will prevent its contact with hot piping in the area and is isolated from electrical power cable which might cause ignition."
Per Page 5 of Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84, "the pipe from the sump to the vented container in the auxiliary building has
Identifier
None
EEEE Description Summary been designed to seismic category Class III which meets the requirement of Regulatory Guide 1.29,
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-56 - Revision 1
Items for Implementation None
paragraph C-2. If failure of this pipe were to occur during a seismic event, the functions of plant features described in paragraph 1 (a through q) of Regulatory Guide 1.29 will not be affected and the plant can be brought to cold shutdown." Per Page 6 of Letter from NRC to NSP dated 7/31/84, "We agree with the licensee that, although lube oil leakage is collected in the sump before it is pumped to a vented container, the sump design at this plant assures us that oil collected there will not lead to fire during normal or design basis accident conditions. The capacity of the sump and the vented containers is adequate to safely contain any anticipated lube oil leakage and the controls provide reasonable assurance that any lube oil collected in the sump can be safely pumped to the vented container in the auxiliary building. Based on our evaluation, the existing lube oil collection system for reactor coolant pumps provides a level of protection equivalent to the requirements specified in Subsection III.0 of Appendix R. Therefore, the exemption from the requirements specified in Subsection III.0 for the lube oil collection system is granted."
The RCP Lube Oil Collection System was designed and installed to consider potential leakage points and this design was specifically approved in the NRC SER.
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-57 - Revision 1
NFPA 805 Section #
3.3.12(5)
Plant Documentation
Subsection Title
Reactor Coolant
Pumps
Requirement/Guidance
For facilities with non-
inerted containments, reactor coolant pumps with an external lubrication system shall be provided with an oil collection system. The oil collection system shall be designed and installed such that leakage from the oil system is safely contained for off normal condtions such as accident conditions or earthquakes. All of the following shall apply.
(5) The collection basin drain line to the collection tank shall be large enough to accommodate the largest potential oil leak such that oil leakage does not overflow the basin.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix E, "The Reactor Coolant Pump Lube Oil Collection System is piped to sump “A” inside containment. The contents of the sump can be pumped to closed vented tank(s) inside the Auxiliary Building via two (2) alternative flow paths…The sump in the basement of containment is a concrete pit having a capacity of 990 gallons. This is more than the capacity needed to contain the total inventory of lube oil (410 gallons) for the two (2) reactor coolant pumps for each unit…The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the Control Room if this level is exceeded. The operator can initiate manual control of the sump pump(s) at any time by overriding the automatic control of sump level."
Per Page 2 of Attachment 1 of Letter from Musolf (NSP) to Director (NRC) dated 4/5/84, "In support of the system as is currently installed it should be noted…the lube oil collection system is seismically designed." Per Page 2 of Attachment 1 of Letter from NSP to NRC dated 4/5/84, "In summary, Northern States Power has made an extensive effort to comply with the requirements of Appendix R. In comparing the lube oil collection system to the requirements of Section III.0 [of
Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84 Letter from Musolf (NSP) to Director (NRC) dated 4/5/84 Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," Rev. 14, dated 4/29/10
Appendix R], concerns were voiced over the use of a closed vented container inside containment because of the need for it to also act as a collection point for seal leakage. Northern States Power believes that the existing configuration meets the intent of Appendix R in that all lube oil is collected to a common point which will prevent its contact with hot piping in the area and is isolated from electrical power cable which might
Identifier
None
Items for Implementation None
EEEE Description Summary cause ignition."
Per Page 5 of Letter from Miller (NRC) to Musolf (NSP) dated 7/31/84, "the pipe from the sump to the vented container in the auxiliary building has been designed to seismic category Class III which meets the requirement of Regulatory Guide 1.29, paragraph C-2. If failure of this pipe were to occur during a seismic event, the functions of plant features described in paragraph 1 (a through q) of Regulatory Guide 1.29 will not be affected and the plant can be brought to cold shutdown." Per Page
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-58 - Revision 1
6 of Letter from NRC to NSP dated 7/31/84, "We agree with the licensee that, although lube oil leakage is collected in the sump before it is pumped to a vented container, the sump design at this plant assures us that oil collected there will not lead to fire during normal or design basis accident conditions. The capacity of the sump and the vented containers is adequate to safely contain any anticipated lube oil leakage and the controls provide reasonable assurance that any lube oil collected in the sump can be safely pumped to the vented container in the auxiliary building. Based on our evaluation, the existing lube oil collection system for reactor coolant pumps provides a level of protection equivalent to the requirements specified in Subsection III.0 of Appendix R. Therefore, the exemption from the requirements specified in Subsection III.0 for the lube oil collection system is granted."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-60 - Revision 1
NFPA 805 Section #
3.4.1
Plant Documentation
Subsection Title
On-Site Fire-
Fighting Capability Subsection 3.4.1(a)
Requirement/Guidance
(a) A fully staffed,
trained, and equipped fire-fighting force shall be available at all times to control and extinguish all fires on site. This force shall have a minimum complement of five persons on duty and shall conform with the following NFPA standards as applicable:
(1) NFPA 600, Standard on Industrial Fire Brigades (interior structural fire fighting) (2) NFPA 1500, Standard on Fire Department Occupational Safety and Heath Program (3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for Fire Department Physicians.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 7.16 of Procedure 5AWI 3.13.0, "Fire Protection Program," "A plant Fire Brigade shall be established to provide initial responses to fires identified in or threatening main plant structures. A Fire Brigade of five persons shall be on-site at all times in addition to the minimum shift crew complement needed to safely shut down the unit(s)."
Per Technical Specification 6.1-1 (Amendment 39, Unit 1; Amendment 33, Unit 2), as identified in Section 6.1.C.6 of letter from Schwencer (NRC) to Mayer (NSP) dated 9/6/79, "A fire brigade of at least five members shall be maintained on site at all times. Fire brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of fire brigade members provided immediate action is taken to restore the fire brigade to within the minimum requirements." (This Technical Specification requirement has since been relocated to Procedure 5AWI 3.13.0 as part of the removal of fire protection requirements from the Technical Specifications.)
The fire brigade has been reviewed against the requirements of NFPA 600, as detailed in Fire Protection Engineering Evaluation (FPEE) 11-031.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-62 - Revision 1
NFPA 805 Section #
3.4.1
Plant Documentation
Subsection Title
On-Site Fire-
Fighting Capability Subsection 3.4.1(b)
Requirement/Guidance
(b) Industrial fire brigade
members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required.
Compliance Statement
Complies
Compliance Basis
Per Section 7.16 of Procedure 5AWI 3.13.0, "Fire Protection Program," fire brigade members are not required to perform safe shutdown activities. The section states, "A Fire Brigade of five persons shall be on-site at all times in addition to the minimum shift crew complement needed to safely shut down the unit(s)."
5.13, "Fire Protection Program Standard," "Industrial Fire Brigade members shall have no other assigned normal plant duties that would prevent immediate response to a fire or other emergency as required."
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-63 - Revision 1
NFPA 805 Section #
3.4.1
Plant Documentation
Subsection Title
On-Site Fire-
Fighting Capability Subsection 3.4.1 (c)
Requirement/Guidance
(c) During every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria.
Exception to (c): Sufficient training and knowledge shall be permitted to be provided by an operations advisor dedicated to industrial fire brigade support.
Compliance Statement
Complies
Compliance Basis
Per Section 5.5.1(c)(3) of Fleet Procedure CD 5.13, "Fire Protection Program Standard," "During every shift, the Fire Brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on those systems."
Per Section 7.16 of Procedure 5AWI 3.13.0, "Fire Protection Program," "The Fire Brigade leader shall be competent to assess the potential safety consequences of a fire and advise Control Room personnel. Such competence by the Fire Brigade leader may be evidenced by possession of an Operator’s license or equivalent knowledge of plant safety related systems."
Sections 4.8.3(A) and 4.8.3(B) of Procedure F5, "Fire Fighting," details the training requirements for all fire brigade members, which includes the identification and location of fire hazards
throughout the plant, as well as the various methods for fighting fires.
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-64 - Revision 1
NFPA 805 Section #
3.4.1
Plant Documentation
Subsection Title
On-Site Fire-
Fighting Capability Subsection 3.4.1(d)
Requirement/Guidance
(d) The industrial fire
brigade shall be notified immediately upon verification of a fire.
Compliance Statement
Complies
Compliance Basis
Per Section 4.6 of Procedure F5, "Fire Fighting," the fire brigade members are notified of a fire incident via a "Fire Alarm page" which is sent by the Control Room Operator. Per Section 1.2.6 of Procedure F5, the Fire Brigade Auto Paging
system automatically pages all required fire brigade responders and support personnel for fire scene command and control. The auto paging system would be actuated immediately from the Unit 1 Lead, Unit 2 Lead, Unit 1 Shift Supervisor, and/or Unit 2 Shift Supervisor telephones upon
Identifier
None
Items for Implementation None
EEEE Description Summary verification of a fire in the plant.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-65 - Revision 1
NFPA 805 Section #
3.4.1
Plant Documentation
Subsection Title
On-Site Fire-
Fighting Capability Subsection 3.4.1(e)
Requirement/Guidance
(e) Each industrial fire
brigade member shall pass an annual physical examination to determine that he or she can perform the strenuous activity required during manual firefighting operations. The physical examination shall determine the ability of each member to use respiratory protection equipment.
Compliance Statement
Complies
Compliance Basis
Per Section 4.8.1 of Procedure F5, "Fire Fighting," "All new members of the Fire Brigades shall have an initial physical examination for strenuous physical activity as experienced in fire fighting...Follow-up physical examinations shall be conducted annually for all Fire Brigade members (i.e., every 9 to 15 months, provided that three consecutive exams do not exceed 39 months)...Initial and follow-up physical examinations shall include respiratory protection qualification testing which screens all respirator users (including Fire Brigade members for cardiopulmonary deficiencies)...Physical examinations shall be conducted by a physician."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-66 - Revision 1
NFPA 805 Section #
3.4.2
Plant Documentation
Subsection Title
Pre-Fire Plans
Requirement/Guidance
Current and detailed pre-
fire plans shall be available to the industrial fire brigade for all areas in which a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.
Compliance Statement
Complies
Compliance Basis
Procedure F5 Appendix A, "Fire Strategies," details each area's layout, hazards, fire protection features, communications capability, and equipment control in its pre-fire plan.
Procedure F5 Appendix A, "Fire Strategies," Rev. 27, dated 11/17/11
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-67 - Revision 1
NFPA 805 Section #
3.4.2.1
Plant Documentation
Subsection Title
Pre-Fire Plans
Requirement/Guidance
The plans shall detail the fire area configuration and fire hazards to be encountered in the fire area, along with any nuclear safety components and fire protection systems and features that are present.
Compliance Statement
Complies
Compliance Basis
Procedure F5 Appendix A, "Fire Strategies," details each area's layout, hazards, fire protection features, communications capability, and equipment control in its pre-fire plan.
Procedure F5 Appendix A, "Fire Strategies," Rev. 27, dated 11/17/11
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-68 - Revision 1
NFPA 805 Section #
3.4.2.2
Plant Documentation
Subsection Title
Pre-Fire Plans
Requirement/Guidance
Pre-fire plans shall be
reviewed and updated as necessary.
Compliance Statement
Complies
Compliance Basis
Section 1.0 of Procedure 5AWI 3.13.1, "Fire Protection Review of Modifications," states, "This Instruction establishes the process for reviewing modifications (engineering changes, equivalency
changes, design changes, temporary modifications, etc.) at PINGP. The review ensures that the fire protection requirements are included. It also ensures adequate evaluation and documentation of the type and quantity of combustible loading in each fire area."
Section 1.1 of Fleet Procedure FP-G-DOC-04, Identifier
None
Items for Implementation None
EEEE Description Summary "Procedure Processing," states "The purpose of this procedure is to establish a common process for initiation, revision, review, and approval of the following document types (listed according to Document Hierarchy): Corporate Directives; Fleet Program/Process Descriptions, Codes of Conduct; Fleet Procedures; Centralized Department Procedures, Corporate Office Procedures; Site Procedures; Forms (that are independent of procedures)" The Pre-Fire Plans are maintained in Procedure F5 Appendix A, "Fire Strategies." This document is controlled, reviewed, and revised through Procedure FP-G-DOC-04 or as required by Procedure 5AWI 3.13.1.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-69 - Revision 1
NFPA 805 Section #
3.4.2.3
Plant Documentation
Subsection Title
Pre-Fire Plans Requirement/Guidance
Pre-fire plans shall be
available in the control room and made available to the plant industrial fire brigade.
Compliance Statement
Complies
Compliance Basis
Per Section 5.20 of Fleet Procedure FP-G-CD-01, "Controlled Documents," distribution of controlled documents is performed using the PASSPORT computer module. Procedure F5 Appendix A, "Fire Strategies," is distributed to the Control Room
Per Section 4.3.3 of F5 Appendix J, “PINGP 1676 SHALL be used as the drill report and as an objective evaluation checklist for all fire drills that are conducted.”
PINGP 1676, “Fire Drill Critique Report” Rev 1, states “Was a copy of Fire Strategies information available and utilized?”
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-70 - Revision 1
NFPA 805 Section #
3.4.2.4
Plant Documentation
Subsection Title
Pre-Fire Plans
Requirement/Guidance
Pre-fire plans shall
address coordination with other plant groups during fire emergencies.
Compliance Statement
Complies
Compliance Basis
Fire brigade coordination with other plant groups during fire emergencies is not addressed in individual pre-fire plans. However, Procedure F5, "Fire Fighting," the parent document to Procedure F5 Appendix A, "Fire Strategies," addresses
coordination with non-fire brigade groups (e.g., security, operations) during fire events. Per Section 1.0, "The purpose of this section is to provide specific instructions on the organization of fire brigades, individual responsibilities in regard to fires, and procedures for extinguishing fires."
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-71 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(a), "Plant Industrial Fire Brigade Training," Section (1)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply:
(1) Plant industrial fire brigade members shall receive training consistent with the requirements contained in NFPA 600, Standard on Industrial Fire Brigades, or NFPA 1500, Standard on Fire Department Occupational Safety and Health Program, as appropriate.
Compliance Statement
Complies
Compliance Basis
The fire brigade has been reviewed against the requirements of NFPA 600, as detailed in Fire Protection Engineering Evaluation (FPEE) 11-031. (NFPA 1500 does not apply as the plant operates a fire brigade, not a fire department.)
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-73 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(a), "Plant Industrial Fire Brigade Training," Section (2)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply. (2) Industrial fire brigade members shall be given quarterly training and practice in fire fighting, including radioactivity and health physics considerations, to ensure that each member is thoroughly familiar with the steps to be taken in the event of a fire.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.8.3 of Procedure F5, "Fire Fighting," the fire brigade training program includes instruction on general plant access, fire fighting tactics and hazards, use of fire fighting equipment, and radiation and contamination considerations.
Per Section 1.2.1 of Procedure F5 Appendix J, "Fire Drills," "Fire drills shall be scheduled so that each Fire Brigade participates in at least four (4) fire drills per year. Drills should be performed at regular intervals with one (1) drill per calendar quarter for each Fire Brigade."
NSP letter dated 5/2/1979, Section 4.1 states that "Drills will be scheduled so that each fire brigade member will participate in at least two drills per year."
NRC SER dated 9/6/1979 references the 5/2/1979 NSP letter and concludes per Section 6.2 that "the fire brigade training program conforms to the provisions of Appendix A to BTP 9.5-1 and is, therefore, acceptable."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-74 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(a), "Plant Industrial Fire Brigade Training," Section (3)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply. (3) A written program shall detail the industrial fire brigade training program.
Compliance Statement
Complies
Compliance Basis
Section 4.8.3 of Procedure F5, "Fire Fighting," details the industrial fire brigade training program.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-75 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(a), "Plant Industrial Fire Brigade Training," Section (4)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply. (4) Written records that include but are not limited to initial industrial fire brigade classroom and hands-on training, refresher training, special training schools attended, drill attendance records, and leadership training for industrial fire brigades shall be maintained for each industrial fire brigade member.
Compliance Statement
Complies
Compliance Basis
Per Section 8.8 of Procedure 5AWI 3.13.0, "Fire Protection Program," "Classroom training sessions, practice sessions, and drills for the brigade shall be documented…"
Per Section 17.5, individual fire brigade training records, drills, practices, and critiques are maintained in accordance with Fleet Procedure FP-G-RM-01, Records Management.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-76 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(b), "Training for Non-Industrial Fire Brigade Personnel"
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(b) Training for Non- Industrial Fire Brigade Personnel. Plant personnel who respond with the industrial fire brigade shall be trained as to their responsibilities, potential hazards to be encountered, and interfacing with the industrial fire brigade.
Compliance Statement
Complies
Compliance Basis
Per Section 8.11 of Procedure 5AWI 3.13.0, "Fire Protection Program," "Level I (A) fire protection training shall be general training given to operations and mechanical maintenance personnel. The training shall include:...Basic principles of fire chemistry and physics...Fire hazards...Fire detection systems...Types of extinguishing systems...Special fire hazards associated with nuclear power…[and] Emergency planning with emphasis on fire emergency"
Section 2.7 of Procedure F5, "Fire Fighting," provides a description of the responsibilities of non-operations personnel, such as radiation protection, security, and nuclear plant service personnel, when responding to fires with the fire brigade.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-77 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(c), "Drills," Section (1)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(c) Drills. All of the following requirements shall apply. (1) Drills shall be conducted quarterly for each shift to test the response capability of the industrial fire brigade.
Compliance Statement
Complies
Compliance Basis
Per Section 1.2.1 of Procedure F5 Appendix J, "Fire Drills," "Fire drills shall be scheduled so that each Fire Brigade participates in at least four (4) fire drills per year...Drills should be performed at regular intervals with one (1) drill per calendar quarter for each Fire Brigade."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-78 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(c), "Drills," Section (2)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(c) Drills. All of the following requirements shall apply.
(2) Industrial fire brigade drills shall be developed to test and challenge industrial fire brigade response, including brigade performance as a team, proper use of equipment, effective use of pre-fire plans, and coordination with other groups. These drills shall evaluate the industrial fire brigade's abilities to react, respond, and demonstrate proper fire- fighting techniques to control and extinguish the fire and smoke conditions being simulated by the drill scenario.
Compliance Statement
Complies
Compliance Basis
Per Section 8.8 of Procedure 5AWI 3.13.0, "Fire Protection Program," fire drills are used to determine the fire brigade's ability to "operate as a team."
Per Section 1.2.3 of Procedure F5 Appendix J, "Fire Drills," "To the extent practical, Fire Brigade members shall use protective equipment, suppression systems, and other equipment used to fight an actual fire during the drills."
Per Section 4.2.1, "The drill should be conducted with as much realism as possible. Fire Brigade members should respond as expeditiously as they would in the event of an actual fire."
Per Section 4.2.2, "The scenario used for the drill shall be included in the fire drill report."
Per Section 4.2.3, "PINGP 1676 shall be used as the drill objectives and as an objective evaluation checklist for all fire drills that are conducted."
Per Section 4.2.5, "A critique shall be conducted as soon as possible after the fire drill is terminated. All items discussed during the critique shall be included in the fire drill report."
PINGP 1676, "Fire Drill Critique Report" specifically includes an evaluation/critique of brigade performance as a team, proper use of equipment, effective use of pre-fire plans, and coordination with other groups.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-79 - Revision 1
Items for Implementation None
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-80 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(c), "Drills," Section (3)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(c) Drills. All of the following requirements shall apply.
(3) Industrial fire brigade drills shall be conducted in various plant areas, especially in those areas identified to be essential to plant operation and to contain significant fire hazards.
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Procedure F5 Appendix J, "Fire Drills” will be revised to include that fire drills will be conducted in various plant areas.
As described in Table S-3, item #3, Procedure F5 Appendix J, "Fire Drills," will be revised to require that fire brigade drills be conducted in various plant areas.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-81 - Revision 1
NFPA 805 Section #
3.4.3
Plant Documentation
Subsection Title
Training and Drills
Subsection 3.4.3(c), "Drills," Section (4)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
(c) Drills. All of the following requirements shall apply.
(4) Drill records shall be maintained detailing the drill scenario, industrial fire brigade member response, and ability of the industrial fire brigade to perform as a team.
Compliance Statement
Complies
Compliance Basis
Details of the drill are recorded per PINGP 1676, "Fire Drill Critique Report".
Per Section 5.1 of Procedure F5 Appendix J, "Fire Drills," "Documentation of drills and practices shall be maintained in accordance with FP-G-RM-01, Records Management, for two (2) years."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-82 - Revision 1
NFPA 805 Section #
3.4.3
Subsection Title
Training and Drills
Subsection 3.4.3(c), "Drills," Section (5)
Requirement/Guidance
Industrial fire brigade
members and other plant personnel who would respond to a fire in conjunction with the brigade shall be provided with training commensurate with their emergency responsibilities.
Compliance Statement
Complies
Compliance Basis
Per Section 4.2.5 of Procedure F5 Appendix J, "Fire Drills," "A critique shall be conducted as soon as possible after the fire drill is terminated. All items discussed during the critique shall be included in the fire drill report."
PINGP 1676, "Fire Drill Critique Report", includes the fire drill scenario and fire drill critique.
Plant Documentation
PINGP 1676, "Fire Drill Critique Report"
(c) Drills. All of the following requirements shall apply. (5) A critique shall be held and documented after each drill.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-83 - Revision 1
NFPA 805 Section #
3.4.4
Plant Documentation
Subsection Title
Fire-Fighting
Equipment
Requirement/Guidance
Protective clothing,
respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment shall be provided for the industrial fire brigade. This equipment shall conform with the applicable NFPA standards.
Compliance Statement
Complies
Compliance Basis
Table 1 of Procedure F5, "Fire Fighting," establishes a list of personal protective clothing and equipment which are available for the fire brigade at several locations throughout the plant. The equipment includes SCBA, turnout gear, fire hose and nozzles, smoke ejectors, emergency lighting, communications equipment, spanner wrenches, and other such tools.
Per Section 7.2.2, in the event of a radioactive fire incident, fire brigade members are required to "Obtain a dosimeter."
Per Section 6.1 of Procedure RPIP 1101, "TLD Issue," a TLD is issued "to any worker entering the Radiologically Controlled Area (RCA)…"
Per Section 7.5.9 of Procedure 5AWI 3.13.0, "Fire Protection Program," the fire protection coordinator is responsible for "procuring, inspecting, and maintaining fire brigade equipment in accordance with applicable NFPA requirements."
FP-RP-DP-01, "Dosimetry Program", Section 3.5. lists the responsibilities of personnel issued a
Identifier
None
Items for Implementation None
EEEE Description Summary TLD. PINGP General Employee Training instructs employees to wear their TLDs at all times when on site.
Per Section 1.2.5 of procedure F5, “Fire Fighting,” All SCBA devices SHALL be approved by NIOSH and MSHA with a minimum duration of 30 minutes.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-85 - Revision 1
NFPA 805 Section #
3.4.5.1
Plant Documentation
Subsection Title
Off-Site Fire
Department Interface Mutual Aid Agreement
Requirement/Guidance
Off-site fire authorities
shall be offered a plan for their interface during fires and related emergencies on site.
Compliance Statement
Complies
Compliance Basis
Per Section 6.1 of Procedure F5, "Fire Fighting," "The primary off site fire department (Red Wing Fire Department) shall provide emergency assistance and shall be called immediately on report of fire."
Per Section 6.2, "Upon arrival, the RWFD shall have primary responsibility for extinguishing the fire...The RWFD may call other mutual aid Departments in the local area for assistance, if necessary."
Per Section 5.6.4.A of Procedure E-PLAN,
Identifier
None
Items for Implementation None
EEEE Description Summary "Emergency Plan," "The Red Wing Fire Department will provide assistance in the event of a fire occurring at the plant."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-86 - Revision 1
NFPA 805 Section #
3.4.5.2
Plant Documentation
Subsection Title
Off-Site Fire
Department Interface Site-Specific Training
Requirement/Guidance
Fire fighters from the off-
site fire authorities who are expected to respond to a fire at the plant shall be offered site-specific training and shall be invited to participate in a drill at least annually.
Compliance Statement
Complies
Compliance Basis
Per Section 6.4.1 of Procedure 5AWI 3.11.0, "Site Training and Staff Selection," "The Emergency Preparedness (EP) Manager maintains and implements an emergency preparedness training plan for...local fire departments...that provide primary support services to the site and its personnel."
"Fire Drills," "A drill involving the local on-duty fire department shall be done at least once per year. Credit may be taken for the annual drill if the local fire department responds to a site fire emergency or participates in training at PINGP or with the fire brigade."
Per Section 7.5.6 of Procedure 5AWI 3.13.0, "Fire
Identifier
None
Items for Implementation None
EEEE Description Summary Protection Program," The site Fire Protection Coordinator establishes "coordination with the local fire department, including joint drill and training sessions to familiarize fire department personnel with plant access routes, layout, equipment and special hazards.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-87 - Revision 1
NFPA 805 Section #
3.4.5.3
Plant Documentation
Subsection Title
Off-Site Fire
Department Interface Security and Radiation Protection
Requirement/Guidance
Plant security and
radiation protection plans shall address off- site fire authority response.
Compliance Statement
Complies
Compliance Basis
Per Section 2.7 of Procedure F5, "Fire Fighting," "Security is responsible to escort the off-site Fire Department personnel and vehicle on to plant property, through the security gate to the building door selected by brigade chief for entry."
Per Section 6.2, "Upon arrival, the RWFD Shall have primary responsibility for extinguishing the fire...Anti-C clothing should be provided by radiation protection personnel, if required.
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-88 - Revision 1
NFPA 805 Section #
3.4.6
Plant Documentation
Subsection Title
Communications
Requirement/Guidance
An effective emergency
communications capability shall be provided for the industrial fire brigade.
Compliance Statement
Complies
Compliance Basis
Per Section 10.3.8 of the USAR, "A fixed public address system interfaced with a UPS powered Private Branch Exchange (PBX) telephone system provide normal and emergency communications. In the event of a PBX failure, power fail telephone stations from the local telephone office and
extensions operating on the Xcel Energy Sherco Plant Telephone Switch could be utilized to conduct emergency communications. In addition, a sound powered communications system is installed with communications jacks located throughout the plant. The sound powered system requires no external power, and headsets for use
Identifier
None
Items for Implementation None
EEEE Description Summary with the system are readily available. ...The site radio system utilizes hand-held portable radios, mobile radios, and stationary radio consoles to facilitate two way communications between out-plant personnel and control points such as the Control Room, Central Alarm Station, or Technical Support Center. The radio transmitters and the radio system controller are powered by the interruptible a-c system with backup transmitters located outside the plant in the Guardhouse and Microwave building."
Per Sections 7.2.1 and 7.2.2 of Procedure E- PLAN, "Emergency Plan," "All emergency operating facilities have at least two means of communications: (1) portable or installed radio systems; and (2) normal telephone communications. ...The normal onsite communications during an emergency will be made via the plant telephone system with a public address system option. The telephone system is powered by non-interruptible power. The public address system includes about 175 loudspeakers located throughout the entire plant area.
A separate paging system has 20 handsets located at strategic plant areas.
...Designated members of the site’s emergency organization carry personal pagers which can be activated from the Technical Support Center. A special emergency code is displayed on the pager."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-90 - Revision 1
NFPA 805 Section #
3.5.1
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Water Supply
Requirement/Guidance
A fire protection water
supply of adequate reliability, quantity, and duration shall be provided by one of the two following methods.
(a) Provide a fire protection water supply of not less than two separate 300,000-gal (1,135,500-L) supplies.
(b) Calculate the fire flow rate for 2 hours. This fire flow rate shall be based on 500 gpm (1892.5 L/min) for manual hose streams plus the largest design demand of any sprinkler or fixed water spray system(s) in the power block as determined in accordance with NFPA 13, Standard for the Installation of the Sprinkler Systems, or NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection. The fire water supply shall be capable of delivering this design demand with the hydraulically least demanding portion of fire main loop out of service.
Compliance Statement
Complies via Previous
Approval; Complies with
Item for Implementation
Compliance Basis The Mississippi River is capable of supplying a virtually unlimited amount of water. Per Section 4.3.1.1 of NRC SER dated 9/6/79, "Water for fire protection is from the Mississippi River through the intake canal. The water is drawn by fire pumps located in the screen house. A secondary water supply is available through the cooling water system emergency intake pipe via crossovers to the fire mains. "We find the water supply meets the objectives identified in Section 2.2 of this report and is, therefore, acceptable." Per Section 4.1 of Plant Procedure F5 Appendix F, "Fire Hazard Analysis," "Water for fire protection system is from the Mississippi River through the intake canal. The water is drawn by two horizontal shaft centrifugal fire pumps each with a design capacity of 2,000 gpm at a pressure 125 psi." Per Section 7.3 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The fire protection water system is supplied from the Mississippi River by two horizontal centrifugal fire pumps rated at 2,000 gpm at 125 psig." As stated in Section 3.5.3, of this attachment, the largest design demand of any sprinkler or fixed water spray system determined in accordance with NFPA 13 and NFPA 15 is a transformer system with a demand of 1534 gpm (including a hose allowance of 500 gpm). Therefore, water quantity required for 2 hour supply is 184,080 gallons. The Mississippi River is capable of supplying this water quantity. A calculation to demonstrate that the fire water supply is capable of delivering the largest design demand with the hydraulically least demanding portion of fire main loop out of service is an item for implementation and identified in Attachment S, Table S-3 item #4.
Items for Implementation As described in Table S-3, item #4, a calculation to demonstrate that the fire water supply is capable of delivering the largest design demand with the hydraulically least demanding portion of fire main loop out of service will be performed.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-92 - Revision 1
NFPA 805 Section #
3.5.2
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
The tanks shall be
interconnected such that fire pumps can take suction from either or both. A failure in one tank or its piping shall not allow both tanks to drain. The tanks shall be designed in accordance with NFPA 22, Standard for Water Tanks for Private Fire Protection.
Exception No. 1: Water storage tanks shall not be required when fire pumps are able to take suction from a large body of water (such as a lake), provided each fire pump has its own suction and both suctions and pumps are adequately separated.
Exception No. 2: Cooling tower basins shall be an acceptable water source for fire pumps when the volume is sufficient for both purposes and water quality is consistent with the demands of the fire service.
Compliance Statement
Complies
Compliance Basis
Complies with Exception No. 1.
Per Section 4.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," “Water for fire protection system is from the Mississippi River through the intake canal. …"The electric motor-driven fire pump and the diesel engine-driven fire pump are located at the east side of the screenhouse separated by a distance of approximately 20 feet. A three-hour rated fire barrier is provided around the electric motor-driven fire pump."
Therefore, PINGP is in compliance with Exception No. 1 to this requirement.
Per Section 7.3 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The fire protection water system is supplied from the Mississippi River…"
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-93 - Revision 1
Items for Implementation None
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-94 - Revision 1
NFPA 805 Section #
3.5.3
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Fire pumps, designed
and installed in accordance with NFPA 20, Standard for the Installation of Stationary Pumps for Fire Protection, shall be provided to ensure that 100 percent of the required flow rate and pressure are available assuming failure of the largest pump or pump power source.
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Per Section 4.1.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Water for fire protection system is from the Mississippi River through the intake canal. The water is drawn by two horizontal shaft centrifugal fire pumps each with a design capacity of 2,000 gpm at a pressure 125 psi. One of the fire pumps is diesel engine-driven and the other is electric motor-driven. A third pump, electric motor-driven also having a capacity of 2000 gpm at a pressure of 125 psi normally assigned to the screen wash function can be aligned to pump into the fire water system. On loss of offsite AC power, the diesel driven pump will be available to supply water to the fire protection system."
Code Compliance Review NFPA 20–1969, "NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1969, Code Compliance Review," Rev. 1 dated 6/1/2010 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," dated 9/15/70
Industry-Related References
NFPA 20, "Standard for the Installation of Centrifugal Fire Pumps," 1969 Edition
Per Section 7.3 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," The fire protection water system is supplied from the Mississippi River by two horizontal centrifugal fire pumps rated at 2,000 gpm at 125 psig. One pump is motor driven and the other pump is diesel driven. The 10” fire header is maintained between 110 and 113 psig by a jockey pump. The jockey pump does not supply water to the header. If the water demand is such that the jockey pump cannot maintain the header pressure, the screen
Identifier
AR 1189183 Fire Protection Engineering Evaluation Code Compliance Deviations, Rev. 1, "NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1969, Code Compliance Review," dated 6/1/2010.
EEEE Description
Fire pumps are tested every 18 months, while NFPA 20 requires that they be tested annually. Based on the design of the fire pumps and the fire water system, the 18 month test frequency currently used by PINGP is deemed adequate. Code Compliance Review
Summary
EEEE Bases for Acceptibility "1. System Design: The redundancy provided by the three pumps will more than compensate for the potential failure of one pump to perform as intended. If debris caused one pump to be disabled, two other pumps are available. Since any single pump can satisfy the largest demand, sufficient firewater is expected to be available at all times.
2. Additional Tests: In addition to the 18 month performance tests, the fire pumps are also tested weekly and/or monthly. Performance of these tests would provide indication of potential obstructions or
wash pump will start (if not running). The screen wash pump, which is rated at 2000 gpm at 125 psig, provides water to the fire header when the bypass valve opens at 106 psig. However the screenwash pump has a bypass line which is orificed to restrict flow to 450 gpm to the fire header. The screen wash pump may be directly aligned to the fire header by a manual action from the Control Room in order to provide the rated 2,000 gpm at 125 psig. Due to the required manual action, this pump can not be credited as a primary fire pump. The motor driven fire pump will automatically start at 98 psig. If the header pressure to drops to 93 psig, the diesel-driven fire pump will start. The motor and diesel-driven fire pumps are designed to pump 2,000 gpm and maintain a minimum of 65 psig in the fire header, measured at the highest point in the system. The motor driven fire pump, diesel-driven fire pump, and the manually-aligned screenwash pump can be used to supply all fire fighting water requirements.
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PINGP Page A-95 - Revision 1
degradations in the fire water supply.
3. The 18 months test interval reflects the testing frequency previously accepted by the NRC. This test frequency was established under NSP's original Technical Specifications. Thus, at the time PINGP's Fire Protection Program was established and approved, the NRC considered the current 18 month test interval acceptable. Eight deviations were dispositioned as acceptable.
"In the event that the motor driven pump is impaired, the control room may manually align the screen wash pump directly to the FP header via CV-31055 with unrestricted flow (2,000 GPM).
If all three fire pumps are impaired, or if the fire suppression water system is incapable of supplying water to a safety related area, a backup fire suppression water system must be established within 24 hours."
Per Section 3.7 of Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," "Both fire pumps are of bronze-fitted cast iron construction…and conform to NFPA Standard 20."
Fire pumps have been reviewed against the requirements of NFPA 20, as detailed in the Fire Protection Engineering Evaluation Code Compliance Deviations, Rev. 1, "NFPA 20, Standard for the Installation of Centrifugal Fire Pumps, 1969, Code Compliance Review," dated 6/1/2010.
Items for Implementation None
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NFPA 805 Section #
3.5.4
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
At least one diesel
engine-driven fire pump or two more seismic Category I Class IE electric motor-driven fire pumps connected to redundant Class IE emergency power buses capable of providing 100 percent of the required flow rate and pressure shall be provided.
Compliance Statement
Complies
Compliance Basis
Per Section 4.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "The water is drawn by two horizontal shaft centrifugal fire pumps each with a design capacity of 2,000 gpm at a pressure 125 psi. One of the fire pumps is diesel engine-driven and the other is electric motor-driven. A third pump, electric motor-driven also having a capacity of 2000 gpm at a pressure of 125 psi normally assigned to the screen wash function can be aligned to pump into the fire water system. On loss of offsite AC power, the diesel driven pump will be available to supply water to the fire protection system."
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NFPA 805 Section #
3.5.5
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Each pump and its driver
and controls shall be separated from the remaining fire pumps and from the rest of the plant by rated fire barriers.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "The electric motor-driven fire pump and the diesel engine-driven fire pump are located at the east side of the screenhouse separated by a distance of approximately 20 feet. A three-hour rated fire barrier is provided around the electric motor-driven fire pump. Electric cables for the electric motor-driven fire pump were re-
Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11
routed to assure that a diesel fuel fire will not damage the controls or power source for the electric motor-driven fire pump. The screen wash pump, used as a backup fire pump, is located at the west side of the screenhouse, well separated from the regular fire pumps."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per Section 3.2.6 of Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80, "We concur with the licensee's conclusion that the control cable for the motor driven fire pump does not need to be relocated or protected. Further, the licensee's proposal to relocate or protect the motor driven fire pump power cabling and provide a fire barrier to enclose the motor driven fire pump will assure that a fire would not cause the loss of both fire pumps. Based on our review, we find the licensee's proposed modifications to assure adequate separation of the fire pumps acceptable."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
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NFPA 805 Section #
3.5.6
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Fire pumps shall be
provided with automatic start and manual stop only.
Compliance Statement
Complies
Compliance Basis Per Section 7.3 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The motor driven fire pump will automatically start at 98 psig. If the header pressure to drops to 93 psig, the diesel-driven fire pump will start."
Procedure C31, "Fire Protection & Detection Systems," Rev. 38, dated 5/11/07 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 Fire Protection Engineering Evaluation, NFPA 20 - 1969, Standard for Centrifugal Fire Pumps, Code Compliance Deviations, Rev. 1
Industry-Related References
NFPA 20 - 1969, Standard for Centrifugal Fire Pumps,
Per Sections 5.2 and 5.4 of Procedure C31, "Fire Protection & Detection Systems," both the motor and diesel fire pumps are shut off manually at local pump panels. Fire Protection Engineering Evaluation, NFPA 20 - 1969, Standard for Centrifugal Fire Pumps, Code Compliance Deviations, Rev. 1
Identifier
None
Items for Implementation None
EEEE Description Summary
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NFPA 805 Section #
3.5.7
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Individual fire pump
connections to the yard fire main loop shall be provided and separated with sectionalizing valves between connections.
Compliance Statement
Complies
Compliance Basis
Per Drawing NF-39228-1, fire pump connections to the yard main fire loop are provided with sectionalizing valves between connections. Valves FP-30-1 and FP-30-2 are provided for the motor driven fire pump; valves FP-30-4 and FP-30-5 are provided for the diesel driven fire pump; and valve FP-21-1 is provided for the jockey pump.
Drawing NF-39228-1, "Flow Diagram Fire Protection and Screen Wash System - Unit 1 and 2, Rev. 77, dated 8-09
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NFPA 805 Section #
3.5.8
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
A method of automatic
pressure maintenance of the fire protection water system shall be provided independent of the fire pumps.
Compliance Statement
Complies
Compliance Basis
Per Section 4.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Pressurization of the fire water system is maintained by an electric motor-driven jockey pump.
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NFPA 805 Section #
3.5.9
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Means shall be provided
to immediately notify the control room, or other suitable constantly attended location, of operation of fire pumps.
Compliance Statement
Complies
Compliance Basis
Per Section 4.1.3 (C) of Procedure F5 Appendix F, "Fire Hazard Analysis," " Flows from manual hose stations are not annunciated, but they will cause the fire pump to start, thereby transmitting a "fire pump running" signal to the Control Room."
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NFPA 805 Section #
3.5.10
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Water Supply
Requirement/Guidance
An underground yard fire
main loop, designed and installed in accordance with NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, shall be installed to furnish anticipated water requirements.
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Per Section 4.3.1.3 of NRC SER dated 9/6/79, "Separate 10-inch supply lines to the 10-inch underground yard main encircling the plant are provided from the fire pump header in the screen house. Valving is provided so a single break in the discharge piping will not remove both fire pumps from service. A backup supply of water to the yard fire main loop is provided by eight crossovers between the cooling water system and the fire water system." Per Drawing NF-39256-1, "Yard Fire Protection Piping," the plant is surrounded by a 10-inch underground fire main loop.
Code Compliance Review – NFPA 24, Standard for Outside Protection, 1969 ed., Rev 1, 12/2010
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
Code Compliance Review – NFPA 24, Standard for Outside Protection, 1969 ed., Rev 1, 12/2010
Summary
All code deviations have been justified as acceptable. Reference individual reports for basis descriptions.
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NFPA 805 Section #
3.5.11
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Water Supply
Requirement/Guidance
Means shall be provided
to isolate portions of the yard fire main loop for maintenance or repair without simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stations provided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.3.1.3 of NRC SER dated 9/6/79, "Separate 10-inch supply lines to the 10-inch underground yard main encircling the plant are provided from the fire pump header in the screen house. Valving is provided so a single break in the discharge piping will not remove both fire pumps from service…Sectionalizing valves with post indicators subdivide the loop into a number of sections enabling a single section to be isolated without impairing the entire loop...The isolation valves in the interior fire header in the turbine building and auxiliary building are of the butterfly valve type with chain operators. Because of this arrangement, the necessity of isolating a portion of the underground yard loop to repair a hydrant or leaking main will not result in shutting off the supply of water to systems protecting safety- related equipment or areas...We find that...fire water piping systems satisfy the objectives identified in Section 2.2 of this report and are, therefore, acceptable."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
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NFPA 805 Section #
3.5.12
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Water Supply
Requirement/Guidance
Threads compatible with
those used by local fire departments shall be provided on all hydrants, hose couplings, and standpipe risers.
Compliance Statement
Complies
Compliance Basis
Per Section 4.3.1.3 of NRC SER dated 9/6/79, "Threads on hydrant outlets and hose couplings are compatible with those of fire departments which serve the plant."
Per Section 1.2.1.C of Procedure F5 Appendix J, "Fire Drills," "A drill involving the local on-duty fire department shall be done at least once per year."
Procedure F5 Appendix J, "Fire Drills," Rev. 14, dated 9/13/10 NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1969 Code Compliance Review, Revision 1 FPEE-11-050, Code Compliance Review, NFPA 14-1986, Standard for the Installation of Standpipe and Hose Systems, D5/D6 Building, Revision 1
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
The NFPA 14 code compliance reviews (Section 444 of NFPA 14-1969; Section 4-1.3 of NFPA 14- 1986) confirmed that the threads are compatible by those by the Red Wing Fire Department. Per General Note 5 on NF-39300-1 and NF-39301-1, all hose threads to be National Standard hose threads at both ends.
Identifier
None
Items for Implementation None
EEEE Description Summary
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NFPA 805 Section #
3.5.13
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
Headers fed from each
end shall be permitted inside buildings to supply both sprinkler and standpipe systems, provided steel piping and fittings meeting the requirements for ANSI B31.1, Code for Power Piping, are used for the headers (up to and including the first valve) supplying the sprinkler systems where such headers are part of the seismically anlayzed hose standpipe system. Where provided, such headers shall be considered an extension of the yard main system. Each sprinkler and standpipe system shall be equipped with an outside screw and yoke (OS&Y) gate valve or other approved shutoff valve.
Compliance Statement
Complies
Compliance Basis
The hose standpipe system is not seismically designed, but the system meets the requirements of NFPA 805. See Section 3.6.4 below. As a result, the requirements of ANSI B31.1 are not applicable.
The hose stations and standpipes provided for PINGP are in accordance with the requirements of BTP 9.5-1, Appendix A for plants which received a construction permit before July 1, 1976 which do not require a seismic category I water system.
Provisions to supply water to standpipes and hose stations for manual fire suppression in the event of a safe shutdown earthquake are outlined in EDMG-2, Guideline for Damage Mitigation Strategies (Attachment B - Fire System Management and Attachment L - Establishing Emergency Water Supply).
Per Section 4.3.4 of Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," "The sprinkler system shut-off valves are 150 lb. position indicating, outside screw and yoke gate valves, flanged ends."
Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76 NRC SER dated 9/6/79 Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," dated 9/15/70 PINGP Unit 1 & 2 Project Design Manual, dated 4/1967 EDMG-2, Guideline for Damage Mitigation Strategies (Attachment B - Fire System Management and Attachment L - Establishing Emergency Water Supply).
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
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Items for Implementation None
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NFPA 805 Section #
3.5.14
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
All fire protection water
supply and fire suppression system control valves shall be under a periodic inspection program and shall be supervised by one of the following methods.
(a) Electrical supervision with audible and visual signals in the main control room or other suitable constantly attended location.
(b) Locking valves in their normal position. Keys shall be made available to authorized personnel.
(c) Sealing valves in their normal positions. This option shall be utilized only where valves are located within fenced areas or under the direct control of the owner/operator.
Compliance Statement
Complies
Compliance Basis
Per Section 8.5.11 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "Each valve (manual, power operated or automatic) in the flow path for safety- related areas and areas posing a fire hazard to safety-related areas shall be verified to be in the correct position and secured to prevent inadvertent misalignment every month."
Per Section 1.1 of Procedure SP 1200, "Fire Protection System Supply to Safety Related Areas Valve Check," "The purpose of this surveillance is to verify that fire protection valves supplying safety related areas, and areas posing a fire hazard to safety related areas, are in the correct position and secured."
Per Section 1.5.1, the acceptance criteria for all water supply and fire suppression system control valves is "Valve verified open per 5AWI 3.10.1 Appendix E" and "Block Wire in place."
Per Section 5.1, "Block Wire is required to prevent inadvertent mispositioning of valves."
Appendix E of Procedure 5AWI 3.10.1, "Methods of Performing Verifications," provides instructions for inspecting valves to ensure they are adequately locked.
Procedure 5AWI 3.10.1, "Methods of Performing Verifications," Rev. 16, dated 6/30/10 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 Procedure SP 1200, "Fire Protection System Supply to Safety Related Areas Valve Check," Rev. 32, dated 11/20/08
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
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Items for Implementation None
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NFPA 805 Section #
3.5.15
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Water Supply
Requirement/Guidance
Hydrants shall be
installed approximately every 250 ft (76 m) apart on the yard main system. A hose house equipped with hose and combination nozzle and other auxiliary equipment specified in NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appuretenances, shall be provided at intervals of not more than 1000 ft (305 m) along the yard main system.
Exception: Mobile means of providing hose and associated equipment, such as hose carts or trucks, shall be permitted in lieu of hose houses. Where provided, such mobile equipment shall be equivalent to the equipment supplied by three hose houses.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.3.1.3 of NRC SER dated 9/6/79, "Eight hydrants are provided in the yard spaced from 180 to 300 feet apart. The hydrants are supplied from the underground fire loop although the laterals feeding the hydrant are not provided with isolation valves…Each of the hydrants is provided with a well maintained corrugated metal hose house containing: 100 feet of 1 1/2-inch and 200 feet of 2 1/2-inch single jacket lined hose, one 2 1/2-inch and one 1 1/2-inch fog nozzle and various other items of fire fighting equipment. The inventory of equipment in the hose houses requires upgrading to include additional 1 1/2-inch hose, hose gaskets and a gated wye to allow connecting two 1 1/2-inch hose lines to a 2 1/2- inch hose line." "We find that, subject to implementation of the above listed modifications, fire water piping systems satisfy the objectives identified in Section 2.2 of this report and are, therefore, acceptable."
Per Section 7.14 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "Thirty two (32) hydrant units are located around the plant, warehouses, cooling towers, and in the switch yard. Each house contains 200 feet of 2 1/2" fire hose with a nozzle, 200 feet of 1 1/2" fire hose with a nozzle, adaptors for various size hose, spanner wrenches, a 2 1/2" hose control unit, gated wye, hydrant wrench, and hose gaskets."
Per Drawing NF-39256-1, "Yard Fire Protection
Code Compliance Review NFPA 24–1969, "Code Compliance Review – NFPA 24–1969, Standard for Outside Protection," dated 12/2010 Drawing NF-38201, "Outdoor Hose House Fire Protection," Rev. E, dated 3/1/73 Drawing NF-39256-1, "Yard Fire Protection Piping," Rev. AG, dated 2/7/92 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012
Industry-Related References
NFPA 24, "Standard for Outside Protection," 1969 Edition
Piping," the fire hydrants are located along the 10 inch fire main and are approximately 180 feet to 300 feet apart.
Drawing NF-38201, "Outdoor Hose House Fire Protection," details the materials required to furnish the hose houses, which include spanner wrenches, hoses, hose couplings, fog nozzles, etc.
Hose house equipment has been reviewed against the requirements of NFPA 24, as detailed
Identifier
None
EEEE Description Summary in the Code Compliance Review Code Compliance Review – NFPA 24–1969, Standard
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Items for Implementation None
for Outside Protection," dated 12/2001.
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
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PINGP Page A-111 - Revision 1
NFPA 805 Section #
3.5.16
Plant Documentation
Subsection Title
Water Supply
Requirement/Guidance
The fire protection water
supply system shall be dedicated for fire protection use only.
Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.
Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.
Compliance Statement
Submit for NRC Approval
Compliance Basis
Per Section 1.2 of System Description B31A, "Fire Protection System," "The motor-driven fire pump can be aligned to provide a backup water supply to the Screenhouse screenwash system in the event of a screenwash pump failure."
Per Section 3.1.C, "If the fire pump is required to supplement the screenwash header flow, the pump must be started manually either locally or remotely. Once the pump is operating and no auto start signal exists, the discharge to the screenwash header valve opens automatically and maintains the screenwash header pressure at approximately 90 psig."
Per Drawing NF-39228-1, "Fire Protection and Screen Wash System - Unit 1 and 2," valve FP-30- 10 ties the fire protection water system to the screenwash header.
Since the fire water system can be aligned for screenwash system use and such use does not meet the requirement or allowed exceptions to the requirement, this configuration is included in Attachment L and NRC approval is being requested.
Drawing NF-39228-1, "Fire Protection and Screen Wash System - Unit 1 and 2," Rev. 77, dated 8/09 System Description B31A, "Fire Protection System," Rev. 12, dated 9/16/11 Code Compliance Review, NFPA 24-1969, Standard for Outside Protection Revision 1, December 2010
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NFPA 805 Section #
3.6.1
Plant Documentation
Subsection Title
Standpipe and
Hose Stations
Requirement/Guidance
For all power block
buildings, Class III standpipe and hose systems shall be installed in accordance with NFPA 14, Standard for the Installation of Standpipe, Private Hydrant, and Hose Systems.
Compliance Statement
Complies by Previous
NRC Approval; Complies with Use of Existing Engineering Equivalency Evaluations
Compliance Basis
Per Item 125 of the table attached to Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76, "Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1," "The classification of the fire system is QA Type III. Design of hose stations to QA Type I in safeguards areas is not required in operating plants in accordance with Appendix A to APCSB 9.5-1. The system is built to power piping ANSI B31.1 code. Crossover lines from the Cooling Water system and valves are QA Type I to the
Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76 Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80 NRC SER dated 9/6/79 Code Compliance Review NFPA 14–1969, "Code Compliance Review – NFPA 14–1969, Standard for the Installation of Standpipe and Hose Systems," dated 2/15/2011 Code Compliance Review NFPA 14–1986, FPEE-11-050, "Code Compliance Review – NFPA 14–1986, Standard for the Installation of Standpipe and Hose Systems," dated 3/9/2011 Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," dated 9/15/70 System Description B31A, "Fire Protection System," Rev. 12, dated 9/16/11
Industry-Related References
FAQ 06-0019, "Definition of 'Power Block' and 'Plant'," Rev. 4, dated 9/28/07 NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1969 and 1986 Editions
Per Section 4.3.1.4 of NRC SER dated 9/6/79, "Interior hose stations are provided throughout most areas of the plant connected to the fire water header. Most hose stations consist of a pin lug type hose rack, 300 psi hose valve with drip vent and 50 to 100 feet of 1 1/2-inch unlined linen hose coupled to an all fog type nozzle with a ball shut off feature. The provision of unlined linen hose is considered unsatisfactory in an industrial application since it cannot be practically tested, deteriorates readily when subjected to moisture, and is more subject to failure from abrasion and cuts than hose with synthetic lining and jacketing. …"All interior unlined linen fire hose will be replaced with 100% synthetic material fire hose
Identifier
Code Compliance Review NFPA 14–1986, FPEE-11-050, "Code Compliance Review – NFPA 14–1986, Standard for the Installation of Standpipe and Hose Systems," dated 3/9/2011
NFPA 14, Standard for the Installation of Standpipe and Hose Systems, 1969 Code Compliance Review,
EEEE Description
Document the review of the standpipe and hose station systems for compliance with the applicable requirements cited in National Fire Protection Association (NFPA) 14 -1986
Document the review of the stand pipe and hose systems for compliance with the applicable requirements cited
Summary
Four deviations have been found acceptable based on the justification provided. One deviation has been found to be unacceptable and will require the performance of hydraulic calculations to verify the design bases of the standpipe and hose station system. GAR 01183456-01, Perform Hydraulic Calculations for FP Systems, is tracking resolution of this issue.
Eleven deviations to the code requirements have been identified, nine of which are
(300 psi test pressure) FM or UL labelled, that is suitable for pin rack storage. …"We find that, subject to implementation of the above described modifications, interior fire hose stations satisfy the objectives identified in Section 2.2 of this report and are, therefore, acceptable."
Per Section 3.8.A of System Description B31A, "Fire Protection System," "Each hose station is equipped with 50; 75' or 100' of 1-1/2" E-Z off, lightweight, synthetic jacket lined fire hose with an adjustable spray nozzle."
Per Page 5 of Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80, "By letter dated December 26, 1979, the licensee provided design details regarding a more timely method of providing fire suppression water for the containment standpipes. Originally, fire suppression water would not be available for a period of 6 to 8 hours following the detection of a
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Revision 1 in National Fire Protection Association (NFPA) 14-1969, Standard for the Installation of Stand Pipe and Hose Systems
Items for Implementation None
identified as acceptable and two with action requests (ARs) to address resolutions to the identified issues.
fire in containment. The licensee has proposed a design modification in which the existing containment fire line will be supplied by a four- inch cross-connect from the four-inch supply to the CRDM Shroud Cooling Coils. The cross- connect will have a capacity of approximately 350 gpm and will be available during plant operations. The cross-connect will include two manually operated four inch gate valves located within six feet of the personnel access air-lock. The limiting time in pressurizing a containment fire line will be only that time required for personnel entry.
"The licensee's proposed design will result in a more timely method of charging the standpipes in containment. Based on our review, we find the licensee's method of charging the standpipes within containment acceptable."
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Per Section 3.1 of Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," "All fire protection system components are QA type IIB and are in accordance with the National Fire Protection Association (NFPA) Standards."
Per Section 3.3 "Standpipe and fire hose stations conforming to NFPA 14 are located on the roofs of the turbine room and the auxiliary building and on the fan deck of each cooling tower."
Per Section 3.4, "Standpipe and water fog hose stations conforming to NFPA Standard #14 are located on all floors of the service, auxiliary, and reactor buildings. These stations are so located that all areas are protected by a fog nozzle when attached to one 75 ft. length of fire hose."
Standpipes, hydrants, and hose systems have been reviewed against the requirements of NFPA 14, as detailed in the Code Compliance Review "Code Compliance Review – NFPA 14–1969, Standard for the Installation of Standpipe and Hose Systems," dated 2/15/2011 and "Code Compliance Review – NFPA 14–1986, Standard
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for the Installation of Standpipe and Hose Systems," dated 3/9/2011.
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PINGP Page A-117 - Revision 1
NFPA 805 Section #
3.6.2
Plant Documentation
Subsection Title
Standpipe and
Hose Stations
Requirement/Guidance
A capability shall be
provided to ensure an adequate water flow rate and nozzle pressure for all hose stations. This capability includes the provision of hose station pressure reducers where necessary for the safety of plant industrial fire brigade members and off-site fire department personnel.
Compliance Statement
Complies
Compliance Basis
Per the NFPA 14–1969 Code Compliance Review, Rev 1 table dated 2/15/2011, "The Mississippi River and the fire pumps can provide adequate water volume and pressure for the most remote hose stations."
Per the NFPA 14–1986 Code Compliance Review, Rev. 1 table dated 3/9/2011, "The two fire pumps can provide adequate pressure for the most remote hose stations."
Pressure reducers are not required on standpipe outlets for 2-1/2 inch hose because it is assumed 2-1/2 inch hose will be attached only when the
NFPA 14–1969 Code Compliance Review Rev. 1, dated 2/15/2011 NFPA 14–1986 Code Compliance Review Rev. 1, FPEE-11-019, dated 3/9/2011 City of Red Wing Job Description, Firefighter/Paramedic: City of Red Wing Job Description, Paid On Call Firefighter, F5, Firefighting, Rev 33. F5 Appendix J, Fire Drills, Rev 14
Industry-Related References
NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1969 and 1986 Editions Minnesota Fire Service Certification, Fire Fighter II Skills Testing NFPA 1001 Fire Fighter II, 2008 Edition
persons likely to use it are trained in handling large streams. Since the brigade is qualified to use large hose streams, they are also qualified to use 1-1/2 inch hose at higher pressures. Training on the use of large hose is conducted annually at the live fire training session.
City of Red Wing Job Description, Firefighter/Paramedic: Minimum Qualifications:
The job requires two years of formal training beyond high school diploma, 1 year of related experience, and Firefighter II certification. National Registry EMT-Paramedic Certification
Identifier
None
Items for Implementation None
EEEE Description Summary required.
City of Red Wing Job Description, Paid On Call Firefighter, The job requires a high school diploma or equivalent and the ability to obtain Fire Fighter II and First Responder certifications. Must obtain a valid Minnesota Class B driver's license with airbrake endorsement or equivalent within 3 years of employment, and have a good driving record. Demonstrated proficiency on personal computers with word processing is desired.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-118 - Revision 1
NFPA 805 Section #
3.6.3
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Standpipe and
Hose Stations
Requirement/Guidance
The proper type of hose
nozzle to be supplied to each power block area shall be based on the area fire hazards. The usual combination spray/straight stream nozzle shall not be used in areas where the straight stream can cause unacceptable damage or present an electrical hazard to fire- fighting personnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltage shock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flow from full open to full closed.
Compliance Statement
Complies
Compliance Basis
Per Section 3.8 of System Description B31A, "Fire Protection System," "Six low gallonage hose stations are located in the plant. Each of the hose stations is equipped with a continuous flow reel, 100' of 1" solid rubber hose, and a 1" low gallonage fog nozzle. The hose stations are located along the 'G' wall on the 715' and 735' levels of the Turbine Building for use in the Relay and Cable Spreading Room, the Unit 1 & Unit 2 480V & 4.16KV Safeguards Switchgear Rooms, and the Control Room. The low gallonage hose stations are designed to prevent the extensive damage which could occur in the rooms from the use of the existing 1-1/2" hose stations."
Per Section 7.10 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "All fire hose stations are equipped with adjustable fog nozzles (95 gpm) and 75 or 100 feet of 1 1/2" fire hose. They can be used on all types of fires."
Per Section 3.4 of Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," "Standpipe and water fog hose stations conforming to NFPA Standard #14
Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," dated 9/15/70 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 System Description B31A, "Fire Protection System," Rev. 12, dated 9/16/11
Industry-Related References FAQ 06-0019, "Definition of 'Power Block' and 'Plant'," Rev. 4, dated 9/28/07 NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1969 and 1986 Editions
are located on all floors of the service, auxiliary, and reactor buildings. These stations are so located that all areas are protected by a fog nozzle when attached to one 75 ft. length of fire hose.” Per Section 4.3.1.4 of NRC SER dated 9/6/79, "Interior hose stations are provided throughout most areas of the plant connected to the fire water header. Most hose stations consist of a pin lug type hose rack, 300 psi hose valve with drip vent and 50 to 100 feet of 1 1/2-inch unlined linen hose coupled to an all fog type nozzle with a ball shut
Identifier None
Items for Implementation
None
EEEE Description Summary off feature… The licensee has agreed to improve the interior fire hose stations by providing the following modifications:… (3) A one-inch booster hose with variable gallonage nozzle with shut off will be provided adjacent to the existing hose stations numbers 21, 23, 24, 64, 69 and 70..." These six low gallonage hose stations have been installed for use in the Relay and Cable Spreading Room, the Unit 1 & Unit 2 480V & 4.16KV Safeguards Switchgear Rooms, and the Control Room.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-119 - Revision 1
NFPA 805 Section #
3.6.4
Plant Documentation
Subsection Title
Standpipe and
Hose Stations
Requirement/Guidance
Provisions shall be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE).
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Item No. 9 of Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76, "As permitted in Appendix A to APCSB 9.5-1, the system has not been analyzed to withstand the SSE. The system can function under all expected natural phenomena characteristic of the region. The system is not subject to man-created site related event damage."
Per Section 4.3.1.4 of NRC SER dated 9/6/79, "interior fire hose stations satisfy the objectives identified in…this report and are, therefore, acceptable."
Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76 NRC SER dated 9/6/79 EDMG-2, Guideline for Damage Mitigation Strategies, Rev. 8
Industry-Related References
NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1969 and 1986 Editions
The hose stations and standpipes provided for PINGP are in accordance with the requirements of BTP 9.5-1, Appendix A for plants which received a construction permit before July 1, 1976 which do not require a seismic category I water system.
Provisions to supply water to standpipes and hose stations for manual fire suppression in the event of a safe shutdown earthquake are outlined in
Identifier
None
Items for Implementation None
EEEE Description Summary EDMG-2, Guideline for Damage Mitigation Strategies (Attachment B - Fire System Management and Attachment L - Establishing Emergency Water Supply).
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-120 - Revision 1
NFPA 805 Section #
3.6.5
Plant Documentation
Subsection Title
Standpipe and
Hose Stations
Requirement/Guidance
Where the seismic
required hose stations are cross-connected to essential seismic non- fire protection water supply systems, the fire flow shall not degrade the essential water system requirement.
Compliance Statement
Complies
Compliance Basis
Per Section 10.3.1.2.1 of the Updated Safety Analysis Report (USAR), "In addition to the two dedicated fire pumps, a third pump, electric motor- driven and having a capacity of 2000 gpm at a pressure of 125 psi, normally assigned to the screen wash function, can be aligned to pump into the fire water system. The cooling water system provides a backup source of water for the fire protection headers. Via normally closed valves, the cooling water system can provide a backup
Calculation ENG-ME-160, "Aux Bldg Fire Protection to Cooling Water Header Cross Tie Hydraulic Calc," Rev. 2 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 System Description B31A, "Fire Protection System," Rev. 12, dated 9/16/11 Updated Safety Analysis Report (USAR) Section 10, Rev. 32P
Industry-Related References
NFPA 14, "Standard for the Installation of Standpipe and Hose Systems," 1969 and 1986 Editions
water supply for the following safeguards equipment fire protection spray and sprinkler systems: auxiliary feedwater pumps, diesel generators, containment cable penetrations, and safeguards cooling water pumps."
Per Section 10.4.1.2.2, "Upon occurrence of the design basis seismic event, the cooling water system flow demand is approximately 36,691 gpm… This is with the two diesel driven safeguards cooling water pumps operating, since
Identifier
None
Items for Implementation None
EEEE Description Summary this creates the highest demand on the suction supply. There is an additional 2000 gpm demand from the diesel fire pump. Initially, the supply to the safeguards cooling water pumps is from both the intake canal and the emergency intake line. The stability of the intake canal banks has been evaluated… The evaluations demonstrate that the intake canal will support the safeguards function of the cooling water system. The volume in the intake canal provides approximately 4 hours for a flow demand of 38,691 gpm…"
Per Section 4.1.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "One of the fire pumps is diesel engine-driven and the other is electric motor-driven. A third pump, electric motor-driven also having a capacity of 2000 gpm at a pressure of 125 psi normally assigned to the screen wash function can be aligned to pump into the fire water system… The screen wash pump [is] used as a backup fire pump… A back-up supply of water to the fire header system is provided by manually- operated crossovers between the Cooling Water System and the main loop and sprinklers within the Fire Protection System. The Cooling Water System and Fire Protection System normally are isolated from each other. At the crossovers, isolation is provided by a normally closed stop
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-121 - Revision 1
valve and a check valve."
Per Section 2 of System Description B31A, "Fire Protection System," "The Cooling Water System also acts as a backup water supply for the fire protection header through 9 separate cross- connects between the two systems. A manual isolation valve and a check valve are installed on each cross-connect to ensure the direction of flow is from the Cooling Water System to the Fire Protection System."
Per the Purpose and Section 5.0 of Calculation ENG-ME-160, "Aux Bldg Fire Protection to Cooling Water Header Cross Tie Hydraulic Calc," "The purpose of this hydraulic calculation is to determine the configuration of cooling water (CL) to fire protection (FP) cross-ties required to supply the Auxiliary Building FP sprinkler header...Based on the results of this analysis, the cooling water system is capable of providing 85% of design flow to DM-7 and 100% of design flow to WPS-22 while meeting the design constraints outlined in this calculation. These flows are adequate to provide auxiliary building fire protection sprinkler systems during those situations where the normal supply must be isolated.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-122 - Revision 1
NFPA 805 Section #
3.7
Plant Documentation
Subsection Title
Fire Extinguishers
Requirement/Guidance
Where provided, fire
extinguishers of the appropriate number, size, and type shall be provided in accordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted to be positioned outside of fire areas due to radiological conditions.
Compliance Statement
Complies
Compliance Basis
Per Section 4.2 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Portable and wheeled fire extinguishers are provided throughout the plant. Most of the extinguishers are of the CO2 to dry chemical type." Table 6-1 of Procedure F5 Appendix F establishes the location and type of each fire extinguisher in the plant.
Fire extinguishers have been reviewed against the requirements of NFPA 10, as detailed in the Code Compliance Review "Code Compliance Review – NFPA 10–1969, Standard for Portable Fire Extinguishers, Code Compliance Deviations" Rev 1, dated 3/26/2010 and "Code Compliance
Code Compliance Review NFPA 10–1969, "Code Compliance Review – NFPA 10–1969, Standard for Portable Fire Extinguishers, Code Compliance Deviations" Rev 1, dated 3/26/2010 Code Compliance Review NFPA 10–1986, "Code Compliance Review – NFPA 10–1986, Standard for Portable Fire Extinguishers," dated 8/16/07 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11
Industry-Related References
NFPA 10, "Standard for the Installation of Portable Fire Extinguishers," 1969 Edition NFPA 10, "Standard for Portable Fire Extinguishers," 1986 Edition
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-124 - Revision 1
NFPA 805 Section #
3.8.1
Plant Documentation
Subsection Title
Fire Alarm
Requirement/Guidance
Alarm initiating devices
shall be installed in accordance with NFPA 72, National Fire Alarm Code ®. Alarm annunciation shall allow the proprietary alarm system to transmit fire- related alarms, supervisory signals, and trouble signals to the control room or other constantly attended location from which required notifications and response can be initiated. Personnel assigned to the proprietary alarm station shall be permitted to have other duties. The following fire-related signals shall be transmitted: (1) Actuation of any fire detection device (2) Actuation of any fixed fire suppression system (3) Actuation of any manual fire alarm station (4) Starting of any fire pump (5) Actuation of any fire protection supervisory device (6) Indication of alarm system trouble condition
Compliance Statement Complies with use of Existing Engineering Equivalency Evaluations
Compliance Basis
Per Section 3.0 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Fire Detection Systems (FDS) are provided throughout the plant. The fire detectors are connected to an alarm which annunciates either in the Control Room or at local panels, which in turn alarms in the Control Room, thus providing the Control Room operator with an indication of system operation, tampering or malfunction. In addition to handling fire detector signals, the system transmits indication of water flow from the sprinkler and deluge extinguishing systems. In all cases, the zone from which the alarm or signal is initiated is indicated. Indication of operation of the CO2 extinguishing system is indicated on a separate panel. The fire detection system is powered from an emergency source of power and all circuits from local control panel to Control Room are electrically supervised."
Per Section 3.0, "The Fire Detection System meets the requirements established by Underwriter’s Laboratories and Factory Mutual. These requirements comply with National Fire Protection Association (NFPA) recommendations in effect when the plant was constructed (1969)."
Per Technical Specification 3.14.A, identified in letter from Davis (NRC) to Mayer (NSP) dated 2/14/78, "The minimum number of fire detection instruments for each fire zone specified in Table TS.3.14-1 shall be operable."
Fire alarm systems have been reviewed against the requirements of NFPA 72D, as detailed in the NFPA 72D–1967 and NFPA 72D–1986 code review checklists.
Letter from Mayer (NSP) to Director (NRC) dated 11/30/79 Letter from Clark (NRC) to Mayer (NSP) dated 12/29/80 Letter from NSP to Director of Nuclear Reactor Regulation (NRC) dated 10/24/80 Letter from Schwencer (NRC) to Mayer (NSP), dated 4/21/80 Letter from Long (NRC) to Parker (NSP) dated 4/28/92 NRC SER dated 9/6/79 NFPA 72D Code Conformance Review Checklist Standard for the Installation, Maintenance and Use of Proprietary Protective Signaling Systems for Watchman, Fire Alarm and Supervisory Service, 1967 Edition, Rev. 1 2/4/2011.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
NFPA 72D, "Proprietary Protective Signaling Systems for Watchman, Fire Alarm and Supervisory Service," 1967 Edition NFPA 72D, "Standard for the Installation, Maintenance and Use of Proprietary Protective Signaling Systems," 1986 Edition
Each deviation has a separate justification, see evaluation for detailed bases.
Each deviation has a separate justification, see evaluation for detailed bases.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-128 - Revision 1
NFPA 805 Section #
3.8.1.1
Plant Documentation
Subsection Title
Fire Alarm
Requirement/Guidance
Means shall be provided
to allow a person observing a fire at any location in the plant to quickly and reliably communicate to the control room or other suitable constantly attended location.
Compliance Statement
Complies
Compliance Basis
Per Section 2.1 of Procedure F5, "Fire Fighting," the "Individual discovering a fire shall immediately report it to the Control Room (CR) using the emergency numbers listed on the telephone, (4911) giving location, type and intensity."
Per Section 10.3.8 of the Updated Safety Analysis Report (USAR), "A fixed public address system interfaced with UPS powered Private Branch Exchange (PBX) telephone system provide
normal and emergency communications. In the event of a PBX failure, power fail telephone stations from the local telephone office and extensions operating on the Xcel Energy Sherco Plant Telephone Switch could be utilized to conduct emergency communications. In addition, a sound powered communications system is
Identifier
None
Items for Implementation None
EEEE Description Summary installed with communications jacks located throughout the plant. The sound powered system requires no external power, and headsets for use with the system are readily available."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-129 - Revision 1
NFPA 805 Section #
3.8.1.2
Plant Documentation
Subsection Title
Fire Alarm
Requirement/Guidance
Means shall be provided
to promptly notify the following of any fire emergency in such a way as to allow them to determine an appropriate course of action:
(1) General site population in all occupied areas
(2) Members of the industrial fire brigade and other groups supporting fire emergency response
(3) Off-site fire emergency response agencies. Two independent means shall be available (e.g., telephone and radio) for notification of off-site emergency services.
Compliance Statement
Complies
Compliance Basis
(1) Per Sections 2.1 and 2.2 of Procedure F5, "Fire Fighting," the individual discovering the fire is required to notify the Control Room via an emergency telephone. "The Control Room Operator shall respond as directed in the Annunciator Response Procedure for a fire detection panel alarm."
Per Section 4.6, a "Fire Alarm page" is sent by the Control Room Operator to notify plant occupants of a fire alarm condition.
Per Section 7.2.1 of Procedure E-PLAN, "Emergency Plan," All emergency operating facilities have at least two means of communications: (1) portable or installed radio systems; and (2) normal telephone communications. The normal onsite communications during an emergency will be made via the plant telephone system with a public address system option. The telephone system is powered by non-interruptible power. The public address system includes about 175 loudspeakers located throughout the entire plant areaThe plant evacuation alarm consists of a 125 VDC operated siren, manually started from the Control Room. This tone consists of a signal starting at approximately 600 cycles per second rising to a
peak of approximately 1450 cycles per second, then returning slowly to the low value of 600 cycles per second and repeating. The Control Room operator can remove the siren tone for emergency voice communication over the loudspeaker PA system.
Identifier
None
Items for Implementation None
EEEE Description Summary
(2) Per Section 4.6 of Procedure F5, "Fire Fighting," the fire brigade members are notified of a fire incident via a "Fire Alarm page" which is sent by the Control Room Operator. Per Section 1.2.6, fire brigade members are equipped with a pager that can be activated from specified telephones.
(3) Per Section 6.1 of Procedure F5, "Fire Fighting," the primary off site fire department (Red Wing Fire Department) is available via telephone.
Per E-Plan Rev. 41, 3/29/2010 Section 7.2.2 Offsite Communications, “Both normal and
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-130 - Revision 1
alternate communication links are provided to offsite agencies. The Xcel Energy telephone network provides normal communications to offsite agencies through telephone lines via the Red Wing US West telephone Exchange, or via Xcel Energy fiber optic SONET communications network. The Control Room, Technical Support Center and Near-Site EOF have a dedicated Xcel Energy radio channel link to the Xcel Energy System Control Center, the Backup EOF, and the Minnesota HSEM Emergency Operating Center in
St. Paul, Minnesota. The Technical Support Center and Near-Site EOF have a National Warning System (NAWAS) extension to the Wisconsin Emergency Management EOC at Madison, the Regional Warning Center at Eau Claire and the Pierce County EOC at Ellsworth, Wisconsin. The Control Room, Technical Support Center and EOF each have a portable cellular phone for emergency communication use, as necessary. The Technical Support Center has access to a computerized auto dial system used for notification of the site’s Emergency Response Organization (ERO). This system consists of a telephone network of several outgoing telephone lines. The Control Room, Technical Support Center and Near-Site EOF have multi-channel radio system for communication with all Plant Radiation Survey Teams, Plant Operations Personnel, Plant Security Areas, county sheriffs, county EOC’s, and Treasure Island Casino (Prairie Island Indian Tribe).”
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-131 - Revision 1
NFPA 805 Section #
3.8.2
Plant Documentation
Subsection Title
Detection
Requirement/Guidance
If automatic fire detection is required to meet the performance or deterministic requirements of Chapter 4, then these devices shall be installed in accordance with NFPA 72, National Fire Alarm Code, and its applicable appendixes.
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Automatic fire detection systems have been
reviewed against the requirements of NFPA 72E, as detailed in "FPP-5 R2, NFPA 72E Code Compliance Evaluation", dated 4/21/04.
Subsequent analysis of identified code deviations was completed and required modifications have been captured in Attachment S, Table S-2.
Letter from Mayer (NSP) to Director (NRC) dated 11/30/79 Letter from Clark (NRC) to Mayer (NSP) dated 12/29/80 Letter from NSP to Director of Nuclear Reactor Regulation (NRC) dated 10/24/80 Letter from Schwencer (NRC) to Mayer (NSP), dated 4/21/80 Letter from Long (NRC) to Parker (NSP) dated 4/28/92 NRC SER dated 9/6/79 Prairie Island Nuclear Generating Plant Appendix 7- NFPA 72E Code Compliance Review FPP-5 NFPA 72, Rev.2 April 21, 2004 NFPA 72E Code Compliance Deviations, FPEE for AR01003334 Design Basis Document DBD-TOP-06, “Fire Protection/Appendix R Design Basis Document,” Rev. 5, dated 10/8/09 Drawing NE-40007-127.1, “0.480Kv Bus 160 Motor Control Center IE Bus 1,” Rev. DB, dated 5/5/79 Drawing NE-40009 Sheet 63, “208/120AC VAC Panel 116,” Rev. BF, dated 10/7/99 Drawing NE-40011 Sheet 36, “BOP Annunciator Point Schematic Diagram Field TB 131-140, SER Points 131-140,” Rev. PS, dated 5/5/00 Drawing NE-40011 Sheet 37, Rev. AV, dated 8/1/05 Drawing NE-40011 Sheet 40, “BOP Annunciator Point Schematic Diagram Field TB 171-180, SER Points 171-180,” Rev. TU, dated 7/27/94 Drawing NE-40011 Sheet 105, “BOP Annunciator Point Schematic Diagram Field TB 821-830, SER Points 821-839,” Rev. G, dated 5/4/00 Drawing NE-40014 Sheet 6, “Fire Detection Control Panel Power Supply,” Rev. R, dated 3/23/94 Drawing NE-40014 Sheet 7, “Fire Detection Control Panel,” Rev. 76, dated 5/09 Drawing NE-40014 Sheet 8, “Fire Detection Control Panel,” Rev. 76, dated 5/09 Drawing NE-40014 Sheet 9, “Fire Detection Control Panel,” Rev. R, dated 11/13/87 Drawing NE-40014 Sheet 10, “Fire Detection Control Panel,” Rev. P, dated 9/14/88 Drawing NE-40014 Sheet 11, “Fire Detection Control Panel,” Rev. Q, dated 9/8/05 Drawing NE-40014 Sheet 12, “Fire Detection Control Panel,” Rev. 76, dated 5/6/06 Drawing NE-40014 Sheet 13, “Fire Detection Control Panel,” Rev. 76, dated 5/09 Drawing NE-40014 Sheet 14, “Fire Detection Control Panel,” Rev. W, dated 4/17/03 Drawing NE-40014 Sheet 15, “Fire Detection Control Panel,” Rev. W, dated 4/22/03 Drawing NE-40014 Sheet 16, “Zones 5 & 7,” Rev. M, dated 11/1/74 Drawing NE-40014 Sheet 17, “Zones 9, 13, & 16,” Rev. C, dated 7/2/84 Drawing NE-40014 Sheet 18, “Zones 18 & 22,” Rev. G, dated 9/25/95 Drawing NE-40014 Sheet 19, “Zones 38, 41, & 45,” Rev. D, dated 12/13/72
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Identifier EEEE Description Summary None Items for Implementation As described in Table S-2, item #8, required modifications have been identified and will be completed in accordance with the associated schedule.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-136 - Revision 1
NFPA 805 Section #
3.9.1
Plant Documentation
NRC SER dated 9/6/79
Subsection Title Automatic and Manual Water- Based Fire Suppression
Requirement/Guidance
If an automatic or manual water-based fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including the following:
(1) NFPA 13, Standard for the Installation of Sprinkler Systems
(2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection
(3) NFPA 750, Standard on Water Mist Fire Protection Systems
(4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix F, "Fire Hazard Analysis," "The Fire Protection System was designed for continuous standby service. The system was designed in accordance with some sections of the standards to the National Fire Protection Association (NFPA) which were applicable in 1969 (1989 for D5/D6), except as identified and justified in the NFPA code compliance review. The system was also based on general recommendations of the Nuclear Energy Property Insurance Association (NEPIA) - now Nuclear Electric Insurance Limited (NEIL)."
Water mist and foam-water systems are not installed at PINGP.
Water-based fire suppression systems have been reviewed against the requirements of NFPA 13 and 15, as detailed in the system specific Code Compliance Reviews listed under Existing Engineering Equivalency Evaluations.
NFPA 13, "Standard for the Installation of Sprinkler Systems," 1969 and 1987 Editions NFPA 15, "Standard for Water Spray Fixed Systems for Fire Protection," 1969 Edition
FPEE-11-009 Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1991 PA-14 and PA-15, Automatic Preaction Systems Turbine Generator and Exciter Bearings Revision 0 FPEE-11-052 Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1969, 1980, 1989, Stairway Suppression Systems SWP-1, SWP-2, SWP-3, SWP-4, SWP-5, SWP-6, SWP-12, SWP-13, SWP-14, SWP-31, D5/D6 Stair (2FP-92-2) Revision 0 Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1989 WPS-32 and 33 D5/D6 Fuel and Lube Oil Storage Tank Room, Code Compliance Review
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-137 - Revision 1
Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1969 PA-1 DG-1 & DG-2 Rooms Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1969, PA-9, Pre-Action Sprinkler System Screenhouse, Revision 1 Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1989 D5 PA-12 Code Compliance Review, NFPA 13, Standard for the Installation of Sprinkler Systems, 1989 D6 PA-13 FPEE-11-011 Code Compliance Review NFPA 13, Standard for the Installation of Sprinkler Systems, 1969 WPS-15, WPS-16 and WPS-17, Wet Pipe Sprinkler Systems, Unit 2 Turbine Building 695ft, Revision 0 FPEE-11-042 Code Compliance Review NFPA 13, Standard for the Installation of Sprinkler Systems, 1969, WPS-10, Wet Pipe Sprinkler Systems Auxiliary Feedwater Pump Rooms, Revision 1 FPEE-11-010 Code Compliance Review NFPA 13, Standard for the Installation of Sprinkler Systems, 1969, WPS-7, WPS-8 and WPS-9, Wet Pipe Sprinkler Systems Unit 1 Turbine Building 695ft, Revision 0 FPEE-11-057 Code Compliance Review NFPA 13, Standard for the Installation of Sprinkler Systems, 1969, WPS-23, WPS-24, Wet Pipe Sprinkler Systems OCS Room (Records Room, Hot Instrument Lab) U1 Auxiliary Building 735ft, Revision 0 Code Compliance Review NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection, 1969, Turbine Oil Reservoir Water Spray Systems 11(DA-3) & 12 (DA-4)
Items for Implementation None
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-138 - Revision 1
NFPA 805 Section #
3.9.2
Plant Documentation NRC SER dated 9/6/79
Subsection Title
Automatic and
Manual Water- Based Fire Suppression Systems
Requirement/Guidance
Each system shall be
equipped with a water flow alarm.
Compliance Statement
Complies
Compliance Basis
Per Section 4.2 of NRC SER dated September 6, 1979, "a fire detection and signaling system is provided throughout many areas of the plant which transmits alarm and supervisory signals to the control room where they are annunciated at the fire panel. In addition to handling fire detector signals, the system transmits indications of water
System Description B31A, "Fire Protection System," Rev. 12, dated 9/16/11
flow from the sprinkler and deluge extinguishing systems."
Per Section 2 of System Description B31A, "Fire Protection System," "Each sprinkler system is equipped with a readily visible, manually operated,
Identifier
None
EEEE Description Summary shutoff valve to isolate the system and a flow switch or alarm check valve which activates an alarm upon initiation of water flow."
Items for Implementation None
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-139 - Revision 1
NFPA 805 Section #
3.9.3
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Automatic and
Manual Water- Based Fire Suppression Systems.
Requirement/Guidance
All alarms from fire
suppression systems shall annunciate in the control room or other suitable constantly attended location.
Compliance Statement
Complies
Compliance Basis
Per Section 4.2 of NRC SER dated 9/6/79, "A fire detection and signaling system is provided throughout many areas of the plant which transmits alarm and supervisory signals to the control room where they are annunciated at the fire panel. In addition to handling fire detector signals, the system transmits indications of water flow from the sprinkler and deluge extinguishing
Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," dated 9/15/70
systems. The system also indicates the operation of the carbon dioxide (CO2) extinguishing systems. In all cases, the zone from which the alarm or supervisory signal is initiated is indicated."
Per Section 4.1.3 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Water flows from the
Identifier
None
Items for Implementation None
EEEE Description Summary automatic wet pipe and preaction deluge suppression systems are annunciated on the fire panel in the Control Room. Flows from manual hose stations are not annunciated, but they will cause the fire pump to start, thereby transmitting a "fire pump running" signal to the Control Room."
Per Section 2.0 of Pioneer Service & Engineering Co. System Description 323.6720, "Prairie Island Fire Protection System," "The deluge and sprinkler systems contain integral alarm circuits to warn of fire, equipment malfunction, or tampering. These audio and visual alarms, together with pressure gages and test and reset switches, provide complete monitoring of the fire protection system from the control room.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-140 - Revision 1
NFPA 805 Section #
3.9.4
Plant Documentation
NRC SER dated 9/6/79
Subsection Title Automatic and Manual Water- Based Fire Suppression Systems.
Requirement/Guidance
Diesel-driven fire pumps
shall be protected by automatic sprinklers.
Compliance Statement
Complies with Item for Implementation
Compliance Basis Per Section 5.19.4 of NRC SER dated 9/6/79, the electric motor-driven and the diesel engine-driven fire pumps are located in the lower level of the screenhouse. The entire screenhouse is protected by a preaction type automatic sprinkler system.
Hazard Analysis," the diesel-driven fire pump is located in Fire Area 41B, and is protected by a full- coverage, pre-action sprinkler system. Attachment S-2 item # 9 will move or install a sprinkler head above the Diesel Driven Fire Pump because of a large obstruction.
Identifier
None
EEEE Description Summary
Items for Implementation
As described in Table S-2, item #9, the modification will move or install a sprinkler head above the Diesel Driven Fire Pump because of a large obstruction.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-141 - Revision 1
NFPA 805 Section #
3.9.5
Plant Documentation
Subsection Title
Automatic and
Manual Water- Based Fire Suppression Systems.
Requirement/Guidance
Each system shall be
equipped with an OS&Y gate valve or other approved shutoff valve.
Compliance Statement
Complies
Compliance Basis
Per Section 7.4.1 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The system shall be functional at all times with a functional flow path capable of taking suction from the river and transferring water through distribution piping with functional sectional
control or isolation valves to the yard hydrant valves and the first isolation valve for each deluge system, hose station, or sprinkler system."
Per Section 3.9 of System Description B31A, "Fire Protection System," the wet pipe sprinkler "systems are monitored by alarm valves and flow switches. A readily visible, manually operated
Identifier
None
Items for Implementation None
EEEE Description Summary shutoff valve is located upstream of each alarm valve.
Per Section 3.11, the deluge systems are provided with "A "Suprotex" valve… to isolate the sprinklers from the water supply…"
Per Section 3.12, "All of the [Preaction Deluge Systems]… except for PADs 12 and 13 (Unit 2, D5/D6 Diesel Generator Building), are equipped "Suprotex" deluge valves for isolating the system water supply… The PAD systems for the D5/D6 Building are equipped with ASCOA "Model F" pre- action valves for isolating the system water supply."
Per Section 3.13, "Manual butterfly isolation valves are provided upstream of the deluge valve…" on the Manual Deluge Systems.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-142 - Revision 1
NFPA 805 Section #
3.9.6
Plant Documentation
Subsection Title
Automatic and
Manual Water- Based Fire Suppression Systems.
Requirement/Guidance
All valves controlling
water-based fire suppression systems required to meet the performance or deterministic requirements of Chapter 4 shall be supervised as described in 3.5.14.
Compliance Statement
Complies
Compliance Basis
Per Section 8.5.11 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "Each valve (manual, power operated or automatic) in the flow path for safety- related areas and areas posing a fire hazard to safety-related areas shall be verified to be in the correct position and secured to prevent inadvertent misalignment every month."
Per Section 1.1 of Procedure SP 1200, "Fire
Procedure 5AWI 3.10.1, "Methods of Performing Verifications," Rev. 16, dated 6/30/10 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 Procedure SP 1200, "Fire Protection System Supply to Safety Related Areas Valve Check," Rev. 32, dated 11/20/08
Protection System Supply to Safety Related Areas Valve Check," "The purpose of this surveillance is to verify that fire protection valves supplying safety related areas, and areas posing a fire hazard to safety related areas, are in the correct position and secured."
Per Section 1.5.1, the acceptance criteria for all water supply and fire suppression system control valves is "Valve verified open per 5AWI 3.10.1 Appendix E" and "Block Wire in place."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per Section 5.1, "Block Wire is required to prevent inadvertent mispositioning of valves."
Appendix E of Procedure 5AWI 3.10.1, "Methods of Performing Verifications," provides instructions for inspecting valves to ensure they are adequately locked.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-144 - Revision 1
NFPA 805 Section #
3.10.1
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
If an automatic total
flooding and local application gaseous fire suppression system is required to meet the performance or deterministic requirements of Chapter 4, then the system shall be designed and installed in accordance with the following NFPA codes:
(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems
(2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems
(3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Per Section 4.0 of Procedure F5 Appendix F, "Fire Hazard Analysis," The Fire Protection System was designed for continuous standby service. The system was designed in accordance with some sections of the standards to the National Fire Protection Association (NFPA) which were applicable in 1969 (1989 for D5/D6), except as identified and justified in the NFPA code compliance review. The system was also based on general recommendations of the Nuclear Energy Property Insurance Association (NEPIA) - now Nuclear Electric Insurance Limited (NEIL)."
Per Section 4.2.7, "The relay and cable spreading room is protected by an automatic 6 ton carbon dioxide (CO2) suppression system with alarm sirens and a sixty-second delay. A detection system and a thermal actuation system is provided. Products of combustion detectors provide an early warning alarm to the Control Room. The auto action mode is normally bypassed when the room is occupied; however, the carbon dioxide system may be manually actuated at any time. The carbon dioxide system is backed up by manual hose stations and extinguishers. The system is designed for total flooding application with a 50 percent concentration for 15 minutes. Storage tank capacity is adequate for two shots.
Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 FPEE-11-038 Code Compliance Review, NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 1972, Relay and Computer Room, Fire Area 18, 10/4/2011.
Industry-Related References
NFPA 12, "Standard on Carbon Dioxide Extinguishing Systems" NFPA 12A, "Standard of Halon 1301 Fire Extinguishing Systems"
The system is actuated by thermal detectors with provisions for manual actuation. The area is also provided with fire detectors which alarm on the control Room fire panel." Per Section 4.3.2 of NRC SER dated 9/6/79, "A carbon dioxide (CO2) type gas fire suppression system is provided in the cable spreading and relay room. The computer room is located within this
Identifier
None
Items for Implementation
EEEE Description Summary room and is also protected by the CO2 system. The system is designed for total flooding application with a 50 percent concentration for 15 minutes. Storage tank capacity is adequate for two shots. The system
Resolution of AR01306187 - NFPA Code issue: Thermal detection for actuation of FP System. Identified deviations require a modification to resolve noncompliances associated with unprotected beam pockets and system supervision. This modification is identified in Attachment S, Table S-2.
is actuated by thermal detectors with provisions for manual actuation. The area is also provided with early warning detectors which indicate an alarm but does not actuate the CO2 suppression system. We find that the gas fire suppression system satisfies the objectives identified in Section 2.2 of this report and is, therefore, acceptable.”
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-145 - Revision 1
NFPA 2001 does not apply, as there are no clean agent extinguishing systems at PINGP.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-146 - Revision 1
NFPA 805 Section #
3.10.2
Plant Documentation
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other constantly attended location identified.
Compliance Statement
Complies
Compliance Basis
Per Section 3.0 of Procedure F5 Appendix F, " Fire Hazard Analysis," "Fire Detection Systems (FDS) are provided throughout the plant. The fire detectors are connected to an alarm which annunciates either in the Control Room or at local panels, which in turn alarms in the Control Room, thus providing the Control Room operator with an indication of system operation, tampering or
malfunction. In addition to handling fire detector signals, the system transmits indication of water flow from the sprinkler and deluge extinguishing systems. In all cases, the zone from which the alarm or signal is initiated is indicated. Indication of operation of the CO2 extinguishing system is indicated on a separate panel." Per Table 6-2, the
Identifier
None
Items for Implementation None
EEEE Description Summary Relay and Cable Spreading Room (Fire Area 18) and the Service Building/Computer Room (Fire Area 94) are monitored in the Detection Zone Control Room Circuit. Per Section 4.1, "System General Precautions", of System Description B31A, the Cardox suppression system is annunciated in the Control Room and "Upon actuation of the Cardox System, the Control Room operator shall dispatch an operator to clear any personnel from the Relay/Computer Rooms..."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-147 - Revision 1
NFPA 805 Section #
3.10.3
Plant Documentation
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Ventilation system design shall take into account prevention from over-pressurization during agent injection, adequate sealing to prevent loss of agent, and confinement of radioactive contaminants.
Compliance Statement
Complies
Compliance Basis Per DBD-TOP-06, “Design Bases Document for Fire Protection / Appendix R”, Section 4.4.3 (E) (1) “The Relay Room penetration seals are protected from overpressurization on a Cardox actuation by allowing the room doors to open to relieve excess pressure. The Unit 1 Relay Room door latch deenergizes after a 50-second time delay, allowing excess pressure to force the door open. The Unit 2 Relay Room door latch deenergizes if the room pressure exceeds 14 inches water column, allowing the door to open when the pressure becomes excessive.”
Design Bases Document for Fire Protection / Appendix R, DBD TOP-06, Revision 5, 10/8/09 SP 1194, Cardox (Carbon Dioxide) 18 Month System Test, Revision 22, dated 7/10/13
Per SP 1194, “Cardox (Carbon Dioxide) 18 Month System Test” Section 8.0, NOTE: “The desired “TRIP” point for DPS-99753 is 14 inches of H2O, or 0.5 psi.”
Confinement of radioactive contaminants is not required as there are no gaseous fire suppression
Identifier
None
Items for Implementation None
EEEE Description Summary systems within the radiologically controlled area.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-148 - Revision 1
NFPA 805 Section #
3.10.4
Plant Documentation
None
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
In any area required to
be protected by both primary and backup gaseous fire suppression systems, a single active failure or a crack in any pipe in the fire suppression system shall not impair both the primary and backup fire suppression capability.
Compliance Statement
N/A
Compliance Basis
Not Applicable. PINGP does not have areas protected by both primary and backup gaseous suppression.
that status of plant equipment is known at all times."
Per Section 3.10, "Repositioning of equipment during the conduct of maintenance and/or testing that is not covered by written instruction may be
Identifier
None
Items for Implementation None
EEEE Description Summary performed by qualified individuals provided these manipulations are authorized by shift supervision. Such manipulations must be formally tracked to ensure proper repositioning."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-150 - Revision 1
NFPA 805 Section #
3.10.6
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Total flooding carbon
dioxide systems shall not be used in normally occupied areas.
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 4.3.2 of NRC SER dated 9/6/79, "A carbon dioxide (CO2) type gas fire suppression system is provided in the cable spreading and relay room. The computer room is located within this room and is also protected by the CO2 system. The system is designed for total flooding
application with a 50 percent concentration for 15 minutes… The area is provided with early warning detectors which indicate an alarm but does not actuate the CO2 suppression system. We find that the gas fire suppression system satisfies the objectives identified in Section 2.2 of this report and is, therefore, acceptable."
Identifier
None
Items for Implementation None
EEEE Description Summary
Per Section 3.18 of System Description B31A, "Fire Protection System," "The Cardox System provides a total flood capability for the Relay/Computer Room in the event of a fire… Upon receipt of either an automatic or a manual signal, an alarm and a 60 second timer are actuated. The timer provides a delay period prior to system actuation to ensure all personnel have adequate time to evacuate the room."
Section 5.14.2 of Procedure C31, "Fire Protection & Detection Systems," defines the steps required to bypass the Cardox Dioxide System in the Computer Room. The Cable Spreading and Relay room and the Computer room are all considered normally unoccupied spaces. The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-151 - Revision 1
NFPA 805 Section #
3.10.7
Plant Documentation
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Automatic total flooding
carbon dioxide systems shall be equipped with an audible pre-discharge alarm and discharge delay sufficient to permit egress of personnel. The carbon dioxide system shall be provided with an odorizer.
Compliance Statement
Complies with Item for
Implementation
Compliance Basis
Per Section 7.8 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The relay and cable spreading room is protected by an automatic 6 ton carbon dioxide (CO2) suppression system with alarm sirens and a sixty- second delay. A detection system and a thermal actuation system are provided. Smoke detectors provide an early warning alarm to the Control Room."
Per Section 3.18.B of System Description B31A,
Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012 System Description B31A, "Fire Protection System." Rev. 12, dated 9/16/11 FPEE-11-038 Code Compliance Review, NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 1972, Relay and Computer Room, Fire Area 18
Industry-Related References
National Fire Protection Association (NFPA) 12 -1972, Standard on Carbon Dioxide Extinguishing Systems.
"Fire Protection System," "Upon receipt of either an automatic or manual signal, an alarm and a 60 second timer are actuated. The timer provides a delay period prior to system actuation to ensure all personnel have adequate time to evacuate the room."
The requirement for an odorizer is not in the 1972 version of NFPA 12, Standard on Carbon Dioxide Extinguishing Systems which is the system design code of record. However, an odorizer will be
Identifier
None
Items for Implementation
EEEE Description Summary added to the system as a modification as outlined in Attachment S.
An odorizer will be added to the carbon dioxide system protecting the relay and cable spreading room.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-152 - Revision 1
NFPA 805 Section #
3.10.8
Plant Documentation
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Positive mechanical
means shall be provided to lock out total flooding carbon dioxide systems during work in the protected space.
Compliance Statement
Complies
Compliance Basis
Per Section 7.9.1 of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "During those periods when the relay and cable spreading room area is occupied, automatic initiation of the CO2 system may be bypassed. During those periods when the area is normally unoccupied, the CO2 system shall be
The mechanical bypass switch (CS-5843001) located in the Control Room disables the automatic function of the CO2 system, thereby preventing actuation of the system while work is on-going in the area protected by the CO2 system.
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-153 - Revision 1
NFPA 805 Section #
3.10.9
Plant Documentation
NRC SER dated 9/6/79
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
The possibility of
secondary thermal shock (cooling) damage shall be considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide.
Compliance Statement
Complies via Previous Approval
Compliance Basis
Per Item 131 of Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76, "Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1," "CO2 outlet nozzles are positioned to direct stream away from concentrations of cables or cabinets."
Per NRC SER dated 9/6/79, "We find that the gas fire suppression system satisfies the objectives identified in Section 2.2 of this report and is, therefore, acceptable."
Letter from Mayer (NSP) to Stello (NRC) dated 12/8/76, "Comparison of Existing Fire Protection Provisions to the Guidelines Contained in Standard Review Plan 9.5.1,"
Installed Halon 1301 systems are not required to meet the performance-based or deterministic requirements of NFPA 805 Chapter 4.
Identifier
None
Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-154 - Revision 1
NFPA 805 Section #
3.10.10
Plant Documentation
None
Subsection Title
Gaseous Fire
Suppression Systems
Requirement/Guidance
Particular attention shall
be given to corrosive characteristics of agent decomposition products on safety systems.
Compliance Statement
Complies
Compliance Basis
NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 2008, Annex G, General Information on Carbon Dioxide G.1. For fire-extinguishing applications, carbon dioxide has a number of desirable properties. It is noncorrosive, non-damaging, and leaves no residue to clean up after the fire.
Industry-Related References
NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 2008
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-155 - Revision 1
NFPA 805 Section #
3.11
Plant Documentation
N/A
Subsection Title
Passive Fire
Protection Features
Requirement/Guidance
This section shall be
used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-156 - Revision 1
NFPA 805 Section #
3.11.1
Plant Documentation
Subsection Title
Building Separation
Requirement/Guidance
Each major building
within the power block shall be separated from the others by barriers having a designated fire resistance rating of 3 hours or by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, Recommended Practice for Protection of Buildings from Exterior Fire Exposures.
Exception: Where a performance-based analysis determines the adequacy of building separation, the requirements of 3.11.1 shall not apply.
Compliance Statement
Complies
Compliance Basis
The power block is defined in Attachment I to this Transition Report.
Per Section 2.2 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Reinforced concrete (Grade B is used in the construction of structural components) requires the following minimum thicknesses to provide a fire resistance of three hours, per UBC: Walls - 6 1/2 inches; Floors – 4 1/2 inches; For all floors, the minimum required concrete (Grade B) cover thickness of the reinforcing bars is 1 inch; for beams and columns the minimum required concrete cover thickness is 1 1/2 inches. All PINGP reinforced concrete walls supporting fire areas in safety-related structures are at least 12 inches in thickness. All floor slabs between floors are at least 5 1/2 inches in thickness with a concrete covering reinforcing steel of at least 1 inch. Two exceptions are the 8 inch concrete floors over the OSC room and the Operator’s Lounge, which have a concrete cover of 3/4 inch. These two floors, although acting as fire barriers, act only as an enclosing ceiling and support no equipment. The D5/D6 Building three
Industry-Related References FAQ 06-0019, "Definition of 'Power Block' and 'Plant'," Rev. 4, dated 9/28/07 NFPA 80A, "Recommended Practice for Protection of Buildings from Exterior Fire Exposure," 1967 and 1987 Editions
hour barriers adequately separate redundant trains, fuel oil/lube oil storage, and the stairwell. Concrete beam and column reinforcing steel has at least a cover of 1 1/2 inches. In some cases, concrete masonry units (calcareous or siliceous gravel) are used in the construction of fire barrier walls with a minimum thickness of 6 inches. The minimum thickness of these walls required for a fire resistance of three hours is 5.3 inches.”
Identifier None Items for Implementation None
EEEE Description Summary
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-157 - Revision 1
NFPA 805 Section #
3.11.2
Plant Documentation
Subsection Title
Fire Barriers
Requirement/Guidance
Fire barriers required by
Chapter 4 shall include a specific fire-resistance rating. Fire barriers shall be designed and installed to meet the specific fire resistance rating using assemblies qualified by fire tests. The qualification fire tests shall be in accordance with NFPA 251, Standard Methods of Tests of Fire Endurance of Building Construction and Materials, or ASTM E 119, Standard Test Methods for Fire Tests of Building Construction and Materials.
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Per Section 2.1 of Procedure F5 Appendix F, "Fire Hazard Analysis," "Walls that do not protect safe shutdown equipment are maintained as unrated barriers. In all cases, good fire prevention practices are followed."
Per Section 2.2, "Fire Hazard Analysis," "Barriers that must be maintained as three hour rated barriers per BTP 9.5-1 or Appendix R are those that:
A. Separate safety-related systems from any potential fires in non-safety related areas that could affect their ability to perform their safety function;
B. Separate redundant divisions or trains of safety- related systems from each other so that both are not subject to damage from a single fire; or
"All safeguard equipment is located within structures or compartments designed to seismic Category 1 requirements which require the use of building materials inherently resistive to structural damage due to fire. Safety-related systems are also separated from high concentrations of combustible material by reinforced concrete or concrete masonry walls. Reinforced concrete (Grade B is used in the construction of structural
Identifier
FPEE-12-006
EEEE Description
FA 85 Boundaries and F5 Appendix K Barriers
Summary
The evaluation assesses the impact of postulated fires on either side of the Fire Area 85 boundaries that communicate with Fire Areas 60 and 75 on the 715ft elevation and Fire Areas 59 and 74 on the 715ft elevation for impact on fire safe shutdown capability. The evaluation also assesses the location of the F5 Appendix K barrier separating Fire Area 85 and Fire Areas 59 and 74 on the 715ft elevation.
Based on the evaluation, there is reasonable assurance that
components) requires the following minimum thicknesses to provide a fire resistance of three hours, per UBC: Walls - 6 1/2 inches; Floors - 4 1/2 inches;
"For all floors, the minimum required concrete (Grade B) cover thickness of the reinforcing bars is 1 inch; for beams and columns the minimum required concrete cover thickness is 1 1/2 inches."
"All PINGP reinforced concrete walls supporting fire areas in safety-related structures are at least 12 inches in thickness. All floor slabs between floors are at least 5 1/2 inches in thickness with a concrete covering reinforcing steel of at least 1 inch. Two exceptions are the 8 inch concrete floors over the OSC room and the Operator’s Lounge, which have a concrete cover of 3/4 inch.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-158 - Revision 1
FPEE-11-020 NFPA Codes Referenced in
NFPA 805 not addressed by separate code reviews
FPEE-12-010 Fire Area 15 Access Control Ceiling, Revision 0
fire will not spread between Fire Area 85 and Fire Areas 60 and 75 on the 735ft elevation and between Fire Area 85 and Fire Areas 59 and 74 on the 715ft elevation and adversely impact fire safe shutdown capability.
The evaluation documents the review of the NFPA codes referenced in NFPA 805 that are not addressed in individual NFPA code compliance reviews.
A deviation was identified regarding the use of a radiant energy shield in Unit 1 Containment that has not been demonstrated to have a ½- hour fire rating when subject to testing following ASTM E-119. AR 01317872 is tracking resolution of this issue. The purpose of this evaluation is to assess the adequacy of the fire area boundary separating Fire Area 15, Access Control, from Fire Area 59, Auxiliary Building Mezzanine Level Unit 1. Both fire areas are located on the 715ft elevation of the Auxiliary Building. Fire Area 15 is part height within Fire Area 59, with a ceiling height of 10ft 5-1/2in, while Fire Area 59 extends the full height of the 715ft elevation to the underside of the 735ft elevation. The ceiling of Fire Area 15 forms a mezzanine accessible by a ladder within Fire Area 59. The barrier is identified as an F5 Appendix K barrier. The ceiling of Fire Area 15 is constructed of 3in of concrete, and all penetrations are sealed full depth with grout,
These two floors, although acting as fire barriers, act only as an enclosing ceiling and support no equipment.
"Concrete beam and column reinforcing steel has at least a cover of 1 1/2 inches. In some cases, concrete masonry units (calcareous or siliceous gravel) are used in the construction of fire barrier walls with a minimum thickness of 6 inches. The minimum thickness of these walls required for a fire resistance of three hours is 5.3 inches."
Per Section 7.16.A of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "The PINGP has been designed generally with physical separation to prevent the spread of a postulated fire in safe shutdown equipment areas to prevent the loss of both Appendix R trains. This separation is maintained in part by fire resistive compartment isolation of plant safety systems. The walls and ceilings of such compartments are rated fire barriers. These walls and ceilings contain penetrations for the passage of pipes and electrical cables from one fire area to another. Therefore, these penetrations are a breach of the fire barriers and must be sealed so as to maintain the integrity of the fire barriers. PINGP is divided into fire areas based on general plant layout and fire protection equipment. Existing barriers, including the containment vessels, were used whenever possible for the fire area boundaries. All safeguards equipment is located within structures or compartments designed to seismic Category I requirements which require the use of building materials inherently resistive to structural damage due to fire. Safety-related systems are also separated from high concentrations of combustible material by reinforced concrete or concrete masonry walls."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-159 - Revision 1
RTV silicone foam, Kaowool, or thermal insulating wool, with damming material and Flamemastic coating on both sides.
Items for Implementation None
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
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NFPA 805 Section #
3.11.3
Subsection Title
Fire Barrier
Penetrations
Requirement/Guidance
Penetrations in fire
barriers shall be provided with listed fire- rated door assemblies or listed rated fire dampers having a fire resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by Chapter 4. (See 3.11.3.4 for penetration seals for through penetration fire stops.) Passive fire protection devices such as doors and dampers shall conform with the following NFPA standards, as applicable:
(1) NFPA 80, Standard for Fire Doors and Fire Windows
(2) NFPA 90A, Standard for the Installation of Air- Conditioning and Ventilating Systems
(3) NFPA 101, Life Safety Code
Exception: Where fire area boundaries are not wall-to-wall, floor-to- ceiling boundaries with all penetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall be required to assess the adequacy of fire barrier forming the fire boundary to determine if
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Per Section 7.16.1.B of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," "Fire doors and frames in Appendix R-required fire barriers are rated to the equivalent fire resistance duration of three hours, in accordance with the criteria established in NFPA 252, Standard Methods of Fire Tests of Door Assemblies, 1968, ed."
Fire doors have been reviewed against the requirements of NFPA 80, as detailed in the Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1968 ed.," Rev 1, dated 11/7/11 and "Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1986 ed.," dated 11/7/12)
Per Section 3.2.5 of Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80, "the licensee proposed to install fire dampers in all ventilation ducts which are unprotected and could endanger areas containing safe shutdown equipment in the event of a fire. Further, the licensee committed to provide three hour fire dampers in those ducts communicating with the Turbine Building. The licensee has shown to our satisfaction that all ventilation ducts which could endanger areas containing safe shutdown equipment will be protected with fire dampers. Based on our review, we find the licensee's commitment to provide fire dampers in ventilation ducts in all fire zones containing equipment necessary for safe shutdown." Fire dampers have been reviewed against the requirements of NFPA 90A, as detailed in the Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, 1969 ed., dated 9/12/07 and "Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, 1978 ed.," dated 9/12/07.3) Fire rated door assembly requirements are addressed in Section 8.2.3.2.1 of NFPA 101 which refers to NFPA 80, and is evaluated in the Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1968 ed., Rev 1, dated 11/7/11 and "Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1986 ed.," dated 11/7/11. Rated fire dampers requirements are addressed
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the barrier will withstand the fire effects of the hazards in the area.
in Section 9.2.1 of NFPA 101 which refers to NFPA 90A, and is evaluated in the Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating
Letter from Schwencer (NRC) to Mayer (NSP) dated 4/21/80; Fire Protection Engineering Evaluation FPEE-12-003, CA-01311055-01, Fire Door Frames, Revision 0, 4/5/2012; Fire Protection Engineering Evaluation, FPEE-CA124448-02, Revision 0, 1/20/2012; Code Compliance Review NFPA 80–1968, FPEE-11-049, "Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1968 ed.," Rev 1, dated 11/7/11,Code Compliance Review NFPA 80–1986, FPEE-11-019, "Code Compliance Review – NFPA 80, Standard for Fire Doors and Windows, 1986 ed.," dated 11/7/11, FPEE-11-022 Code Compliance Review NFPA 90A–1969, "Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, Code Compliance Review NFPA 90A–1978, "Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, 1978 ed.," dated 11/29/2011Procedure F5 Appendix F "Fire Hazard Analysis," Rev. 25A, dated 8/8/11; Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012
Industry-Related References
NFPA 80, "Standard for Fire Doors and Fire Windows," 1968 and 1986 Editions NFPA 90A, "Installation of Air Conditioning and Ventilating Systems," 1969 and 1978 Editions NFPA 101, "Life Safety Code," 2000 Edition
Systems, 1969 ed., dated 9/12/07 and "Code Compliance Review – NFPA 90A, Standard for the Installation of Air Conditioning and Ventilating Systems, 1978 ed.," dated 9/12/07
Identifier
CE0112159401 FPEE 0113625201 CA 0124445802 CA 0131104601 CA 0131105701 FPEE 0124191701 FPEE 10-006AR 117907003 FPEE 11001 FPEE 11019 FPEE 11021 FPEE 11022 FPEE 11049 FPEE 12002 CA 013274301 FPEE 12003 CA 0131105501 FPEE 12004 CA 0131380801
Items for Implementation None
EEEE Description
Engineering evaluations can
be found in the SharePoint Portal.
Summary
See individual evaluation
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
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NFPA 805 Section #
3.11.4
Subsection Title
Through
Penetration Fire Stops
Requirement/Guidance
Through penetration fire
stops for penetrations such as pipes, conduits, bus ducts, cables, wires, pneumatic tubes and ducts, and similar building service equipment that pass through fire barriers shall be protected as follows.
(a) The annular space between the penetrating item and the through opening in the fire barrier shall be filled with a qualified fire-resistive penetration seal assembly capable of maintaining the fire resistance of the fire barrier. The assembly shall be qualified by tests in accordance with a fire test protocol acceptable to the AHJ or be protected by a listed fire-rated device for the specified fire-resistive period.
(b) Conduits shall be provided with an internal fire seal that has an equivalent fire-resistive rating to that of the fire barrier through opening fire stop and shall be permitted to be installed on either side of the barrier in a location that is as close to the barrier as possible.
Exception: Openings inside conduit 4 in (10.2 cm) or less in diameter shall be sealed at the
Compliance Statement
Complies via Previous
Approval
Compliance Basis
Per Section 2.4 of Procedure F5 Appendix F, "Fire Hazard Analysis," "All (cable) penetration seals passed the criteria of no flame passage, temperature, and hose stream test of IEEE 634. Fire stops are not provided at intermediate points in vertical or horizontal cable spans. Penetrations are sealed with packed thermal fiber or foam and covered with thermal board and approximately 1/8 in. coat of thermal mastic. Where the penetration is through a structure forming the boundary between ventilation zones, fire dampers have been installed except where determined unnecessary by evaluation. Conduit penetrations through walls, floors, and ceilings of the relay/cable spreading rooms are provided with fire stops."
Per Section 2.5, "Most piping penetrations in walls and floors, in safety-related areas of the plant, are sealed. In those instances where seals are not provided, evaluations exist to justify conditions. The small area surrounding pipe is sealed with a qualified penetration seal, designed for the maximum fire severity on either side of the barrier, and to allow for thermal movement. In all cases, when new penetrations are sealed, the sealing material is noncombustible, or, as in the case of silicone foam, has been reviewed and found acceptable by the NRC."
Per Enclosure 1 to Letter from Clark (NRC) to Mayer (NSP) dated 12/29/80, "The test slab containing the penetrations was placed on a horizontal furnace and exposed to the ASTM E119 time/temperature curve for 3-hours. The ASTM E119 time/temperature curve is the standard time/temperature curve used for fire endurance tests. After three hours the test slab was lifted in a horizontal position and subjected to a hose stream test. The hose stream test was performed in accordance with the recommendations of IEEE 634-1978, "Standard Cable Fire Stop Qualification Test," which is an acceptable method. The acceptance criteria for the fire stops was based on the acceptance criteria in IEEE 634-1978. We find the acceptance criteria for the test provide reasonable assurance that the penetration seals will be capable of preventing a fire from spreading from one fire area to another. All of the tested
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fire barrier with a fire- rated internal seal unless the conduit extends greater that 5 fit (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot gas seal in this application
penetration seals qualified as three-hour seals. "Based on our review and the test data, we find that the penetration seals are qualified as 3-hour fire rated seals. Therefore, we conclude that the licensee's proposed modification regarding upgraded penetration firestops is acceptable."
Procedure D52, "Installation Guidelines for the Permanent & Temporary Sealing of Electrical/Mechanical Openings between Established Fire Areas," provides guidelines and procedures for the installation of penetration seals, including material requirements. Section 1.0 states, "The purpose of this installation guideline is to establish the controls & instructions necessary for new, existing and/or temporary electrical/mechanical openings that will be constructed or have been breached during construction/maintenance work. These guidelines are designed to meet the conditions set forth by
Letter from Clark (NRC) to Mayer (NSP) dated 12/29/80 Engineering Manual 2.1.14, "Engineering Design, Fabrication, and Installation Summary for Fire Barriers and Penetration Seals," Rev. 1, dated 1/26/00 Procedure D52, "Installation Guidelines for the Permanent & Temporary Sealing of Electrical/Mechanical Openings between Established Fire Areas," Rev. 13, dated 1/27/09 Procedure F5 Appendix F, "Fire Hazard Analysis," Rev. 25A, dated 8/8/11 Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Rev. 15, dated 4/11/2012
Operations Manual F5 Appendix K (fire penetrations), T.S.3.7.12 for Aux Building Special Vent Zone and H27 for Steam Exclusion."
Per Section 7.16.C of Procedure F5 Appendix K, "Fire Protection Systems Functional Requirements," Piping and electrical penetrations are provided with qualified seals where they penetrate boundaries between fire areas. Seals are qualified for the maximum fire severity present on either side of the barrier. Seals are installed in the annulus with a qualified penetration seal designed to allow for thermal movement.
Identifier
None
Items for Implementation None
EEEE Description Summary
Operations Manual Section D52, "Installation Guidelines for the Permanent and Temporary Sealing of Electrical/Mechanical Openings Between Established Fire Areas", provides the overall requirements to be met for the design of seals on the fire barrier penetrations, thus maintaining fire confinement capability and limiting the spread of a potential fire. D52 includes the requirements for internal conduit seals. The original penetration inventory, sketches, and retrofit modifications are located in QUAD-5-80-008, Rev. 6, PI Rev. 0, and are filmed under modification 79Y084 (Film Reel #0984-0037). The current penetration inventory is maintained in the Penetration Seal Database."
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
PINGP Page A-164 - Revision 1
The basis for approval has been reviewed. There have been no plant modifications or other changes that would invalidate the basis for approval.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
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NFPA 805 Section #
3.11.5
Subsection Title
Electrical Raceway
Fire Barrier System (ERFBS)
Requirement/Guidance
ERFBS required by
Chapter 4 shall be capable of resisting the fire effects of the hazards in the area. ERFBS shall be tested in accordance with and shall be tested in accordance with and shall meet the acceptable criteria of NRC Generic Letter 86- 10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area". The ERFBS needs to adequately address the design requirements and limitations of supports and intervening items and their impact on the fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal exposure, barrier design, raceway size and type, cable size, fill and type shall be demonstrated.
Exception No. 1: When the temperatures inside the fire barrier system exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire Barrier Systems Used to Separate Redundant Safe
Compliance Statement
Complies with use of
Existing Engineering Equivalency Evaluation
Compliance Basis
Electrical raceway fire barrier systems are installed at PINGP.
Per Section 2.1 of Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," "Except as provided for in paragraph G.3 of this section, where cables or equipment, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achieve and maintain Hot Shutdown conditions are located within the same fire area outside of primary containment, one (1) of the following means of ensuring that one (1) of the redundant trains is free of fire damage shall be provided: A. Separation of cables and equipment and associated non-safety circuits of redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers shall be protected to provide fire resistance equivalent to that required of the barrier; B. Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area; or C. Enclosure of cable and equipment and associated non-safety circuits of one (1) redundant train in a fire barrier having a one (1) hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area;"
Per Section 2.6, "the SBO/ESU Project established and enhanced basic routing paths for trained cables. Certain fire areas were designated as basic Train A (or B) routes or areas, meaning that normally Train A (or B) cables would be routed through the area. Thus, in a Train A area, it is expected that Train A systems and components would be affected by a fire and that Train B would be relied on for safe shutdown. This methodology allows protecting (wrapping) the least affected train in each area. (The previous analyses essentially provided for Train A operation in the event of a control room fire and ensuring that Train B would be available for other fires.) For the
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Shutdown Training Within the Same Fire Area," Supplement , functionality of the cable at these elevated temperatures shall be demonstrated. Qualification demonstration of these cables shall be performed in accordance with the electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure."
Exception No. 2: ERFBS systems employed prior to the issuance of Generic Letter 86-10. Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at
most part, Train B would be protected and credited in Train A areas and Train A would be protected and credited in Train B areas."
Per Section 2.7.2, "All cables and components, identified in the circuit analyses as being required for operation, are included in the Safe Shutdown data base along with the cable route and component location. Once all cables and components were entered, reports were generated that summarize required information needed for the overall analysis. A compliance assessment summary, and a compliance assessment report is included for every fire area that contains required safe shutdown cables."
Engineering Manual 4.3.1-E, "Engineering Design, Fabrication and Installation Summary for Fire Barriers," provides detailed specifications and installation instructions for electrical raceway fire barrier systems in accordance with NRC GL 86-10 Supplement 1.
FPEE-2005-001 Qualification of PCI PROMATEC INTERAM" Flexible Fire Protective System, evaluation performed a detailed comparison and analysis of the fire wrap design configurations to fire-tested configurations and concluded that the system applied to 1CB-31 is considered qualified for a fire endurance rating of 1-hour in accordance with USNRC GL 86-10 Supplement 1.
Engineering Manual EM 4.3.1-E, "Engineering Design, Fabrication and Installation Summary for Fire Barriers," Rev. 2, dated 4/19/01 Procedure F5 Appendix E, "Fire Protection Safe Shutdown Analysis Summary," Rev. 14, dated 4/29/10 FPEE-2005-001 Qualification of PCI PROMATEC INTERAM" Flexible Fire Protective System
Industry-Related References
USNRC IN-95-52, Supplement Fire Endurance Test Results for Electrical Raceway Fire Barrier Systems Constructed from 3M Company Interam Fire Barrier Materials.
Northern States Power - Minnesota Attachment A - NEI 04-02 Table B-1
FPEE-2005-001 Qualification of PCI PROMATEC INTERAM" Flexible Fire Protective System
Items for Implementation None
This Fire Protection Engineering Evaluation (FPEE) is written to demonstrate the acceptability of the fire protected conduit 1CB-31. This protection has been achieved with the 3M Interam Flexible Fire Wrap System as necessary to achieve the required fire rating of one-hour.
The evaluation performed a detailed comparison and analysis of the fire wrap design configurations to fire-tested configurations and concluded that the system applied to 1CB- 31 is considered qualified for a fire endurance rating of 1-hour in accordance with USNRC GL 86-10 Supplement 1.
Northern States Power - Minnesota Attachment B – NEI 04-02 Table B-2 – Nuclear Safety Capability Assessment Methodology Review
PINGP Page B-1 – Revision 1
B. NEI 04-02 Table B-2 – Nuclear Safety Capability Assessment Methodology Review
97 Pages Attached
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-2 - Revision 1
2.4.2.1 Nuclear Safety Capability Systems and Equipment Selection (Taken From NFPA 805, 2001 Edition)
A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed. The equipment list shall contain an inventory of those critical components required to achieve the nuclear safety performance criteria of Section 1.5. Components required to achieve and maintain the nuclear safety functions and components whose fire-induced failure could prevent the operation or result in the maloperation of those components needed to meet the nuclear safety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.0 Deterministic Methodology This section discusses a generic deterministic methodology and criteria that licensees can use to perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific analysis approved by NRC is reflected in the plant’s licensing basis. The methodology described in this section is also an acceptable method of performing a post-fire safe shutdown analysis. This methodology is indicated in Figure 3-1. Other methods acceptable to NRC may also be used. Regardless of the method selected by an individual licensee, the criteria and assumptions provided in this guidance document may apply. The methodology described in Section 3 is based on a computer database oriented approach, which is utilized by several licensees to model Appendix R data relationships. This guidance document, however, does not require the use of a computer database oriented approach.
The requirements of Appendix R Sections III.G.1, III.G.2 and III.G.3 apply to equipment and cables required for achieving and maintaining safe shutdown in any fire area. Although equipment and cables for fire detection and suppression systems, communications systems and 8-hour emergency lighting systems are important features, this guidance document does not address them.
Applicability
Applicable
Comments
Alignment Statement
Aligns with Intent
Alignment Basis PINGP utilized a deterministic methodology to assess conformance of the PINGP SSA with the Nuclear Safety Capability Assessment (NSCA), as described in NFPA 805, Section 1.5.1. This assessment evaluated the current program for readiness to transition, identified gaps and gap closure actions, and culminated in the PINGP’s safe and stable position with respect to the deterministic guidance given in Chapter 3 of NEI 00-01, Revision 1, “Deterministic Methodology.”
Note: A Gap analysis between NEI 00-01 Revision 1 and Revision 2 was performed under EC 23408 and the applicable findings were incorporated in the Nuclear Safety Capability Assessment (NSCA) Genesis model. The following sections of this analysis (taken from Table B-2 of NEI 04-02) provide a detailed comparison of the PINGP deterministic methodology, against the guidance provided by NEI 00-01 Revision 1, Chapter 3 “Deterministic Methodology”.
Reference Documents
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis
NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
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PINGP Page B-3 - Revision 1
The comparison concluded that the PINGP NSCA falls within the following two categories when compared with NEI 00-01 Revision 1, and the NEI 00-01 revision 1 to revision 2 gap analysis conducted under EC23408:
1. Aligns or Aligns with Intent a. May include open items that are not required to be closed to ensure
alignment status. b. May include commitments in Attachment S that are being performed to
assure alignment status.
2. Not in Alignment [but Prior NRC Approval] a. PINGP credits two existing licensing actions (exemptions) to meet the
criteria of Sections 3.1.1.4 and 3.4.1.6.
The PINGP methodology was used to determine if the Nuclear Safety Performance Criteria are being met for maintaining fuel in a safe and stable condition for all modes and applicable plant configurations.
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PINGP Page B-4 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1 Safe Shutdown Systems and Path Development
This section discusses the identification of systems available and necessary to perform the required safe shutdown functions. It also provides information on the process for combining these systems into safe shutdown paths. Appendix R Section III.G.1.a requires that the capability to achieve and maintain hot shutdown be free of fire damage. It is expected that the term “free of fire damage” will be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section III.G.1.b requires that repairs to systems and equipment necessary to achieve and maintain cold shutdown be completed within 72 hours. It is the intent of the NRC that requirements related to the use of manual operator actions will be addressed in a forthcoming rulemaking.
The goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and components remains free of fire damage for a single fire in any single plant fire area. This goal is accomplished by determining those functions important to achieve and maintain hot shutdown. Safe shutdown systems are selected so that the capability to perform these required functions is a part of each safe shutdown path. The functions important to post-fire safe shutdown generally include, but are not limited to the following:
- Reactivity Control - Pressure Control Systems - Inventory Control Systems - Decay Heat Removal Systems - Process Monitoring - Support Systems
Electrical systems Cooling systems
These functions are of importance because they have a direct bearing on the safe shutdown goal of being able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactor pressure vessel, and the primary containment. If these functions are preserved, then the plant will be safe because the fuel, the reactor and the primary containment will not be damaged. By assuring that this equipment is not damaged and remains functional, the protection of the health and safety of the public is assured.
In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated circuits with the potential for spurious equipment operation and/or loss of power source, and the common enclosure failures. Spurious operations/actuations can affect the accomplishment of the post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious operations of concern are the following:
- A loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability
- A flow loss or blockage in the inventory makeup or decay heat removal systems being used for the required safe shutdown path.
Spurious operations are of concern because they have the potential to directly affect the ability to achieve and maintain hot shutdown, which could affect the fuel and cause damage to the reactor pressure vessel or the primary containment. Common power source and common enclosure concerns could also affect these and must be addressed.
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PINGP Page B-5 - Revision 1
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns with Intent Systems, functions, equipment, cables and logics required to maintain fuel in a safe and stable condition have been identified and are included in
the NSCA (Genesis) Model.
Demonstration of the Nuclear Safety Performance Criteria for safe and stable conditions was performed in two analyses. � At-Power analysis, Mode 1 through Mode 3. This analysis is discussed in Section 4.2.4. � Non-Power analysis, which includes Mode 4 and below. This analysis is discussed in Section 4.3. The NFPA 805 licensing basis for PINGP for a Safe and Stable condition in the event of a fire starting with the reactor in at-power operating Modes 1, 2, or 3 (Power Operation, Startup, or Hot Standby, respectively) is to maintain Safe and Stable conditions in Hot Standby without Residual Heat Removal (RHR). PINGP will maintain Hot Standby conditions until a decision is made to either place the reactor in a non-power operating mode, i.e., Hot Shutdown (Mode 4) or Cold Shutdown (Mode 5), or to return to power operations. Determination of the final state will be based upon the extent of the fire damage, the inventory remaining in the Refueling Water Storage Tank (RWST), the ability to provide makeup water to the RWST, and the ability to re-establish inventory in the Condensate Storage Tank (CST) or realign the CST to an alternate source. Mission Time A PINGP thermal-hydraulic analysis was performed for a mission time of 24 hours to assure that safe and stable conditions can be achieved within that time period. This mission time ensures that sufficient time is available for the Emergency Response Organization to respond to the event, assess the extent of fire damage, and assist the plant operating staff with maintaining Safe and Stable conditions or transitioning the plant to a non-power operating mode. To sustain Safe and Stable conditions, Key Safety Functions are met as follows: � Reactivity and Inventory Control The reactor design ensures that Keff < 0.99 can be achieved by use of the control rods from any operating mode. Subsequent injection (using Charging or Safety Injection Pumps) of soluble poison can be used to assure continuation of Mode 3, Hot Standby, under all circumstances. The charging system and the Safety Injection system will remain available beyond the mission time for Safe and Stable. The RWST is the credited source of borated water and is capable of providing water for at least 38
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 EC 23925 – NFPA 805 LAR Supplement – Control Room HVAC Evaluation
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hours, per EC-20736, “Reactivity Control.” Operator actions to establish makeup sources of inventory to the RWST are described in existing plant procedure C12.5, “Boron Concentration Control.” � Decay Heat Removal One or both steam generators, as well as a motor driven or turbine driven Auxiliary Feedwater (AFW) pump will remain available without additional actions to provide symmetrical or asymmetrical decay heat removal beyond the mission time for Safe and Stable. The Condensate Storage Tank (CST) is the initial source for the AFW pumps. Per EC-20738, “Decay Heat Removal,” the CST will provide a continuous water supply for the AFW pumps for 20 hours. Beyond 20 hours, the CST can be refilled or the AFW pumps can be re-aligned to the cooling water system to provide an unlimited water source. This realignment is accomplished through existing plant procedures 1(2)E-1, “Loss of Reactor or Secondary Coolant,” and C28.1, AOP2, “Loss of Condensate Supply to Auxiliary Feedwater Pump Suction.” � Vital Auxiliaries – Power and Support Systems The Emergency Diesel Generators (EDGs) and Diesel Driven Cooling Water Pumps (DDCLPs) have an on-site fuel oil supply that will last for 14 days, assuming one EDG on each unit and one DDCLP, or 7 days if both EDGs are operating for each unit and both DDCLPs are operating. . Offsite sources of fuel oil are available to replenish fuel oil levels if needed via established contracts. Control room HVAC can be lost in several fire areas due to the loss of instrument air and other fire-induced HVAC component damage. EC 23925, “NFPA 805 LAR Supplement – Control Room HVAC Evaluation”, demonstrates that control room temperatures will remain below equipment limits for up to 36 hours with actions taken only within the control room itself. However, a portable fan may be installed in the control room prior to 36 hours to allow temperatures to remain below equipment limits indefinitely. If required, the portable fan will normally be powered by a designated welding receptacle or in cases where the welding receptacle power is lost due to the fire, by a 480VAC portable generator located outside of the building [Reference Attachment S-3 for additional details] If conditions warrant placing the plant in Hot Shutdown (Mode 4) or Cold Shutdown (Mode 5), NSPM will initiate operation of the RHR System. Although the RHR system is not required for maintaining safe and stable conditions, the RHR system is included in the “at power” NSCA Genesis model to demonstrate its availability for transition. Initiation of RHR system operations does not imply that the end state will be Cold Shutdown (Mode 5).
The NSCA (Genesis) model is utilized to demonstrate that the Nuclear Safety Performance Criteria is met on a fire-area-by-fire-area basis by logically correlating the Nuclear Safety Performance Criteria of NFPA 805, Section 1.5.1, to physical plant success paths, equipment, cables, and location (fire areas).
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PINGP Page B-7 - Revision 1
High/low pressure interfaces have been considered and precluded from spurious operation of the pathway via plant configuration change to operate with 480VAC breakers open for RHR suction valves: MV-32231 (U1), MV-32165 (U1), MV-32233 (U2) and MV-32193 (U2). The series counterparts to these valves already have their 480VAC breakers in the open position during power operations. This configuration change will preclude spurious operation of the high / low interface path by removing motive power from both series valves in each path.
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NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1 Criteria/Assumptions The following criteria and assumptions may be considered when identifying systems available and necessary to perform the required safe shutdown functions and combining these systems into safe shutdown paths.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance Only EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
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NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.1 Criteria/Assumptions [BWR] GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths For The BWR" addresses the systems and equipment originally designed into the GE boiling water reactors (BWRs) in the 1960s and 1970s, that can be used to achieve and maintain safe shutdown per Section III.G.1 of 10CFR 50, Appendix R. Any of the shutdown paths (methods) described in this report are considered to be acceptable methods for achieving redundant safe shutdown.
Applicability
Not Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Applicable Not Applicable to PINGP – BWR Specific. EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
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NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.2 Criteria/Assumptions [BWR] GE Report GE-NE-T43-00002-00-03-R01 provides a discussion on the BWR Owners' Group (BWROG) position regarding the use of Safety Relief Valves (SRVs) and low pressure systems (LPCI/CS) for safe shutdown. The BWROG position is that the use of SRVs and low pressure systems is an acceptable methodology for achieving redundant safe shutdown in accordance with the requirements of 10CFR50 Appendix R Sections III.G.1 and III.G.2. The NRC has accepted the BWROG position and issued an SER dated Dec. 12, 2000.
Applicability
Not Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Not Applicable to PINGP – BWR Specific. EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-11 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.3 Criteria/Assumptions [PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the makeup/charging pumps). Hot shutdown conditions can be maintained via natural circulation of the RCS through the steam generators. The cooldown rate must be controlled to prevent the formation of a bubble in the reactor head. Therefore, feedwater (either auxiliary or emergency) flow rates as well as steam release must be controlled.
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Aligns Safe and stable is defined as Mode 3 Hot Standby for PINGP. Pressurizer heaters are not credited for increasing RCS pressure, but can be utilized if the circuits are not affected in a particular fire area. Pressurizer heaters are primarily evaluated for spurious operation concerns. If a pressurizer heater is not available, RCS pressure can be increased by utilizing a water solid pressurizer and controlling RCS makeup and letdown flow."
Spurious operation of pressurizer heaters has been considered in the safe and stable model.
RCS makeup is provided by either charging or SI with suction from the Reactor Water Storage Tank (RWST). The plant is capable of maintaining safe and stable conditions in both a natural or forced circulation mode. See Section 3.1 above for a definition of safe and stable for PINGP.
Feedwater for decay heat removal is provided by auxiliary feedwater (AFW) system (turbine or motor driven). The plant is capable of either symmetrical or asymmetrical cooldown methods. The normal source for AFW is the CST with the cooling water system serving as a source for extended operation. See Section 3.1 above for details.
Adequate cooldown rate is maintained to ensure void formation in the reactor vessel does not occur to ensure performance goals of maintaining pressurizer level within scale and preventing fuel clad damage will be met. Analysis supporting the assumption that pressurizer heaters are not required when SI is credited for RCS makeup is documented in EC 20824.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 EC 20824 - NFPA 805 Thermal Hydraulic Calculation for Subcooling Margin, Owners Acceptance of Vendor Calculation P2117-2400-06-00 Maintaining Subcooling Margin with Failure of Press Htrs
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-12 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.4 Criteria/Assumptions The classification of shutdown capability as alternative shutdown is made independent of the selection of systems used for shutdown. Alternative shutdown capability is determined based on an inability to assure the availability of a redundant safe shutdown path. Compliance to the separation requirements of Sections III.G.1 and III.G.2 may be supplemented by the use of manual actions to the extent allowed by the regulations and the licensing basis of the plant, repairs (cold shutdown only), exemptions, deviations, GL 86-10 fire hazards analyses or fire protection design change evaluations, as appropriate. These may also be used in conjunction with alternative shutdown capability.
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Not in Alignment [but Prior NRC Approval] Fire areas 13 and 18 (Control Room and Relay Room) were defined as “Alternative / Dedicated Shutdown” areas under 10CFR50 Appendix R. Unlike 10CFR50 Appendix R, NFPA 805 does not define areas or scenarios as alternative shutdown or dedicated shutdown.
Fire areas 13 and 18 are being transitioned as risk-informed, performance-based areas with the following high-level details:
• Not all scenarios within fire areas 13 and 18 require control room abandonment.
• Fire PRA employs the use of a hot shutdown panel (HSD) in conjunction with other controls when required to abandon the control room, however, the HSD panels and controls do not meet the definition of a primary control station (PCS) as defined by RG 1.205. Therefore, no credit has been taken for PCS actions in the Fire PRA for these fire areas.
• Actions to enable the HSD panel and establish required controls are listed in the PRA Alternative Shutdown Notebook.
• Just prior to abandoning the control room, control room action is taken on the control board to close the PORV block valves. Additional control room actions are then taken at a switch panel that is being installed within the control room (see Attachment S Item 27) to isolate excess letdown, head vents, pressurizer vents, pressurizer PORV and the pressurizer heaters.
As an exception to this section, PINGP is transitioning existing approved licensing action for a “repair action” to assure isolation of pressurizer PORVs for a fire occurring in the control room or relay room (Fire Areas 13 and 18 respectively), that could cause spurious operation of PORV isolation valves. Therefore, this section is “Not in Alignment but Prior NRC Approval”. The details for this licensing action can be found in Attachments K and T.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 PRA Alternative Shutdown Notebook Attachment S Attachment K Attachment T
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-13 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.5 Criteria/Assumptions At the onset of the postulated fire, all safe shutdown systems (including applicable redundant trains) are assumed operable and available for post-fire safe shutdown. Systems are assumed to be operational with no repairs, maintenance, testing, Limiting Conditions for Operation, etc. in progress. The units are assumed to be operating at full power under normal conditions and normal lineups.
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Comments
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Alignment Basis
Reference Documents
Aligns The PINGP NFPA-805 safe and stable analysis assumes the availability / operability of credited systems at the onset of the fire. This assurance is provided by including appropriate systems / components in the PINGP monitoring program as defined in LAR Section 4.6.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-14 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.6 Criteria/Assumptions No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident, earthquake), single failures or non-fire induced transients need be considered in conjunction with the fire.
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Aligns No accidents or other design basis events, including single failures and non-fire induced transients, are considered in conjunction with the fire.
Reference Table B-1 Section 3.6.4 for how PINGP aligns with the earthquake provisions of NFPA 805.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-15 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.7 Criteria/Assumptions For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of fire damage. Offsite power should be assumed to remain available for those cases where its availability may adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power if the consequences of offsite power availability are more severe than its presumed loss). No credit should be taken for a fire causing a loss of offsite power. For areas where train separation cannot be achieved and alternative shutdown capability is necessary, shutdown must be demonstrated both where offsite power is available and where offsite power is not available for 72 hours.
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Aligns Credit is not taken for off-site power in a specific fire area, unless an analysis is performed to demonstrate the availability of off-site power following the worst-case postulated fire in the area in which all unprotected cables in the area are assumed damaged. The NSCA (Genesis) Model does not credit the loss of off-site power due to the fire.
Distribution components and cables of the credited plant AC and DC systems are modeled in the NSCA (Genesis) model in a cascading fashion. Within the NSCA (Genesis) model, the distribution equipment, source cables and credited load equipment are combined to form a complete cascading success path from all credited sources to load.
All credited sources are analyzed for availability within the NSCA (Genesis) model for the fire area in which they are credited. Additionally, the ability to trip and close alternate source breakers, including the utilization of the load sequencer when applicable, have been analyzed within the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-16 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.8 Criteria/Assumptions Post-fire safe shutdown systems and components are not required to be safety-related.
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Aligns PINGP credits both safety-related and non-safety related equipment to achieve safe and stable conditions.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-17 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.9 Criteria/Assumptions The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within this 72-hour period need not be included in the post-fire safe shutdown analysis. At least one train can be repaired or made operable within 72 hours using onsite capability to achieve cold shutdown.
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Aligns with Intent NFPA 805 Nuclear Safety Performance Criteria (NSPC) requires the licensee to demonstrate that the plant can achieve and maintain a “safe and stable” condition, but it has no explicit requirement to demonstrate that cold shutdown can be achieved.
Equipment required to achieve cold shutdown is included in the NSCA (Genesis) model to demonstrate availability / identify fire damage that may require repair should plant operations decide to proceed to non-power operations from a safe and stable condition. There is no time or power supply requirement associated with completion of repairs under NFPA 805.
Reference Section 3.1 above for PINGP’s definition of Safe and Stable.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-18 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.10 Criteria/Assumptions Manual initiation from the main control room or emergency control stations of systems required to achieve and maintain safe shutdown is acceptable where permitted by current regulations or approved by NRC; automatic initiation of systems selected for safe shutdown is not required but may be included as an option.
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Aligns with Intent Automatic functions (e.g., load sequencer), when credited, have been analyzed within the NSCA (Genesis) model for availability and spurious operation. Manual functions are credited by the analysis and are listed in the fire risk evaluations FREs (Recovery Actions) and in the non-power operations NPO (manual actions) EC 20612.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 FRE ECs (Many)
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-19 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.1.11 Criteria/Assumptions Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and maintain safe shutdown for each affected unit must be demonstrated.
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Aligns The NSCA (Genesis) model evaluates the effects on both PINGP units, of a single fire occurring in a single fire area, for all analyzed areas in the plant. This methodology assures that fuel in both units is maintained in a safe and stable condition.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-20 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2 Shutdown Functions The following discussion on each of these shutdown functions provides guidance for selecting the systems and equipment required for safe shutdown. For additional information on BWR system selection, refer to GE Report GENE-T43-00002-00- 01-R01 entitled "Original Safe Shutdown Paths for the BWR."
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Applicable
Comments
Alignment Statement
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Reference Documents
Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-20 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.1 Reactivity Control [BWR] Control Rod Drive System
The safe shutdown performance and design requirements for the reactivity control function can be met without automatic scram/trip capability. Manual scram/reactor trip is credited. The post-fire safe shutdown analysis must only provide the capability to manually scram/trip the reactor.
[PWR] Makeup/Charging
There must be a method for ensuring that adequate shutdown margin is maintained by ensuring borated water is utilized for RCS makeup/charging.
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Aligns with Intent Long-term reactivity control is accomplished by adding borated water from the refueling Water Storage Tank (RWST) to ensure reactivity margin throughout safe and stable.
RWST is supplied to the RCS through either the Chemical and Volume Control System (CVCS) or the Safety Injection System (SI).
The NSCA (Genesis) model has shown that the credible boron dilution event (MSO) is only possible for fires occurring in Fire Areas 13 and 18. This event will be addressed as part of the RI / PB methods for these areas.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-21 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.2 Pressure Control Systems [BWR] Safety Relief Valves (SRVs)
The SRVs are opened to maintain hot shutdown conditions or to depressurize the vessel to allow injection using low pressure systems. These are operated manually. Automatic initiation of the Automatic Depressurization System is not a required function.
[PWR] Makeup/Charging
RCS pressure is controlled by controlling the rate of charging/makeup to the RCS. Although utilization of the pressurizer heaters and/or auxiliary spray reduces operator burden, neither component is required to provide adequate pressure control. Pressure reductions are made by allowing the RCS to cool/shrink, thus reducing pressurizer level/pressure. Pressure increases are made by initiating charging/makeup to maintain pressurizer level/pressure. Manual control of the related pumps is acceptable.
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Aligns RCS Pressure Control is necessary to prevent exceeding RCS design pressure-temperature limits and to minimize void formation in the reactor. RCS pressure control is also necessary to control high/low pressure interfaces and to prevent rupture of any primary coolant boundary. Some key goals of pressure control are: • Preventing a LOCA due to spurious operation of high/low pressure interface components • Isolating Normal Pressurizer and Auxiliary Spray • Closing or isolating the Pressurizer Power-Operated Relief Valves (PORVs) • Isolating RCS head vent flowpaths • Control of Normal and Excess Letdown flowpaths • Securing the Pressurizer Heaters when spuriously operating
Over-pressure protection of the RCS is mainly provided by the pressurizer safety valves (RC-10-1, RC-10-2, 2RC-10-1, 2RC-10-2). The power- operated relief valves and associated block valve, RCS head vent valves, and Pressurizer vent valves can be used for over-pressure protection. Low temperature overpressure protection mitigates an overpressure event during low temperature operation. The reactor head vent valves are credited for low temperature overpressure protection.
Normal pressurizer spray and auxiliary spray are also methods used to depressurize the RCS; however, these flow paths are not credited for safe
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-22 - Revision 1
and stable and are only evaluated for spurious operation concerns.
Safe and stable conditions can be achieved and maintained by controlling the RCS makeup through the charging path (charging pumps or safety injection pumps) and letdown/leakage flows. Because normal and excess letdown flowpaths may not be available, the RCS head vents or pressurizer head vents can be throttled for letdown. In addition, RCP seal leakage flow is also another means for letdown but is not explicitly credited by the safe and stable analysis.
In addition to controlling the rate of charging/makeup for pressure increases, PINGP safe and stable credits cooling and shrinking of the RCS through steam heat removal; this is accomplished by controlling the flow of auxiliary feedwater to the steam generators and discharging the volume as steam via the safety valves, steam diversion flow paths and the steam generator power-operated relief valves (PORVs).
Pressurizer heaters and sprays have been analyzed for spurious operation in the NSCA (Genesis) model.
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-23 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.3 Inventory Control [BWR] Systems selected for the inventory control function should be capable of supplying sufficient reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.
[PWR] Systems selected for the inventory control function should be capable of maintaining level to achieve and maintain hot shutdown. Typically, the same components providing inventory control are capable of providing pressure control. Manual initiation of these systems is acceptable. Automatic initiation functions are not required.
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Aligns with Intent To ensure fuel is maintained in a safe and stable condition, RCS makeup is required for maintenance of RCS integrity (isolating flow diversion paths and maintaining the pressurizer level within the indicating range), and to compensate for RCS inventory losses due to seal leakoff and shrinkage during cooldown. RCS makeup is provided by the Chemical and Volume Control System (CVCS) or the Safety Injection System (SI). Typically, the CVCS system is credited if Train A systems are available, and the SI system is credited if Train B systems are available. As an example, the NSCA (Genesis) model includes analysis of the following in support of inventory control: • Isolation of the Reactor Coolant System diversion flowpaths. • Alignment of RWST to either charging pump suction or SI Pump suction. • Ensuring RCP seal cooling to prevent a small break LOCA (Reference Attachment S Tables S-1, S-2 for seal modification). • Operation of the charging system or SI system. • Maintaining Reactor Coolant Pump seal return flowpath. • Isolating RWST diversion flowpaths to the Containment Spray and RHR system. • De-energizing non-credited components that may affect safe and stable.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-24 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.4 Decay Heat Removal [BWR] Systems selected for the decay heat removal function(s) should be capable of:
- Removing sufficient decay heat from primary containment, to prevent containment over-pressurization and failure.
- Satisfying the net positive suction head requirements of any safe shutdown systems taking suction from the containment (suppression pool).
- Removing sufficient decay heat from the reactor to achieve cold shutdown.
This does not restrict the use of other systems.
[PWR] Systems selected for the decay heat removal function(s) should be capable of:
- Removing sufficient decay heat from the reactor to reach hot shutdown conditions. Typically, this entails utilizing natural circulation in lieu of forced circulation via the reactor coolant pumps and controlling steam release via the Atmospheric Dump valves.
- Removing sufficient decay heat from the reactor to reach cold shutdown conditions.
This does not restrict the use of other systems.
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Aligns with Intent Following a reactor trip, decay heat is removed via the steam generators by natural or forced circulation within the reactor coolant loops. Feedwater is supplied to the steam generators by the Auxiliary Feedwater (AFW) System. At a minimum, one AFW pump and one steam generator is required for removal of decay heat from the RCS.
Removing sufficient decay heat to reach cold shutdown conditions is not a requirement for PINGP to maintain fuel in a safe and stable condition. PINGP defines safe and stable as Mode 3 (reference section 3.1 above for further details of PINGP safe and stable).
Decay heat safety function has been incorporated in the NSCA (Genesis) model with all available success paths. The NSCA (Genesis) model analyzes for availability as well as spurious operation.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-25 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.5 Process Monitoring The process monitoring function is provided for all safe shutdown paths. IN 84-09, Attachment 1, Section IX "Lessons Learned from NRC Inspections of Fire Protection Safe Shutdown Systems (10CFR50 Appendix R)" provides guidance on the instrumentation acceptable to and preferred by the NRC for meeting the process monitoring function. This instrumentation is that which monitors the process variables necessary to perform and control the functions specified in Appendix R Section III.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 list of process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specific instruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). In general, process monitoring instruments similar to those listed below are needed to successfully use existing operating procedures (including Abnormal Operating Procedures).
PWR:
- Reactor coolant temperature (hot leg / cold leg) - Pressurizer pressure and level - Neutron flux monitoring (source range) - Level indication for tanks needed for safe shutdown - Steam generator level and pressure - Diagnostic instrumentation for safe shutdown systems
The specific instruments required may be based on operator preference, safe shutdown procedural guidance strategy (symptomatic vs. prescriptive), and systems and paths selected for safe shutdown.
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Aligns with Intent PINGP has modeled the required process monitoring functions in the NSCA (Genesis) model.
The performance of the primary system (RCS) is monitored by RCS pressure, pressurizer level, and RCS Hot Leg and Cold Leg Temperatures. The secondary system (AFW) is monitored by the SG level indicators.
Reactivity is monitored by the neutron source range flux monitors.
When utilizing the hot shutdown panels (HSD), steam generator pressure is available via local mechanical (remote from HSD panel) indicators. Additionally, when controlling charging from the HSD panel, local pressure indication is credited (Unit 1: 1LI-433C Unit 2: 2LI-433C).
Additional tank level monitoring is provided by local mechanical level
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-26 - Revision 1
indicators. Local tank level indicators include: Unit 1 and Unit 2 Refueling Water Storage Tanks Unit 1 and Unit 2 Condensate Storage Tanks 12 and 22 Diesel Driven Cooling Water Pump Fuel Oil Day Tank
Diagnostic instrumentation is generally provided by local devices that do not require power to operate. If power or control was required for the diagnostic instrument, it was included in the model for analysis purposes.
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-27 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.6.1 Electrical Systems AC Distribution System
Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage system such as 4.16 KV Class 1E buses either directly from the buses or through step down transformers/ load centers/ distribution panels for 600, 480 or 120 VAC loads.
For redundant safe shutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, power may be supplied from either offsite power sources or the emergency diesel generator depending on which has been demonstrated to be free of fire damage. No credit should be taken for a fire causing a loss of offsite power. Refer to Section 3.1.1.7.
DC Distribution System
Typically, the 125VDC distribution system supplies DC control power to various 125VDC control panels including switchgear breaker controls. The 125VDC distribution panels may also supply power to the 120VAC distribution panels via static inverters. These distribution panels typically supply power for instrumentation necessary to complete the process monitoring functions.
For fire events that result in an interruption of power to the AC electrical bus, the station batteries are necessary to supply any required control power during the interim time period required for the diesel generators to become operational.
Once the diesels are operational, the 125 VDC distribution system can be powered from the diesels through the battery chargers.
The DC control centers may also supply power to various small horsepower Appendix R safe shutdown system valves and pumps. If the DC system is relied upon to support safe shutdown without battery chargers being available, it must be verified that sufficient battery capacity exists to support the necessary loads for sufficient time (either until power is restored, or the loads are no longer required to operate).
Aligns Distribution components of the credited plant AC and DC systems are
modeled in the NSCA (Genesis) model in a cascading fashion. Within the
NSCA (Genesis) model, individual source cables and equipment are combined to form a complete cascading success path from source to load. Additionally, the NSCA (Genesis) model provides the means to analyze a loss of DC control power to the credited AC buses.
Upon loss of AC power to the battery chargers, the batteries are credited to supply power to interim loads, including diesel generator startup loads, until an AC source is re-established to the credited chargers.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-28 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.2.6.2 Cooling Systems Various cooling water systems may be required to support safe shutdown system operation, based on plant-specific considerations. Typical uses include:
HVAC Systems may be required to assure that safe shutdown equipment remains within its operating temperature range, as specified in manufacturer’s literature or demonstrated by suitable test methods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxic gases, and gaseous fire suppression agents). HVAC systems may be required to support safe shutdown system operation, based on plant-specific configurations. Typical uses include:
- Main control room, cable spreading room, relay room - ECCS pump compartments - Diesel generator rooms - Switchgear rooms
Plant-specific evaluations are necessary to determine which HVAC systems are essential to safe shutdown equipment operation.
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Aligns HVAC Systems have been demonstrated to be not required to support the Nuclear Safety Performance Criteria in the NSCA (Genesis) model. Component Cooling, seal injection, seal barrier cooling, cooling to heat exchangers as well as HVAC cooling of components have been considered for safe and stable.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 EC 23925 – NFPA 805 LAR Supplement - Control Room HVAC Evaluation
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-30 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.3 Methodology for Shutdown System Selection
Refer to NEI-00-01 Rev 1 Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown systems and developing the shutdown paths.
The following methodology may be used to define the safe shutdown systems and paths for an Appendix R analysis:
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-31 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.3.1 Identify Safe Shutdown Functions Review available documentation to obtain an understanding of the available plant systems and the functions required to achieve and maintain safe shutdown.
Documents such as the following may be reviewed: - Operating Procedures (Normal, Emergency, Abnormal) - System descriptions - Fire Hazard Analysis - Single-line electrical diagrams - Piping and Instrumentation Diagrams (P&IDs) - [BWR] GE Report GE-NE-T43-00002-00-01-R02 entitled "Original Shutdown Paths for the BWR"
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Comments
Alignment Statement
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Aligns Consistent with the guidance of this section of NEI 00-01, Revision 1, the following input documents were utilized in the development of the PINGP NSCA (Genesis) model:
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-32 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.3.2 Identify Combinations of Systems That Satisfy Each Safe Shutdown Function
Given the criteria/assumptions defined in Section 3.1.1, identify the available combinations of systems capable of achieving the safe shutdown functions of reactivity control, pressure control, inventory control, decay heat removal, process monitoring and support systems such as electrical and cooling systems (refer to Section 3.1.2). This selection process does not restrict the use of other systems. In addition to achieving the required safe shutdown functions, consider spurious operations and power supply issues that could impact the required safe shutdown function.
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Aligns PINGP utilized the methodology of NEI 00-01 Revision 1, Section 3.1.1 and Section 3.1.2, coupled with procedure EM 3.4.3, “Safe Shutdown Circuit Analysis” to establish the systems, components, cables and functions to satisfy the Nuclear Safety Performance Criteria NSPC as defined in NFPA 805 Section 1.5.1.
The NSCA (Genesis) model is used to analyze the post-fire impact of spurious operations and power supply issues that can affect the safe and stable conditions of the plant. Additionally, the NSCA (Genesis) model presents the combination of systems and paths in a logic tree fashion. Attachment S, Table S-3, Item 51 identifies an implementation action to incorporate the applicable details of the vendor cable selection / circuit analysis procedure (EPM-DP-EP-004) into the plant circuit analysis procedure EM 3.4.3. This action will enhance the plant documentation.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 EM 3.4.3 S-3 Item 51
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-33 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.3.3 Define Combination of Systems for Each Safe Shutdown Path
Select combinations of systems with the capability of performing all of the required safe shutdown functions and designate this set of systems as a safe shutdown path. In many cases, paths may be defined on a divisional basis since the availability of electrical power and other support systems must be demonstrated for each path. During the equipment selection phase, identify any additional support systems and list them for the appropriate path.
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Aligns PINGP utilizes the NSCA (Genesis) model to maintain the logical relationships between the components and paths that make up the required NSPC function. Additionally, supporting systems were identified and included in the logical relationship when required. Power supplies are modeled in a cascading fashion such that a loss of an upstream supply will affect all downstream supplies and credited equipment serviced by those supplies.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-34 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.1.3.4 Assign Shutdown Paths to Each Combination of Systems
Assign a path designation to each combination of systems. The path will serve to document the combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to this document for an example of a table illustrating how to document the various combinations of systems for selected shutdown paths.
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Alignment Statement
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Aligns The NSCA (Genesis) model assigns identifiers to each success path. The success paths are grouped by NSPC function within each analysis area within the model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-35 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2 Safe Shutdown Equipment Selection
The previous section described the methodology for selecting the systems and paths necessary to achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for identifying the specific safe shutdown equipment necessary for the systems to perform their Appendix R function.
The selected equipment should be related back to the safe shutdown systems that they support and be assigned to the same safe shutdown path as that system. The list of safe shutdown equipment will then form the basis for identifying the cables necessary for the operation or that can cause the maloperation of the safe shutdown systems.
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Comments
Alignment Statement
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Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-36 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1 Criteria/Assumptions Consider the following criteria and assumptions when identifying equipment necessary to perform the required safe shutdown functions:
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Comments
Alignment Statement
Alignment Basis
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Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-37 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.1 Criteria/Assumptions Safe shutdown equipment can be divided into two categories. Equipment may be categorized as (1) primary components or (2) secondary components. Typically, the following types of equipment are considered to be primary components:
- Pumps, motor operated valves, solenoid valves, fans, gas bottles, dampers, unit coolers, etc.
- All necessary process indicators and recorders (i.e., flow indicator, temperature indicator, turbine speed indicator, pressure indicator, level recorder)
- Power supplies or other electrical components that support operation of primary components (i.e., diesel generators, switchgear, motor control centers, load centers, power supplies, distribution panels, etc.).
Secondary components are typically items found within the circuitry for a primary component. These provide a supporting role to the overall circuit function. Some secondary components may provide an isolation function or a signal to a primary component via either an interlock or input signal processor. Examples of secondary components include flow switches, pressure switches, temperature switches, level switches, temperature elements, speed elements, transmitters, converters, controllers, transducers, signal conditioners, hand switches, relays, fuses and various instrumentation devices.
Determine which equipment should be included on the Safe Shutdown Equipment List (SSEL). As an option, include secondary components with a primary component(s) that would be affected by fire damage to the secondary component. By doing this, the SSEL can be kept to a manageable size and the equipment included on the SSEL can be readily related to required post-fire safe shutdown systems and functions.
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Aligns with Intent PINGP did not differentiate between primary and secondary components within the NSCA (Genesis) model however, all analysis performed in support of added equipment for NFPA 805 included the secondary components, or at a minimum, the function of the secondary components within the model.
For example: a pump that is required to meet a NSPC, would appear as a component on the logic path for that function. However, the mechanical oil pressure switch that stops the pump on low oil pressure may not appear on the logic path, but the function of the pressure switch would be included in the circuit analysis considerations for the pump.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-38 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.2 Criteria/Assumptions Assume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing with brazed or soldered joints, are not included in this assumption). Fire damage should be evaluated with respect to the ability to manually open or close the valve should this be necessary as a part of the post-fire safe shutdown scenario.
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Aligns PINGP assumes no fire-induced damage to manual valves or piping. Additionally, NSPM determined that no valves are subject to the rising stem concern as described in NEI 00-01 Revision 2 (reference EC 23408, NEI 00-01 gap analysis).
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-39 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.3 Criteria/Assumptions Assume that manual valves are in their normal position as shown on P&IDs or in the plant operating procedures.
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Aligns For safe and stable, manual valves are assumed to be in their normal operating positions per their respective plant documentation.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-40 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.4 Criteria/Assumptions Assume that a check valve closes in the direction of potential flow diversion and seats properly with sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the flow rate capability of the safe shutdown systems being used for inventory control, decay heat removal, equipment cooling or other related safe shutdown functions.
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Aligns PINGP assumes check valves are installed properly in that they will prevent reversal of flow. PINGP also assumes that the check valve’s integrity is such that they will not produce a leak rate that is other than inconsequential.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-41 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.5 Criteria/Assumptions Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flow transmitters) are assumed to fail upscale, midscale, or downscale as a result of fire damage, whichever is worse. An instrument performing a control function is assumed to provide an undesired signal to the control circuit.
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Aligns with Intent PINGP assumes that instrumentation circuits fail in their worst-case positions when damaged by the fire unless analysis was performed to show that the failure mode is incredible.
Credited instrumentation circuits are modeled in the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-42 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.6 Criteria/Assumptions Identify equipment that could spuriously operate or mal-operate and impact the performance of equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1 of RIS 2004-03 during the equipment identification process.
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Aligns Spurious operation was considered in the selection of components as well as cable selection. The selected components (including logical relationships) and cables are maintained in the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-43 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.1.7 Criteria/Assumptions Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a result of fire. Determine and consider the fire area location of the instrument tubing when evaluating the effects of fire damage to circuits and equipment in the fire area.
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Aligns An instrument tubing analysis has been performed for credited instrumentation along the credited success paths (including credited process monitoring). The analysis did not identify any impacts to the safe and stable model resulting from fire-induced effects on liquid-filled instrument tubing.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-44 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2 Methodology for Equipment Selection
Refer to NEI-00-01 Rev 1 Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdown equipment.
Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdown analysis:
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Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-45 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2.1 Identify the System Flow Path for Each Shutdown Path
Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept.
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Aligns with Intent PINGP utilized marked-up P&IDs to identify both required flow paths as well as diversion flow paths. This information was used as a basis for the safe and stable logic trees and equipment listings that are contained in the NSCA (Genesis) model. The marked up P&IDs are not required to be retained as the logic trees within the NSCA (Genesis) model are sufficient to show flowpaths with credited components.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-46 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2.2 Identify the Equipment in Each Safe Shutdown System Flow Path Including Equipment That May Spuriously Operate and Affect System Operation
Review the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to assure that all equipment in each system’s flow path has been identified. Assure that any equipment that could spuriously operate and adversely affect the desired system function(s) is also identified. If additional systems are identified which are necessary for the operation of the safe shutdown system under review, include these as systems required for safe shutdown. Designate these new systems with the same safe shutdown path as the primary safe shutdown system under review (Refer to Figure 3-1).
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Aligns PINGP utilized P&IDs, single line diagrams, loop diagrams, connection diagrams and procedures to identify equipment required to meet NSPC. Spurious operation of equipment that could affect the NSPC was considered in the selection process.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-47 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2.3 Develop a List of Safe Shutdown Equipment and Assign the corresponding System and Safe Shutdown Path(s) Designation to Each
Prepare a table listing the equipment identified for each system and the shutdown path that it supports. Identify any valves or other equipment that could spuriously operate and impact the operation of that safe shutdown system.
Assign the safe shutdown path for the affected system to this equipment. During the cable selection phase, identify additional equipment required to support the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe shutdown and it documents various equipment-related attributes used in the analysis.
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Aligns The NSCA (Genesis) model contains the listing of the components that were identified as being required to meet the NSPC.
In addition to the equipment listing, shutdown paths are displayed logically within the model. Electrical distribution equipment is logically arranged in a cascading manner such that a failure of an upstream component will cascade down to show a loss of the downstream components. Other supporting equipment is logically tied to the component(s) or functions which it supports.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-48 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2.4 Identify Equipment Information Required for the Safe Shutdown Analysis
Collect additional equipment-related information necessary for performing the post-fire safe shutdown analysis for the equipment. In order to facilitate the analysis, tabulate this data for each piece of equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL. Examples of related equipment data should include the equipment type, equipment description, safe shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of equipment. Other information such as the following may be useful in performing the safe shutdown analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed electrical position, high/low pressure interface concern, and spurious operation concern.
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Aligns with Intent PINGP has identified and recorded similar data to that given in this guidance. PINGP maintains the NSCA equipment listing and analysis data within the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-49 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.2.2.5 Identify Dependencies Between Equipment, Supporting Equipment, Safe Shutdown Systems and Safe Shutdown Paths
In the process of defining equipment and cables for safe shutdown, identify additional supporting equipment such as electrical power and interlocked equipment. As an aid in assessing identified impacts to safe shutdown, consider modeling the dependency between equipment within each safe shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these relationships.
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Aligns PINGP utilizes the NSCA (Genesis) model to provide logical relationships between equipment, cables, power supplies and supporting equipment, to the fire areas in which they are located. The NSCA (Genesis) model provides the necessary logical relationships to allow analysis of fire- induced failures on a fire-area-by-fire-area basis. The logical relationships provide for proper cascading of equipment losses from an upstream component to the downstream components.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety functions shall be identified. This includes circuits that are required for operation, that could prevent the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluation shall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and shorts-to-ground, to identify circuits that are required to support the proper operation of components required to achieve the nuclear safety performance criteria, including spurious operation and signals. This will ensure that a comprehensive population of circuitry is evaluated.
2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria.
(a) Common Power Supply Circuits. Those circuits whose fire induced failure could cause the loss of a power supply required to achieve the nuclear safety performance criteria shall be identified. This situation could occur if the upstream protection device (i.e. breaker or fuse) is not properly coordinated with the downstream protection device.
(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve the nuclear safety performance criteria and whose fire-induced failure could cause the loss of the required components shall be identified. The concern is that the effects of a fire can extend outside of the immediate fire area due to fire-induced electrical faults on inadequately protected cables or via inadequately sealed fire area boundaries.
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3 Safe Shutdown Cable Selection and Location
This section provides industry guidance on the recommended methodology and criteria for selecting safe shutdown cables and determining their potential impact on equipment required for achieving and maintaining safe shutdown of an operating nuclear power plant for the condition of an exposure fire. The Appendix R safe shutdown cable selection criteria are developed to ensure that all cables that could affect the proper operation or that could cause the maloperation of safe shutdown equipment are identified and that these cables are properly related to the safe shutdown equipment whose functionality they could affect. Through this cable-to-equipment relationship, cables become part of the safe shutdown path assigned to the equipment affected by the cable.
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Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-51 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1 Criteria/Assumptions To identify an impact to safe shutdown equipment based on cable routing, the equipment must have cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment so that impacts from these cables can be properly assessed in terms of their ultimate impact on safe shutdown system equipment.
Consider the following criteria when selecting cables that impact safe shutdown equipment:
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Comments
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Not Required Generic Guidance Only
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-52 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.1 Criteria/Assumptions The list of cables whose failure could impact the operation of a piece of safe shutdown equipment includes more than those cables connected to the equipment. The relationship between cable and affected equipment is based on a review of the electrical or elementary wiring diagrams. To assure that all cables that could affect the operation of the safe shutdown equipment are identified, investigate the power, control, instrumentation, interlock, and equipment status indication cables related to the equipment. Consider reviewing additional schematic diagrams to identify additional cables for interlocked circuits that also need to be considered for their impact on the ability of the equipment to operate as required in support of post-fire safe shutdown. As an option, consider applying the screening criteria from Section 3.5 as a part of this section. For an example of this see Section 3.3.1.4.
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Aligns PINGP utilized schematic and connection diagrams to identify all cables associated with the component being analyzed. The analysis included cables from both on-scheme and off-scheme categories when adequate circuit isolation could not be shown to exist.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-53 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.2 Criteria/Assumptions In cases where the failure (including spurious actuations) of a single cable could impact more than one piece of safe shutdown equipment, include the cable with each piece of safe shutdown equipment.
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Aligns The NSCA (Genesis) model maintains cable to equipment relationships with the ability to have a one-to-many relationship. In other words, cables appearing in more than one component’s schematic are captured by the circuit analysis, and are included in the NSCA (Genesis) model under each component that may be affected by the potential fault on that cable.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-54 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.3 Criteria/Assumptions Electrical devices such as relays, switches and signal resistor units are considered to be acceptable isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in the loop to determine that an acceptable isolation device has been installed at each point where the loop must be isolated so that a fault would not impact the performance of the safe shutdown instrument function.
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Aligns The NSCA (Genesis) model credits only acceptable (positive) isolation devices including isolation devices for instrumentation signals that are credited for use in multiple locations of the plant.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-55 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.4 Criteria/Assumptions Screen out cables for circuits that do not impact the safe shutdown function of a component (i.e., annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these circuits is necessary. However, they must be isolated from the component’s control scheme in such a way that a cable fault would not impact the performance of the circuit.
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Aligns Cables were only excluded when proper isolation existed, and reliance on the excluded circuit(s) was not required by the analysis.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-56 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.5 Criteria/Assumptions For each circuit requiring power to perform its safe shutdown function, identify the cable supplying power to each safe shutdown and/or required interlock component. Initially, identify only the power cables from the immediate upstream power source for these interlocked circuits and components (i.e., the closest power supply, load center or motor control center). Review further the electrical distribution system to capture the remaining equipment from the electrical power distribution system necessary to support delivery of power from either the offsite power source or the emergency diesel generators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to the safe shutdown equipment list. Evaluate the power cables for this additional equipment for associated circuits concerns.
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Aligns Power supply logics have been created within the NSCA (Genesis) model to identify the cascading effects of a loss of upstream power supplies. Power supplies are cascaded from the offsite power sources and / or the emergency diesel generator. The effects of loss of DC control power (common power supply, common enclosure) have also been incorporated into the analysis for the applicable buses. Discrepancies identified during the model strategy reviews were dispositioned via the VFDR process.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-57 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.6 Criteria/Assumptions The automatic initiation logics for the credited post-fire safe shutdown systems are not required to support safe shutdown. Each system can be controlled manually by operator actuation in the main control room or emergency control station. If operator actions outside the MCR are necessary, those actions must conform to the regulatory requirements on manual actions. However, if not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits must not adversely affect any post-fire safe shutdown system function.
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Aligns For safe and stable, PINGP does not credit automatic functions unless circuit analysis supports the availability of the function, along with the required supporting equipment and power.
The NSCA (Genesis) model was utilized to determine the effects of the fire on automatic functions for both availability (when credited) and for effects of spurious operation. Discrepancies identified during the model strategy reviews were dispositioned via the VFDR process.
Operator actions taken outside of the control room are addressed in LAR Attachment G. The VFDR process was used to identify those actions that are not allowed under the deterministic methodology of NFPA 805.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-58 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.1.7 Criteria/Assumptions Cabling for the electrical distribution system is a concern for those breakers that feed associated circuits and are not fully coordinated with upstream breakers. With respect to electrical distribution cabling, two types of cable associations exist. For safe shutdown considerations, the direct power feed to a primary safe shutdown component is associated with the primary component. For example, the power feed to a pump is necessary to support the pump. Similarly, the power feed from the load center to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordination is not provided, the same cables discussed above would also support the power supply. For example, the power feed to the pump discussed above would support the bus from which it is fed because, for the case of a common power source analysis, the concern is the loss of the upstream power source and not the connected load. Similarly, the cable feeding the MCC from the load center would also be necessary to support the load center.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns Loss of DC control power to NFPA 805 credited plant buses is incorporated in the NSCA (Genesis) model, and the impact of secondary fires and loss of power supply have been considered (common power supply, common enclosure). Logics contained in the NSCA (Genesis) model detail the cascading effects of power supply losses when they occur as a result of fire-induced damage. Additionally, coordination concerns identified during the coordination review were incorporated in the NSCA (Genesis) model. Discrepancies identified during the model strategy reviews were dispositioned via the VFDR process.
An update to Attachment S Table S-3 was added to track the results of AR 01342798 that identified the need to modify the 4kV fault current study ENG-EE-177 to properly reflect the plant lineups to meet NFPA 805. AR 01342798 identifies that ENG-EE-177 is overly conservative with respect to NFPA 805 requirements. This pending change to ENG-EE-177 only impacts the NFPA 805 analysis for buses 11, 12, 21 and 22; none of which are credited by the NSCA or PRA models for powering equipment.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 ENG-EE-177 Attachment S Table S-3 AR 01342798
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-59 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.2 Associated Circuit Cables Associated Circuit Cables Appendix R, Section III.G.2, requires that separation features be provided for equipment and cables, including associated non-safety circuits that could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts-to-ground, of redundant trains of systems necessary to achieve hot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and further clarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference 6.1.6. They are as follows:
- Spurious actuations - Common power source - Common enclosure
Cables Whose Failure May Cause Spurious Actuations Safe shutdown system spurious actuation concerns can result from fire damage to a cable whose failure could cause the spurious actuation/ maloperation of equipment whose operation could affect safe shutdown. These cables are identified in Section 3.3.3 together with the remaining safe shutdown cables required to support control and operation of the equipment.
Common Power Source Cables The concern for the common power source associated circuits is the loss of a safe shutdown power source due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on a non-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordination between the upstream supply breaker/fuse feeding the safe shutdown power source and the load breaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdown bus. This would result in the loss of power to the safe shutdown equipment supplied from that power source preventing the safe shutdown equipment from performing its required safe shutdown function. Identify these cables together with the remaining safe shutdown cables required to support control and operation of the equipment. Refer to Section 3.5.2.4 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.
Common Enclosure Cables The concern with common enclosure associated circuits is fire damage to a cable whose failure could propagate to other safe shutdown cables in the same enclosure either because the circuit is not properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result in ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This fire spread to an adjacent fire area could impact safe shutdown equipment in that fire area, thereby resulting in a condition that exceeds the criteria and assumptions of this methodology (i.e., multiple fires). Refer to Section 3.5.2.5 for an acceptable methodology for analyzing the impact of these cables on post-fire safe shutdown.
Applicability
Applicable
Comments
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
Not Required Generic Guidance Only EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-61 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3 Methodology for Cable Selection and Location
Refer to Figure 3-4 for a flowchart illustrating the various steps involved in selecting the cables necessary for performing a post-fire safe shutdown analysis. Use the following methodology to define the cables required for safe shutdown including cables that may cause associated circuits concerns for a post-fire safe shutdown analysis:
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance Only EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-62 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3.1 Identify Circuits Required for the Operation of the Safe Shutdown Equipment
For each piece of safe shutdown equipment defined in section 3.2, review the appropriate electrical diagrams including the following documentation to identify the circuits (power, control, instrumentation) required for operation or whose failure may impact the operation of each piece of equipment:
For electrical power distribution equipment such as power supplies, identify any circuits whose failure may cause a coordination concern for the bus under evaluation. If power is required for the equipment, include the closest upstream power distribution source on the safe shutdown equipment list. Through the iterative process described in Figures 3-2 and 3-3, include the additional upstream power sources up to either the offsite or the emergency power source.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP safe and stable circuits were identified using schematics, connection diagrams, single lines and loop diagrams. Logic diagrams detailing the cascade power sources to equipment are included in the NSCA (Genesis) model. Loss of DC control power to power operated circuit breakers is addressed via the OCT analysis in the NSCA (Genesis) model. Discrepancies identified during the model strategy reviews were dispositioned via the VFDR process.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-63 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3.2 Identify Interlocked Circuits and Cables Whose Spurious Operation or Maloperation Could Affect Shutdown
In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes, cables and equipment. Assign to the equipment any cables for interlocked circuits that can affect the equipment.
While investigating the interlocked circuits, additional equipment or power sources may be discovered. Include these interlocked equipment or power sources in the safe shutdown equipment list (refer to NEI-00-01 Rev 1 Figure 3-3) if they can impact the operation of the equipment under consideration.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP circuit analysis considered interlocked, and required supporting equipment.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-64 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3.3 Assign Cables to the Safe Shutdown Equipment
Given the criteria/assumptions defined in Section 3.3.1, identify the cables required to operate or that may result in maloperation of each piece of safe shutdown equipment.
Tabulate the list of cables potentially affecting each piece of equipment in a relational database including the respective drawing numbers, their revision and any interlocks that are investigated to determine their impact on the operation of the equipment. In certain cases, the same cable may support multiple pieces of equipment. Relate the cables to each piece of equipment, but not necessarily to each supporting secondary component.
If adequate coordination does not exist for a particular circuit, relate the power cable to the power source. This will ensure that the power source is identified as affected equipment in the fire areas where the cable may be damaged.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns A listing of required cables from the existing Appendix R model was obtained and combined with circuit analysis that was created in support of NFPA-805. This information was entered into the NSCA (Genesis) model for analysis of safe and stable conditions under NFPA-805. The NSCA (Genesis) model relates cable to equipment, cable to raceway, and raceway to location for analysis purposes. Power supplies are modeled in the NSCA (Genesis) model in cascading fashion. Logic diagrams within the model display the potential success paths. All discrepancies identified during the strategy reviews performed using the model, were dispositioned via the VFDR process.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-65 - Revision 1
2.4.2.3 Nuclear Safety Equipment and Cable Location (Taken From NFPA 805, 2001 Edition)
Physical location of equipment and cables shall be identified.
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3.4 Identify Routing of Cables Identify the routing for each cable including all raceway and cable endpoints. Typically, this information is obtained from joining the list of safe shutdown cables with an existing cable and raceway database.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns Cable routing information was obtained from the plant Passport database system. This information was entered into the NSCA (Genesis) model in support of the safe and stable analysis under NFPA-805. Discrepancies discovered during the transition process were corrected using the established plant processes.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-66 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.3.3.5 Identify Location of Raceway and Cables by Fire Area
Identify the fire area location of each raceway and cable endpoint identified in the previous step and join this information with the cable routing data. In addition, identify the location of field-routed cable by fire area. This produces a database containing all of the cables requiring fire area analysis, their locations by fire area, and their raceway.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP utilizes the Genesis database to track cable to equipment, cable to raceway, and raceway to location information. As stated above, the routing information was obtained from the Passport system. When routing was determined to be missing or inadequate, raceway drawings and walkdowns were utilized to generate the correct information. The information tracked by Genesis is consistent with the guidance of NEI 00-01, Revision 1 for safe and stable and is utilized by the NSCA (Genesis) model to provide logic diagrams demonstrating success paths for each fire area.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-67 - Revision 1
2.4.2.4 Fire Area Assessment (Taken From NFPA 805, 2001 Edition)
Fire Area Assessment. An engineering analysis shall be performed in accordance with the requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5. [See Chapter 4 for methods of achieving these performance criteria (performance-based or deterministic)].
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4 Fire Area Assessment and Compliance Strategies
By determining the location of each component and cable by fire area and using the cable to equipment relationships described above, the affected safe shutdown equipment in each fire area can be determined. Using the list of affected equipment in each fire area, the impacts to safe shutdown systems, paths and functions can be determined. Based on an assessment of the number and types of these impacts, the required safe shutdown path for each fire area can be determined. The specific impacts to the selected safe shutdown path can be evaluated using the circuit analysis and evaluation criteria contained in Section 3.5 of this document. Having identified all impacts to the required safe shutdown path in a particular fire area, this section provides guidance on the techniques available for individually mitigating the effects of each of the potential impacts.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-68 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1 Criteria/Assumptions The following criteria and assumptions apply when performing fire area compliance assessment to mitigate the consequences of the circuit failures identified in the previous sections for the required safe shutdown path in each fire area.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-69 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.1 Criteria/Assumptions Assume only one fire in any single fire area at a time.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The PINGP safe and stable analysis under NFPA-805 considers only one fire occurring in one area at a time.
Common enclosure concerns, common enclosure caused by loss of DC control power have been considered in the model and by other analysis (Reference Section 3.3.1.7 above).
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-70 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.2 Criteria/Assumptions Assume that the fire may affect all unprotected cables and equipment within the fire area. This assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the exposure fire that is required by the regulation.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP assumed that a fire could affect all unprotected equipment and cables in the fire area. PINGP did not credit fire dynamics (intensity or size) when analyzing the deterministic areas for fire-induced damage. For areas utilizing RI / PB methodology, fire modeling was often employed to demonstrate that success paths would remain available.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-71 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.3 Criteria/Assumptions Address all cable and equipment impacts affecting the required safe shutdown path in the fire area. All potential impacts within the fire area must be addressed. The focus of this section is to determine and assess the potential impacts to the required safe shutdown path selected for achieving post-fire safe shutdown and to assure that the required safe shutdown path for a given fire area is properly protected.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The PINGP analysis considered all cable and equipment impacts as a result of the fire and addressed the impacts to achieve success paths for each NSPC within the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-72 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.4 Criteria/Assumptions Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in accordance with NRC requirements.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns with Intent To achieve and maintain safe and stable conditions, and to minimize the use of recovery actions, the least fire-impacted success path was credited for each fire area.
All recovery actions that varied from the deterministic requirements were addressed via the VFDR process. Areas with VFDRs that were not solved by a modification utilized the RI / PB methodology.
Feasibility and / or Reliability of the recovery actions is being addressed as part of the NFPA 805 process as described in LAR Attachment G and Attachment S, Table S-3, Items 57, 58, and 62.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 Attachment G Attachment S, Table S-3, Items 57, 58, and 62
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-73 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.5 Criteria/Assumptions Where appropriate to achieve and maintain cold shutdown within 72 hours, use repairs to equipment required in support of post-fire shutdown.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The 10CFR50 Appendix R requirement to achieve and maintain cold shutdown within 72 hours is not a requirement of NFPA 805; NFPA 805 requires maintaining fuel in a safe and stable condition. PINGP achieves safe and stable conditions at Mode 3. Reference Section 3.1 (above) for additional information regarding PINGP’s safe and stable conditions.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-74 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.6 Criteria/Assumptions Appendix R compliance requires that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage (III.G.1.a). When cables or equipment, including associated circuits, are within the same fire area outside primary containment and separation does not already exist, provide one of the following means of separation for the required safe shutdown path(s):
- Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a fire barrier having a 3-hour rating (III.G.2.a)
- Separation of cables and equipment and associated non-safety circuits of redundant trains within the same fire area by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.b).
- Enclosure of cable and equipment and associated non-safety circuits of one redundant train within a fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic fire suppression system shall be installed in the fire area (III.G.2.c).
For fire areas inside non-inerted containments, the following additional options are also available:
- Separation of cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustibles or fire hazards (III.G.2.d);
- Installation of fire detectors and an automatic fire suppression system in the fire area (III.G.2.e); or
- Separation of cables and equipment and associated non-safety circuits of redundant trains by a noncombustible radiant energy shield (III.G.2.f).
Use exemptions, deviations and licensing change processes to satisfy the requirements mentioned above and to demonstrate equivalency depending upon the plant's license requirements.
Applicability
Applicable
Comments
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-75 - Revision 1
Alignment Statement
Not in Alignment [but Prior NRC Approval]
Alignment Basis
Areas crediting methods or mitigating strategies that
varied from the deterministic requirements were
addressed through the VFDR process.
In each case, a success path was assured for each
performance goal within each fire area.
Areas with VFDRs that could not be solved by a
modification, were addressed using the RI / PB
methods and not the deterministic methods.
As an exception to this section, PINGP is transitioning
existing approved licensing action for the oil collection
system in Fire Areas 1 and 71 (containment).
Additionally, as an exception to this section, PINGP is
transitioning existing approved licensing action for a
“repair action” to assure isolation of pressurizer PORVs
for a fire occurring in the cable spreading room or relay
room (Fire Areas 13 and 18 respectively) that could
cause spurious operation of PORV isolation valves.
Therefore, this section is “Not in Alignment but Prior
NRC Approval”. The details for this licensing action
can be found in Attachments K and T.
All fire areas utilizing these transitioning licensing
actions (exemptions) are evaluated using the RI / PB
methodology.
The basis for approval has been reviewed. There have
been no plant modifications or other changes that
would invalidate the basis for approval.
Reference Documents EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2
Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations
Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-76 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.7 Criteria/Assumptions Consider selecting other equipment that can perform the same safe shutdown function as the impacted equipment. In addressing this situation, each equipment impact, including spurious operations, is to be addressed in accordance with regulatory requirements and the NPP’s current licensing basis.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP considered multiple success paths and alternative equipment when deciding on how to best meet the NSPC. In each case, a success path was assured for each performance goal within each fire area by choosing the path least impacted by the fire, so as to minimize the reliance upon recovery actions. Spurious operations were addressed as detailed in the applicable sections of this document.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-77 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.1.8 Criteria/Assumptions Consider the effects of the fire on the density of the fluid in instrument tubing and any subsequent effects on instrument readings or signals associated with the protected safe shutdown path in evaluating post-fire safe shutdown capability. This can be done systematically or via procedures such as Emergency Operating Procedures.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns PINGP considered multiple success paths and alternative equipment when deciding on how to best meet the NSPC. In each case, a success path was assured for each performance goal within each fire area by choosing the path least impacted by the fire, so as to minimize the reliance upon recovery actions. Spurious operations were addressed as detailed in the applicable sections of this document. The impact of fire-induced effects on instrument tubing was addressed by EC 19988. This included process instrumentation as well as credited flowpath instrumentation.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805 EC 19988, NFPA 805 LAR Attachment B- RIS 04-003 (Updated) Analysis, B-2 Table Support
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-78 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2 Methodology for Fire Area Assessment
Refer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area assessment.
Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R compliance:
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Not Required Generic Guidance
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-79 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2.1 Identify the Affected Equipment by Fire Area
Identify the safe shutdown cables, equipment and systems located in each fire area that may be potentially damaged by the fire. Provide this information in a report format. The report may be sorted by fire area and by system in order to understand the impact to each safe shutdown path within each fire area (see Attachment 5 for an example of an Affected Equipment Report).
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The NSCA (Genesis) model provides a listing of equipment and associated cables, as well as a logical relationship for meeting NSPC requirements used to identify success paths for each fire area under the safe and stable analysis for NFPA 805. Support systems and interfaces were also identified on the logic trees within the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-80 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2.2 Determine the Shutdown Paths Least Impacted By a Fire in Each Fire Area
Based on a review of the systems, equipment and cables within each fire area, determine which shutdown paths are either unaffected or least impacted by a postulated fire within the fire area. Typically, the safe shutdown path with the least number of cables and equipment in the fire area would be selected as the required safe shutdown path. Consider the circuit failure criteria and the possible mitigating strategies, however, in selecting the required safe shutdown path in a particular fire area. Review support systems as a part of this assessment since their availability will be important to the ability to achieve and maintain safe shutdown. For example, impacts to the electric power distribution system for a particular safe shutdown path could present a major impediment to using a particular path for safe shutdown. By identifying this early in the assessment process, an unnecessary amount of time is not spent assessing impacts to the frontline systems that will require this power to support their operation.
Based on an assessment as described above, designate the required safe shutdown path(s) for the fire area. Identify all equipment not in the safe shutdown path whose spurious operation or maloperation could affect the shutdown function. Include these cables in the shutdown function list. For each of the safe shutdown cables (located in the fire area) that are part of the required safe shutdown path in the fire area, perform an evaluation to determine the impact of a fire-induced cable failure on the corresponding safe shutdown equipment and, ultimately, on the required safe shutdown path.
When evaluating the safe shutdown mode for a particular piece of equipment, it is important to consider the equipment’s position for the specific safe shutdown scenario for the full duration of the shutdown scenario. It is possible for a piece of equipment to be in two different states depending on the shutdown scenario or the stage of shutdown within a particular shutdown scenario. Document information related to the normal and shutdown positions of equipment on the safe shutdown equipment list.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The NSCA (Genesis) model was utilized to identify the NSPC paths least affected by the fire occurring in each fire area. The NSCA (Genesis) model identifies, and logically relates the NSPC components to their support equipment, such that a fire-induced failure of the support equipment will cascade as a loss of the NSPC component, and its path when applicable.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-81 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2.3 Determine Safe Shutdown Equipment Impacts
Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire area, and what those possible impacts are.
Applicability
Applicable
Comments
Alignment Statement
Alignment Basis
Reference Documents
Aligns The NSCA (Genesis) model generates reports (in logic tree format) of all affected equipment and cables within the fire area of concern. These reports include a listing of cascaded losses of support equipment. The circuit analysis tied to the cables within the model describes the potential impacts due to the fire.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-82 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2.4 Develop a Compliance Strategy or Disposition to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable
The available deterministic methods for mitigating the effects of circuit failures are summarized as follows (see Figure 1-2):
- Provide a qualified 3-fire rated barrier. - Provide a 1-hour fire rated barrier with automatic suppression and detection. - Provide separation of 20 feet or greater with automatic suppression and detection and demonstrate that there are no intervening combustibles within the 20 foot separation distance. - Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability. - Provide a procedural action in accordance with regulatory requirements. - Perform a cold shutdown repair in accordance with regulatory requirements. - Identify other equipment not affected by the fire capable of performing the same safe shutdown function. - Develop exemptions, deviations, Generic Letter 86-10 evaluation or fire protection design change evaluations with a licensing change process.
Additional options are available for non-inerted containments as described in 10 CFR 50 Appendix R section III.G.2.d, e and f.
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Aligns PINGP fire areas that are remaining deterministic, utilize one of the options listed in NFPA 805 Section 4.2.3 to assure success paths. For those areas differing from the deterministic requirements, VFDRs were created and resolved via the non-deterministic (RI / PB) methodology or through the modification process for areas remaining deterministic.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-83 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.4.2.5 Document the Compliance Strategy or Disposition Determined to Mitigate the Effects Due to Fire Damage to Each Required Component or Cable
Assign compliance strategy statements or codes to components or cables to identify the justification or mitigating actions proposed for achieving safe shutdown. The justification should address the cumulative effect of the actions relied upon by the licensee to mitigate a fire in the area. Provide each piece of safe shutdown equipment, equipment not in the path whose spurious operation or maloperation could affect safe shutdown, and/or cable for the required safe shutdown path with a specific compliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area Assessment Report documenting each cable disposition.
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Aligns PINGPs safe and stable model incorporates the resolution (compliance strategy) for components and cables affected by the fire (reference LAR Attachment C). The overall effect of recovery actions (mitigation activities), has been considered via the VFDR and FRE processes for RI / PB areas. For deterministic areas, VFDRs were resolved via the modification process. The NSCA (Genesis) model was utilized to develop input for the NSCA strategy reviews, which facilitated the process of generating the appropriate VFDRs.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-84 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5 Circuit Analysis and Evaluation This section on circuit analysis provides information on the potential impact of fire on circuits used to monitor, control and power safe shutdown equipment. Applying the circuit analysis criteria will lead to an understanding of how fire damage to the cables may affect the ability to achieve and maintain post-fire safe shutdown in a particular fire area. This section should be used in conjunction with Section 3.4, to evaluate the potential fire-induced impacts that require mitigation. Appendix R Section III.G.2 identifies the fire-induced circuit failure types that are to be evaluated for impact from exposure fires on safe shutdown equipment. Section III.G.2 of Appendix R requires consideration of hot shorts, shorts-to-ground and open circuits.
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Not Required Generic Guidance EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-85 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1 Criteria/Assumptions Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.
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Alignment Statement
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Not Required Generic Guidance EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-86 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1.1 Criteria/Assumptions Consider the following circuit failure types on each conductor of each unprotected safe shutdown cable to determine the potential impact of a fire on the safe shutdown equipment associated with that conductor.
- A hot short may result from a fire-induced insulation breakdown between conductors of the same cable, a different cable or from some other external source resulting in a compatible but undesired impressed voltage or signal on a specific conductor. A hot short may cause a spurious operation of safe shutdown equipment.
- An open circuit may result from a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit may prevent the ability to control or power the affected equipment. An open circuit may also result in a change of state for normally energized equipment. (e.g. [for BWRs] loss of power to the Main Steam Isolation Valve (MSIV) solenoid valves due to an open circuit will result in the closure of the MSIVs). Note that RIS 2004-03 indicates that open circuits, as an initial mode of cable failures, are considered to be of very low likelihood. The risk-informed inspection process will focus on failures with relatively high probabilities.
- A short-to-ground may result from a fire-induced breakdown of a cable insulation system, resulting in the potential on the conductor being applied to ground potential. A short-to-ground may have all of the same effects as an open circuit and, in addition, a short-to-ground may also cause an impact to the control circuit or power train of which it is a part.
Consider the three types of circuit failures identified above to occur individually on each conductor of each safe shutdown cable on the required safe shutdown path in the fire area.
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Aligns For safe and stable, PINGP considers hot shorts, open circuits and shorts-to-ground in alignment with NEI 00-01 Revision 1 as detailed in the following sections of this document.
Refer to Section 3.5.2.1 for discussion of open circuits.
Refer to Section 3.5.2.2 for discussion of shorts-to-ground.
Refer to Section 3.5.2.3 for discussion of hot shorts and wire-to wire shorts.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-87 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1.2 Criteria/Assumptions Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst must consider the position of the safe shutdown equipment for each specific shutdown scenario when determining the impact that fire damage to a particular circuit may have on the operation of the safe shutdown equipment.
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Aligns PINGP circuit analysis assumes that the initial contact and operational positions are per normal lineup as described in plant documentation. Cable selection included components that require repositioning to achieve and maintain safe and stable conditions. All logical relationships and applicable cable selection is maintained in the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-88 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1.3 Criteria/Assumptions Assume that circuit failure types resulting in spurious operations exist until action has been taken to isolate the given circuit from the fire area, or other actions have been taken to negate the effects of circuit failure that is causing the spurious actuation. The fire is not assumed to eventually clear the circuit fault. Note that RIS 2004-03 indicates that fire- induced hot shorts typically self-mitigate after a limited period of time.
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Aligns PINGP does not credit self-mitigation of hot shorts in its safe and stable analysis.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-89 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1.4 Criteria/Assumptions When both trains are in the same fire area outside of primary containment, all cables that do not meet the separation requirements of Section III.G.2 are assumed to fail in their worst case configuration.
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Aligns with Intent All unprotected cables, and equipment within the analysis area, that did not meet the separation requirements of NFPA 805 Section 4.2.3 were identified and the appropriate failure modes were considered and addressed through either deterministic or RI / PB methods.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-90 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.1.5 Criteria/Assumptions The following guidance provides the NRC inspection focus from Bin 1 of RIS 2004-03 in order to identify any potential combinations of spurious operations with higher risk significance. Bin 1 failures should also be the focus of the analysis; however, NRC has indicated that other types of failures required by the regulations for analysis should not be disregarded even if in Bin 2 or 3. If Bin 1 changes in subsequent revisions of RIS 2004-03, the guidelines in the revised RIS should be followed.
Cable Failure Modes. For multiconductor cables testing has demonstrated that conductor-to-conductor shorting within the same cable is the most common mode of failure. This is often referred to as "intra-cable shorting." It is reasonable to assume that given damage, more than one conductor-to-conductor short will occur in a given cable. A second primary mode of cable failure is conductor-to-conductor shorting between separate cables, commonly referred to as "inter-cable shorting." Inter-cable shorting is less likely than intra-cable shorting. Consistent with the current knowledge of fire-induced cable failures, the following configurations should be considered:
A. For any individual multiconductor cable (thermoset or thermoplastic), any and all potential spurious actuations that may result from intra-cable shorting, including any possible combination of conductors within the cable, may be postulated to occur concurrently regardless of number. However, as a practical matter, the number of combinations of potential hot shorts increases rapidly with the number of conductors within a given cable. For example, a multiconductor cable with three conductors (3C) has 3 possible combinations of two (including desired combinations), while a five conductor cable (5C) has 10 possible combinations of two (including desired combinations), and a seven conductor cable (7C) has 21 possible combinations of two (including desired combinations). To facilitate an inspection that considers most of the risk presented by postulated hot shorts within a multiconductor cable, inspectors should consider only a few (three or four) of the most critical postulated combinations.
B. For any thermoplastic cable, any and all potential spurious actuations that may result from intra-cable and inter-cable shorting with other thermoplastic cables, including any possible combination of conductors within or between the cables, may be postulated to occur concurrently regardless of number. (The consideration of thermoset cable inter-cable shorts is deferred pending additional research.)
C. For cases involving the potential damage of more than one multiconductor cable, a maximum of two cables should be assumed to be damaged concurrently. The spurious actuations should be evaluated as previously described. The consideration of more than two cables being damaged (and subsequent spurious actuations) is deferred pending additional research.
D. For cases involving direct current (DC) circuits, the potential spurious operation due to failures of the associated control cables (even if the spurious operation requires two concurrent hot shorts of the proper polarity, e.g., plus-to-plus and minus-to-minus) should be considered when the required source and target conductors are each located within the same multiconductor cable.
E. Instrumentation Circuits. Required instrumentation circuits are beyond the scope of this associated circuit approach and must meet the same requirements as required power and control circuits. There is one case where an instrument circuit could potentially be considered an associated circuit. If fire-induced damage of an instrument circuit could prevent operation (e.g., lockout permissive signal) or cause maloperation (e.g., unwanted start/stop/reposition signal) of systems
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-91 - Revision 1
necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an associated circuit and handled accordingly.
Likelihood of Undesired Consequences
Determination of the potential consequence of the damaged associated circuits is based on the examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other scenarios that could significantly impair the NPP’s ability to achieve and maintain hot shutdown. When considering the potential consequence of such failures, the [analyst] should also consider the time at which the prevented operation or maloperation occurs. Failures that impede hot shutdown within the first hour of the fire tend to be most risk significant in a first-order evaluation. Consideration of cold-shutdown circuits is deferred pending additional research.
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Aligns For cable selection and circuit analysis purposes, PINGP did not differentiate or pre-screen cables based upon cable type / insulation material (e.g., thermoset and thermoplastic cables). Additionally, where applicable, the guidance of RIS 2004-03 was considered and incorporated into the NSCA (Genesis) model.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-92 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2 Types of Circuit Failures Appendix R requires that nuclear power plants must be designed to prevent exposure fires from defeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits that provide control and power to equipment on the required safe shutdown path and any other equipment whose spurious operation/ maloperation could affect shutdown in each fire area must be evaluated for the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of fire damage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, it must be assured that one redundant train of equipment capable of achieving hot shutdown is free of fire damage for fires in every plant location. To provide this assurance, Appendix R requires that equipment and circuits required for safe shutdown be free of fire damage and that these circuits be designed for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect to the electrical distribution system, the issue of breaker coordination must also be addressed.
This section will discuss specific examples of each of the following types of circuit failures:
- Open circuit - Short-to-ground - Hot short.
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Not Required Generic Guidance
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-93 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2.1 Circuit Failures Due to an Open Circuit
This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit will typically prevent the ability to control or power the affected equipment. An open circuit can also result in a change of state for normally energized equipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV.
NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:
- Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment.
- In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evaluate this to determine if equipment fails safe.
- Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in secondary damage.
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Aligns Open circuits were considered for all unprotected cables within the fire area of concern, in keeping with the guidance of this section.
NSPM concluded that by using the industry accepted screening criteria; fire-induced damage to current transformer (CT) circuits would not adversely impact PINGP’s ability to achieve and maintain safe and stable conditions.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-94 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2.2 Circuit Failures Due to a Short-to- Ground
This section provides guidance for addressing the effects of a short-to-ground on circuits for safe shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation system resulting in the potential on the conductor being applied to ground potential. A short-to-ground can cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-ground may affect other equipment in the electrical power distribution system in the cases where proper coordination does not exist. Consider the following consequences in the post-fire safe shutdown analysis when determining the effects of circuit failures related to shorts-to-ground:
- A short to ground in a power or a control circuit may result in tripping one or more isolation devices (i.e. breaker/fuse) and causing a loss of power to or control of required safe shutdown equipment.
- In the case of certain energized equipment such as HVAC dampers, a loss of control power may result in loss of power to an interlocked relay or other device that may cause one or more spurious operations.
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Aligns PINGP has considered the effects of shorts-to-ground in alignment with the guidance given in this section to maintain safe and stable conditions.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-95 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2.3 Circuit Failures Due to a Hot Short This section provides guidance for analyzing the effects of a hot short on circuits for required safe shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between conductors of the same cable, a different cable or some other external source resulting in an undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed voltage would be to cause equipment to operate or fail to operate in an undesired manner.
Consider the following specific circuit failures related to hot shorts as part of the post-fire safe shutdown analysis:
- A hot short between an energized conductor and a de-energized conductor within the same cable may cause a spurious actuation of equipment. The spuriously actuated device (e.g., relay) may be interlocked with another circuit that causes the spurious actuation of other equipment. This type of hot short is called a conductor-to-conductor hot short or an internal hot short.
- A hot short between any external energized source such as an energized conductor from another cable (thermoplastic cables only) and a de-energized conductor may also cause a spurious actuation of equipment. This is called a cable-to- cable hot short or an external hot short. Cable-to-cable hot shorts between thermoset cables are not postulated to occur pending additional research.
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Aligns For safe and stable, PINGP utilizes the “hot probe” method to determine the effects of hot shorts on the circuits during the initial circuit analysis activity. The hot probe method is not dependent upon the source (inter, intra or power supply).
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-96 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2.4 Circuit Failures Due to Inadequate Circuit Coordination
The evaluation of associated circuits of a common power source consists of verifying proper coordination between the supply breaker/fuse and the load breakers/fuses for power sources that are required for safe shutdown. The concern is that, for fire damage to a single power cable, lack of coordination between the supply breaker/fuse and the load breakers/fuses can result in the loss of power to a safe shutdown power source that is required to provide power to safe shutdown equipment.
A coordination study should demonstrate the coordination status for each required common power source. For coordination to exist, the time-current curves for the breakers, fuses and/or protective relaying must demonstrate that a fault on the load circuits is isolated before tripping the upstream breaker that supplies the bus. Furthermore, the available short circuit current on the load circuit must be considered to ensure that coordination is demonstrated at the maximum fault level.
The methodology for identifying potential associated circuits of a common power source and evaluating circuit coordination cases of associated circuits on a single circuit fault basis is as follows:
- Identify the power sources required to supply power to safe shutdown equipment.
- For each power source, identify the breaker/fuse ratings, types, trip settings and coordination characteristics for the incoming source breaker supplying the bus and the breakers/fuses feeding the loads supplied by the bus.
- For each power source, demonstrate proper circuit coordination using acceptable industry methods.
- For power sources not properly coordinated, tabulate by fire area the routing of cables whose breaker/fuse is not properly coordinated with the supply breaker/fuse. Evaluate the potential for disabling power to the bus in each of the fire areas in which the associated circuit cables of concern are routed and the power source is required for safe shutdown. Prepare a list of the following information for each fire area:
- Cables of concern. - Affected common power source and its path. - Raceway in which the cable is enclosed. - Sequence of the raceway in the cable route. - Fire zone/area in which the raceway is located.
For fire zones/areas in which the power source is disabled, the effects are mitigated by appropriate methods.
- Develop analyzed safe shutdown circuit dispositions for the associated circuit of concern cables routed in an area of the same path as required by the power source. Evaluate adequate separation based upon the criteria in Appendix R, NRC staff guidance, and plant licensing bases.
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Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
Aligns Original installation and subsequent PINGP processes have followed proper design standards for circuit protection. Engineering manuals, short circuit evaluations and loading evaluations pertaining to proper circuit protection and coordination are in place to maintain circuit protection and coordination where required.
Loss of DC control power has been analyzed in the NSCA (Genesis) model, including common power supply and common enclosure considerations resulting from this fire-induced phenomenon.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment B - NEI 04-02 Table B-2
PINGP Page B-98 - Revision 1
NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance
3.5.2.5 Circuit Failures Due to Common Enclosure Concerns
The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to fire damage to a circuit either whose isolation device fails to isolate the cable fault or protect the faulted cable from reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.
The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.
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Aligns Original installation and subsequent PINGP processes have followed proper design standards for circuit protection. An analysis of the latest coordination studies was performed in support of section 3.5.2.4 of the B- 2 Table, and can be found in section 3.5.2.4. Credited plant physical barriers pertaining to the defined NFPA-805 fire areas were verified as part of the compartment analysis.
EC 23407 - B2 Methodology Evaluation EC 23408 - NEI 00-01 Revision 1 to Revision 2 Gap Analysis NSCA (Genesis) Model EC 20612 - Non Power / NSCA Operations Review for NFPA 805
Northern States Power - Minnesota Attachment C – NEI 04-02 Table B-3 Fire Area Transition
PINGP Page C-1 – Revision 1
C. NEI 04-02 Table B-3 – Fire Area Transition
315 Pages Attached
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 2- Revision 1
Unit Fire Area Description 1 1 Containment Unit 1 Fire Area 1 includes Fire Area(s): 68 Containment Annulus Unit 1 Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.3.4 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Note: Unit 1, one SG could be affected but the redundant SG remains available.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 3- Revision 1
Process Monitoring If Unit 1 Process Monitoring Train A is not available, use Train B RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
VFDR-001-01
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B)
Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
VFDR001-01-02
VFDR001-01-03
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 4- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST.
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST.
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Unit 1 - CC Train A or B
Unit 2 - CC Train A or B
CL Train A or B
Reference Documents
Safe/Genesis V 4.0.2 EC23621, Fire Risk Evaluation, Fire Area 01, Unit 1 Containment, Rev. 1, March 2014 Licensing Actions
Appendix R Exemption, Containment, RCP oil collection system not in strict compliance (III.O criteria), Units 1 and 2, Fire Areas 1 and 71 Reference Attachment K – Existing Licensing Action Transition for details Existing Engineering Equivalency Evaluations (EEEE)
EEEE Title FPEE 2011-003, Fire Protection Engineering Evaluation of Appendix R Compliance with Section III.G.2.D Since Containment Annulus Pre-Action Sprinkler System PA-3, PA-4, PA-6, and PA-7 May Not Actuate
Summary The purpose of this evaluation is to justify the treatment of the containment annulus in the same way the containment is treated in 10 CFR 50 Appendix R. Revision 1 to this FPEE evaluates the functional requirements for the existing pre-action fire suppression systems inside the containment annulus. FA 68 and 72 are in compliance with Appendix R, Section III.G.2.d, because redundant trains of equipment required for safe shutdown are separated by greater than 20 feet free of intervening combustibles. Although the Annulus is not inside the containment pressure boundary, it is inside the Reactor Containment Building and it qualifies to be treated like an area inside containment because access to the area is restricted in the same way access is restricted to containment during power operation. Reference 4.15 supports this position that the annulus is inside the Reactor Containment Building. Since FA 68 and FA 72 can be treated like areas inside containment, the partial area fire detection and automatic fire suppression in the Annulus do not need to be credited to meet the requirements of Appendix R, Section III.G.2. In addition, the functional requirements and surveillances required to ensure operability are not required for the cable tray sprinkler systems in the annulus of either unit, FA 68 and FA 72.
EEEE Title 97FP02-DOC-01 R0, RCP Lube Oil Compliance Review, 97FP02-DOC-01 R0 Add 1 App R RCP LO Sys Compliance Review
Summary The purpose of this compliance review is to evaluate the as-built Reactor Coolant Pump (RCP) Lube Oil Collection System (LOCS) at
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 5- Revision 1
Prairie Island Nuclear Generating Plant (PINGP) with respect to the licensing and design basis commitments relative to the requirements of 10 CFR 50, Section 111.0, "Reactor Coolant Pump Lube Oil Collection Systems". Based on the installation of a metal shroud around the oil lift pump and piping, the RCP lube oil collection system at Prairie Island for Units 1 and 2 would adequately collect oil from postulated pressurized and unpressurized leak sites.
Variances from Deterministic Requirements (VFDR)
VFDR-001-1-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of LOOP 1L-433, Pressurizer Level Cold Calibration Instrument and LOOP 1L-426-RP, Pressurizer Level Red Channel. This could result in the loss of the ability to observe adequate RCS Inventory. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3.4.b, due to lack of separation between redundant trains of Pressurizer Level Indication. Components and Cables: Pressurizer Level Cold Calibration Instrument, LOOP 1L-433 (1CF-31) Pressurizer Level Red Channel, LOOP 1L-426RP (1CR-36) Compliant Case: One train of Pressurizer Level Indication (LOOP 1L-433) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes risk is low for a fire to cause concurrent damage to redundant Pressurizer Level Indication.
This VFDR has been evaluated and it was determined to meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-001-1-02 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31245
(11 Reactor Coolant Pump (RCP) Thermal Barrier Heat Exchanger (TBHX)) and CV-31335 (11 RCP Seal Injection). This could cause a loss of all RCP seal cooling to 11 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3.4.b, due to lack of separation between redundant trains of RCP seal cooling.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Components and Cables: 11 REAC CLNT PMP SL WTR OUTL ISOL ,CV-31335 ( 1C-1080, 1C-1076, 1C-1075) 11 RCP TBHX CC, CV-31245 (1C-2163, 1C-2178, 1C-2179, 1C-2180, 1C-4641, 1C-4643) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling.
This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-001-1-03 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31246
(12 Reactor Coolant Pump (RCP) Thermal Barrier Heat Exchanger (TBHX) and CV-31336 (12 RCP Seal Injection). This could cause a loss of all RCP seal cooling to 12 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3.4.b, due to lack of separation between redundant trains of RCP seal cooling. Components and Cables: 12 RCP seal injection outlet valve, CV-31336 (1C-1081, 1C-1082, 1C-1086) 12 RCP TBHX CC, CV-31246 (1C-2162, 1C-2174, 1C-2175, 1C-2176, 1C-4640, 1C-4642) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area.
Disposition Recovery Action: No recovery actions credited. Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
1 Detection 10, 20, 29, 32
Ionization N N N Y N
Suppression - - - - - - -
Feature 1CV-T42I
RES N N N Y N Raceway 1CV-T42I is wrapped with a radiant energy shield
68 Detection 21 Ionization N N N Y N Suppression PA-3 Pre-Action N N N N N This suppression system is in the Annulus.
Suppression PA-4 Pre-Action N N N N N This suppression system is in the Annulus. Feature - - - - - - -
1 Feature N/A Lube Oil Collection System
N Y N N N
Hose Station N N N Y N
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
There are automatic fire suppression systems in the Annulus portion of the fire area. Water will drain via grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 1 now includes the Unit 1 Annulus, which was Fire Area 68 prior to the transition to NFPA 805.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 2 Ventilation Fan Room Unit 1 & 2 Fire Area 2 includes Fire Area(s): 76 Vent and Fan Room Unit 2 Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 9- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 10- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B Unit 1 - CC Train A or B Unit 2 - CC Train A or B CL Train A or B
EEEE Title FPEE-12-001, CA-01314253-01, VFD-63 & VFD-64
Summary The purpose of this evaluation is to assess VFD-63 and VFD-64, located on the 755ft elevation of the Auxiliary Building in Fire Area 2 and Fire Area 76, respectively. Exposure fires in Fire Area 8, Fire Area 2, and Fire Area 76 would not be expected to cause failure of the 6 inch diameter 14 gauge steel duct pipe between the fire damper in the 1ft x 2ft steel enclosure and G-wall. The duct pipes are securely attached with bolts to both sides of the steel enclosures. The construction of the steel enclosures and duct pipes, along with method of attachment, are capable of withstanding the impact of postulated exposure fires to which they could be exposed. Fire Area 8, Fire Area 2, and Fire Area 76 are in compliance with Appendix R since all required safe shutdown functions are available from the control room, with manual actions following C37.9 AOP1 and C37.9 AOP2 credited in Fire Area 2 to establish temporary Control Room and Relay Room HVAC. As such, fires in Fire Area 8, Fire Area 2, and Fire Area 76 that spread through G-wall due to failure of 6 inch diameter 14 gauge steel pipe between G-wall and VFD-63 and VFD-64 will not impact safe shutdown capability.
EEEE Title CA-01311057-01, Fire Doors 146/181 and 147/273
Summary The purpose of this evaluation is to assess two pairs of fire doors in series, Fire Doors 181/146 and 147/273. Each door in the series pairs has a 1-1/2hr fire rating, providing 3-hr fire rated protection for the barriers in which they are located. Each pair of doors in series is on opposite ends of an airlock. The door pairs are required for steam exclusion and High Energy Line Break (HELB), along with Appendix R, necessitating that the pair of doors can both open at the same time. As such, the doors are not provided with positive latching mechanisms. The doors swing into Fire Area 61A from Fire Area 2 and Fire Area 76. A fire in Fire Area 61A will not spread into Fire Area 2 or Fire Area 76 even though the doors are not provided with positive latching because any increase in pressure due to a fire in Fire Area 61A will push the doors back onto their door stops, preventing them from opening. Fire Area 61A is in compliance with Appendix R since all required safe shutdown functions are available from the control room. As such, fire spreading through the doors due to postulated fires in either Fire Area 2 or Fire Area 76 will not impact on fire safe shutdown capability.
EEEE Title FPEE 01086132-01, Condition/Fire Protection Evaluation Adequacy of the fire boundary between Fire Area 76 (Vent and Fan Room, Unit
2, 755' Auxiliary Building) and Fire Area 92 (Water Chiller Room, Unit 2, 755' Auxiliary Building) without an installed three-hour fire damper
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 11- Revision 1
Summary The purpose of this evaluation is to address the adequacy of the fire boundary between Fire Area 76 (Vent and Fan Room, Unit 2, 755' Auxiliary Building) and Fire Area 92 (Water Chiller Room, Unit 2, 755' Auxiliary Building) without an installed three-hour fire damper in a duct transversing the barrier. Although the duct work is not in a fire-tested configuration, the construction of the duct itself provides a one-hour measure of fire protection. Additionally, a lack of combustibles and ignition sources in the vicinity of the duct, as well as control of transient combustibles, minimizes the risk of fire. Should a fire occur, area-wide detection would quickly alert operators to the presence of products of combustion. The existing duct configuration (without damper) provides adequate protection from a fire in Fire Area 76 propagating into Fire Area 92.
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
None
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
2 Detection 30, 108 Ionization N N Y Y N
Suppression - Deluge N N N N N Filter system
Feature - - - - - - -
76 Detection 53 Ionization N N N Y N Suppression - Deluge N N N N N Filter system
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water from all sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 12- Revision 1
the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 3 Water Chiller Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 14- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 15- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 16- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 4 Fuel Handling Area Fire Area 4 includes Fire Area(s): 39 Radwaste Building 40 Maintenance Storage Shed 61A Aux Building Hatch Area 62 Spent Fuel Pool Area 67 Resin Disposal Building 93 Drum Storage/Low Level Rad Waste Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 18- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Vital Auxiliaries Unit 1 - Offsite Power (1R) supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power (2RY) supplying Electrical Distribution Trains A or B
Summary The purpose of this evaluation is to assess the unrated penetrations through the transoms and the unrated ¼-in checker steel plates and tray/conduit penetrations above Doors 94 and 95 for impact on the fire area boundaries separating Fire Areas 73 and 58 from Fire Area 4. This evaluation also assesses the adequacy of the assemblies to provide adequate protection for the 3-hour barrier in which they are located. The bases for this conclusion include the following: Postulated fires north of Unit 1 Door 95 and Unit 2 Door 94 would not adversely impact on redundant safe shutdown capability in Fire Area 58 and Fire Area 73 based on approved exemption requests for lack of automatic suppression that rely on enclosing Division B safe shutdown cable in one hour rated fire barriers, enclosing Division A safe shutdown cable trays in the vicinity of MCCs and at specified coordinates in one hour rated fire barriers, low combustible loading, automatic detection, and fire brigade response. Postulated fire spread south through the unrated penetrations and ¼-in checker steel plate from Fire Area 58 or Fire Area 73 to Fire Area 4 will not impact on safe shutdown capability since the only cables of concern for safe shutdown in the event of a fire in Fire Area 4 are 1CT-1, 16408-1, and 15407-3. These cables provide offsite power from the CT transformer to Bus 15 and Bus 16 and run from FA 4 to FA 58/73. The 1R transformer remains available to provide offsite power to Bus 15 and Bus 16 in these areas. Since these cables are also routed in Fire Areas 58 and 73, Doors 95 and 94 do not separate redundant trains of safe shutdown equipment. A minimum of 50% of the combustible loading is associated with Kerite cables in the overhead trays that pass through the unrated penetration seals in the ¼-in checker steel plate. Kerite cable is of thermoset construction and will not result in a fire without an external fire source. Trash receptacle and transient combustible fires that also involve the racks of plastic rolls and material in the metal enclosed cabinets south of Unit 1 Door 95 or the 8.5% hydrogen cylinders south of Unit 2 Door 94 could generate enough energy to ignite the cables in the cable trays. Flame propagation along thermoset cables is about 3mm per second or 3-1/2ft per hour. Due to the swing of Unit 1 Door 95 and Unit 2 Door 94 into the corridor portions of Fire Area 4, the trash receptacle and transient combustible fire would be at least 4ft away from the unrated penetrations and ¼-in checker steel plate, with a burn time of upwards of one hour required before a cable tray fire would impact on the barrier. Automatic detection in the corridors on both sides of Unit 1 Door 95 and Unit 2 Door 94 would actuate early in the fire. Detection alarms would result in prompt response by the fire brigade. Hose stations are available on both sides of Unit 1 Door 95 and Unit 2 Door 94 for fire
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 20- Revision 1
fighting purposes. Fire brigade response would be effective in controlling and extinguishing postulated fires prior to the fire spreading through the unrated penetrations and impacting on fire safe shutdown capability on the north sides of Unit 1 Door 95 and Unit 2 Door 94. The main open area of Fire Area 4 extends up to the 755ft elevation. The minimal continuity of intervening combustibles (although some combustible materials may be present in the intervening space) combined with the open area above the location of combustible materials in this location effectively ensures that a fire in one corridor of Fire Area 4 will not spread across the open area and back into the opposite corridor. As such, a fire in one corridor in Fire Area 4 will not result in a simultaneous fire in the opposite corridor, and the unsealed penetrations and ¼-in checker steel plates above Unit 1 Door 95 and Unit 2 Door 94 will not be challenged by the same fire simultaneously.
EEEE Title CA-01311057-01, Fire Doors 146/181 and 147/273
Summary The purpose of this evaluation is to assess two pairs of fire doors in series, Fire Doors 181/146 and 147/273. Each door in the series pairs has a 1-1/2hr fire rating, providing 3-hr fire rated protection for the barriers in which they are located. Each pair of doors in series is on opposite ends of an airlock. The door pairs are required for steam exclusion and High Energy Line Break (HELB), along with Appendix R, necessitating that the pair of doors can both open at the same time. As such, the doors are not provided with positive latching mechanisms. The doors swing into Fire Area 61A from Fire Area 2 and Fire Area 76. A fire in Fire Area 61A will not spread into Fire Area 2 or Fire Area 76 even though the doors are not provided with positive latching because any increase in pressure due to a fire in Fire Area 61A will push the doors back onto their door stops, preventing them from opening. Fire Area 61A is in compliance with Appendix R since all required safe shutdown functions are available from the control room. As such, fire spreading through the doors due to postulated fires in either Fire Area 2 or Fire Area 76 will not impact on fire safe shutdown capability.
EEEE Title FPEE 01121594-01, Basis for the use of (2) 1 1/2 Hour Rated Fire Doors in a (3) Hour Fire Rated Barrier
Summary The purpose of this evaluation is to address the adequacy of Class B (1.5 hour) fire doors in Appendix R-required fire barriers. This evaluation, through the use of a bounding condition, finds that the lesser-rated doors, where used in PINGP, are acceptable given the combustible loading and available suppression systems of the areas.
EEEE Title FPEE-11-049, NFPA 80, 1968 Code Compliance Deviations
Summary The purpose of this analysis is to document the review of required fire doors in F5 Appendix K fire barriers for compliance with the requirements of National Fire Protection Association (NFPA) 80 -1968, Standard for Fire Doors and Windows (Code of Record). Six deviations have been justified as “acceptable”; therefore, no further action is necessary. Eight deviations require additional actions to resolve the noncompliance associated with use of gates not doors in required fire barriers, insulation attached to the inside of a required door, access openings through three doors, four doors for which a fire rating could not be determined (two of which include transoms), the qualification of frames in fire door assemblies, transoms with penetrations above two rated doors, doors that would not close under airflow, and two sets of door combinations without positive latching. Each item is being tracked in the corrective action program.
EEEE Title
Summary
FPEE-12-005, CA-01311402-03, Fire Doors 136 & 139
The purpose of this analysis is to document the review of required fire doors in F5 Appendix K fire barriers for compliance with the requirements of National Fire Protection Association (NFPA) 80 -1968, Standard for Fire Doors and Windows (Code of Record). Six deviations have been justified as “acceptable”; therefore, no further action is necessary. Eight deviations require additional actions to resolve the noncompliance associated with use of gates not doors in required fire barriers, insulation attached to the inside of a required door, access openings through three doors, four doors for which a fire rating could not be determined (two of which include transoms), the qualification of frames in fire door assemblies, transoms with penetrations above two rated doors, doors that would not close under airflow, and two sets of door combinations without positive latching. Each item is being tracked in the corrective action program.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 21- Revision 1
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
4 Detection 8, 30, 33
Ionization N N Y N N
Suppression SWP-1 Wet Pipe N N N N N Stairwell
Feature - - - - - - -
39 Detection 34, 81 Ionization, Thermal, Smoke
N N N N N
Suppression SWP-1 Wet Pipe N N N N N Stairwell System
Feature - - - - - - -
40 Detection - - - - - - -
Suppression - - - - - - -
Feature - - - - - - -
61A Detection 30 Ionization N N N N N
Detection 53 Ionization N N N N N
Suppression - - - - - - - Feature - - - - - - -
62 Detection 30 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - -
67
Detection
34 Ionization, Smoke
N N N N N
Detection 81 Ionization N N N N N
Suppression SWP-1 Wet Pipe N N N N N Stairwell System Feature - - - - - - -
93 Detection 104 Thermal, Flame
N N N N N
Suppression DM-7 Deluge N N N N N
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 22- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
There is an automatic fire suppression system in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 23- Revision 1
Unit Fire Area Description 1, 2 8 Turbine Building Fire Area 8 includes Fire Area(s): 9 Maintenance Shops 14 Working Material Lunch Room 21 Unit 1 4.16 KV Normal Switchgear, (Bus 13, 14) 23 Unit 2 4.16 KV Normal Switchgear (Bus 23, 24) 27 Water Conditioning Equpiment Area 69 Turbine Building Ground Floor & Mezzanine Floors Unit 1 70 Turbine Building Ground Floor & Mezzanine Floors Unit 2
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW Pump to 11 SG
Unit 2 - 21 MDAFW Pump to 21 SG
VFDR-008-1-01
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 24- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-427) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 – D5 supplying Electrical Distribution Train A from Bus 25 via 4KV bus cross-connect breakers (BKR-15-8, BKR-25-17) Unit 2 - D5 supplying Electrical Distribution Train A
Unit 1 - CC Train A
Unit 2 - CC Train A
CL Train A
Reference Documents
Safe/Genesis V 4.0.2
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
EEEE Title FPEE-12-002; CA-01327430-1, Steam Line Pipe Penetrations without Penetration Seals PENF-1526, PENF-1528, PENF-1689, & PENF-
1692
Summary The purpose of this evaluation is to assess the impact on fire safe shutdown capability of four main steam line penetrations that are not provided with 3-hour fire rated penetration seals in G wall separating the Turbine Operating Deck from the 735ft elevation of the Auxiliary Building. The types, quantities, and continuity of combustible materials in Fire Area 60 or Fire Area 75 would not result in a sufficient heat release rate or fire to spread to, or spread through, the main steam line penetrations into Fire Area 8. The Turbine Stop Valves in Fire Area 8 are between 50ft and 70ft from the main steam line penetrations through G-wall. The only postulated fires that could result in damage to the Turbine Stop Valves and be large enough to potentially spread to the main steam line penetrations would involve either a turbine bearing oil fire or a catastrophic failure of the turbine oil system. The turbine bearings are protected by an automatically-actuated preaction sprinkler system that will respond to postulated turbine bearing oil fires and result in prompt fire brigade response. Postulated catastrophic turbine oil system failures could result in very severe fires; however, the area where oil piping runs and oil can spread are protected by automatic wet pipe sprinkler systems. Postulated turbine bearing oil fires and postulated catastrophic turbine oil system failure fires would result in significant heat release rates and smoke production; however, the large volume of the turbine building combined with the existing roof exhaust fans and smoke hatches that are fitted with automatic releases would release smoke and hot gas to the environment and delay the effects of such fires from banking down to the level of the main steam line penetrations located 50ft below the roof. The main steam line penetrations have limited annular gaps, 10in, for passage of fire effects to Fire Area 60 and Fire Area 75. There is very limited continuity of combustible materials in Fire Areas 60 and Fire Area 75 for fire to spread from Fire Area 8 to the vicinity of the MSlVs and solenoid valves. Postulated fires in Fire Area 8, Fire Area 60, and Fire Area 75 would not adversely impact redundant safe shutdown capability consisting of the Turbine Stop Valves in Fire Area 8 and the MSIVs and solenoid valves in Fire Area 60 and Fire Area 75.
EEEE Title FPEE-12-001, CA-01314253-01, VFD-63 & VFD-64
Summary The purpose of this evaluation is to assess VFD-63 and VFD-64, located on the 755ft elevation of the Auxiliary Building in Fire Area 2 and Fire Area 76, respectively. Exposure fires in Fire Area 8, Fire Area 2, and Fire Area 76 would not be expected to cause failure of the 6 inch diameter 14 gauge steel duct pipe between the fire damper in the 1ft x 2ft steel enclosure and G-wall. The duct pipes are securely attached with bolts to both sides of the steel enclosures. The construction of the steel enclosures and duct pipes, along with method of attachment, are capable of withstanding the impact of postulated exposure fires to which they could be exposed. Fire Area 8, Fire Area 2, and Fire Area 76 are in compliance with Appendix R since all required safe shutdown functions are available from the control room, with manual actions following C37.9 AOP1 and C37.9 AOP2 credited in Fire Area 2 to establish temporary Control Room and Relay Room HVAC. As such, fires in Fire Area 8, Fire Area 2, and Fire Area 76 that spread through G-wall due to failure of 6 inch diameter 14 gauge steel pipe between G-wall and VFD-63 and VFD-64 will not impact safe shutdown capability.
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 26- Revision 1
adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-008-1-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of the steam supply valve (CV-31998) to 11 Turbine Driven Aux Feedwater Pump or spurious closure of CV-31153, 11 TDAFWP recirculation and lube oil cooling. 12 MDAFWP is failed due to loss of power to BUS-16. This could prevent the credited 11 TDAFWP from providing AFW flow to the Steam Generators to support Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, lack of separation between redundant trains of Decay Heat Removal. Components and Cables: 11 Turbine Driven AFW Pump Main Steam Supply, CV-31998 (1CA-1109, 1CA-1111, 1CA-1248) 11 Turbine Driven AFW Pump Recirc. Lube Oil CLG Control Valve, CV-31153 (1CA-1111, 1CA-1248) Compliant Case: The 11 TDAFWP (CV-31998 and CV-31153) should remain unaffected by a fire to provide AFW to 11 Steam Generator.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #11 will re-wire and re-route cables 1CA-1109, 1CA-1111, and 1CA-1248 so a fire in FA 8 cannot cause spurious closure of CV-31998 and CV-31153. This VFDR has been evaluated and it has been determined to meet the acceptance criteria of NFPA 805 Section 4.2.3 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 27- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
8 Detection 24, 49 Heat N N Y N N Detection 107 Heat, Smoke N N N N N Smoke detector in Elevator Machine
room
Detection 27 Ionization, Thermal. Heat
N N N N N
Detection 107 Smoke N N N N N Smoke detector in Elevator Machine room
Suppression PA-14 Pre-Action N N Y Y N Unit 1 Generator and Exciter
Suppression PA-15 Pre-Action N N Y Y N Unit 2 Generator and Exciter
Suppression WPS-9 Wet Pipe N N Y N N Protects Elevator Machine room and pit Suppression SWP-6 Wet Pipe N N Y N N Stairway System
Detection 107 Smoke N N N N N Smoke detector in elevator Machine room
Suppression SWP-6 Wet Pipe N N N N N Stairway System
Feature - - - - - - - 14 Detection 15 Ionization N N N N N
Suppression WPS-25 Wet Pipe N N N N N Feature - - - - - - -
21 Detection 84 Ionization N N N Y N Suppression - - - - - - - Feature - - - - - - -
23 Detection 86 Ionization N N N Y N Suppression - - - - - - -
Feature - - - - - - - 27 Detection 4 Ionization N N N N N
Suppression SWP-6 Wet Pipe N N N N N
Feature - - - - - - -
69 Detection 3 Ionization N N N Y N
Detection 4 Ionization, Heat, Smoke
N N N Y N
Detection 15 Ionization, Flame N N N Y N
Detection 107 Heat, Smoke N N N Y N Smoke detector in elevator Machine room
Suppression WPS-7 Wet Pipe N N N N N
Suppression WPS-8 Wet Pipe N N N Y N
Suppression WPS-9 Wet Pipe N N N N N
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 28- Revision 1
Suppression WPS-18 Wet Pipe N N N N N Suppression DA-1 Deluge N N N N N
Suppression DA-3 Deluge N N N Y N
Suppression PA-14 Pre-Action N N N N N Suppression SWP-3 Wet Pipe N N N Y N Stairwell system
Suppression SWP-5 Wet Pipe N N N N N Stairwell system Feature - - - - - - -
70 Detection 36 Ionization N N N Y N Detection 37 Ionization, Heat, Smoke N N N Y N Detection 44 Ionization, Thermal N N N Y N
Suppression WPS-15 Wet Pipe N N N N N
Suppression WPS-16 Wet Pipe N N N Y N
Suppression WPS-17 Wet Pipe N N N N N
Suppression WPS-21 Wet Pipe N N N N N
Suppression DA-4 Deluge N N N Y N
Suppression DA-5 Deluge N N N N N
Suppression PA-15 Pre-Action N N N N N
Suppression SWP-13 Wet Pipe N N N N N Stairwell System
Suppression SWP-14 Wet Pipe N N N N N Stairwell System
Feature - - N N N N N
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk
Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
There are automatic fire suppression systems in the fire area. Safety related electrical equipment is mounted on pedestals above the floor level minimizing the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 29- Revision 1
Unit Fire Area Description 1 10 Train A Event Monitoring Equipment Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 30- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 31- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
10 Detection 26 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 32- Revision 1
Unit Fire Area Description 1 11 Unit 1 Normal Switchgear & Control Rod Drive Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 33- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room or use Alternate Rod Insertion. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 34- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
11 Detection 87 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 35- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 36- Revision 1
Unit Fire Area Description 1 12 OSC Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 37- Revision 1
Safety Injection Pump (Train B) to inject borated water from the RWST Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A and B
Unit 2 - Offsite Power supplying Electrical Distribution Trains A and B
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
12 Detection 25 Ionization N N N N N
Suppression WPS-23, 24
Wet Pipe N N N Y N
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 38- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
There are automatic fire suppression systems in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 39- Revision 1
Unit Fire Area Description 1, 2 13 Control Room Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach Note: One train may be affected but the redundant train is likely to remain available.
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
VFDR-013-1-03 VFDR-013-2-03
Process Monitoring Note: Unit 1 and 2, one train of process monitoring could be affected but the redundant train remains available. RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp
VFDR-013-1-05 VFDR-013-2-05
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 40- Revision 1
(LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Train A or B RCS Head Vents or Pressurizer Vents Unit 2 - Charging System (Train A) or Safety Injection (Train B) Train A or B RCS Head Vents or Pressurizer Vents
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
VFDR-013-1-01 VFDR-013-2-01
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B, or D1 Emergency Diesel Generator supplying Electrical Distribution Train A Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B, or D5 Emergency Diesel Generators supplying Electrical Distribution Train A
Safe/Genesis V 4.0.2 EC 23605, Fire Risk Evaluation, Fire Area 13, Control Room, Rev. 1, March 2014. Procedure F5 Appendix B, Control Room Evacuation (Fire), Rev. 45 Licensing Actions
Appendix R Exemption, Control Room, Allowance for removal of fuses, Units 1 and 2, Fire Area 13 Reference Attachment K – Existing Licensing Action Transition for details
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating”
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 41- Revision 1
acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
VFDR-013-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Critera is not met for Vital Auxiliary Systems.
This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation
between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems.
Components and Cables: Many
Compliant Case: Train A and Train B Cooling Water Strainers (CV-31652, CV-31653, CV-31654, CV-31655, MTR-111C-21, MTR-111C-22,
MTR-121C-21 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix B.
Modification identified in Table S-2, Item #20 to correct fuse/breaker coordination on PNL-137.
This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification and recovery actions credited.
VFDR-013-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Reactivity Control. This could cause a loss of Reactivity Control. The Nuclear Safety Performance Criteria is not met for Reactivity Control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Reactivity Control. Components and Cables: Many
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 42- Revision 1
Compliant Case: The ability to maintain Reactivity Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment C) to manually trip Unit 1 Turbine at the Front Standard and close MV-32006 and MV-32010 (Attachment B). This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain RCS Inventory Control. This could cause a loss of Inventory Control. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Inventory and Pressure Control. Components and Cables: Many Compliant Case: The ability to maintain RCS Inventory and Pressure Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in F5 Appendix B [Reference 15] (Attachments A, B, E and F) to isolate Letdown, Excess Letdown, Head vents, Pressurizer vents, de-energize Sump B valves, de-energize the Containment Spray Pump, locally trip the Reactor Coolant Pumps at the Bus, and actions to re-align 12 Charging Pump suction to the RWST and restart a charging pump. Evaluate Recovery Actions to de-energize and manually close MV-32084 and MV-32085. Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after suction from the RWST is restored to inject borated water into the RCS. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and plant modifications credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 43- Revision 1
VFDR-013-1-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Decay Heat Removal. This could cause a loss of Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Decay Heat Removal. Components and Cables: Many Compliant Case: The ability to maintain Decay Heat Removal should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment E) to ensure steam supply valve from 11 SG is open and (Attachment I) to manually start the 11 TDAFWP and align to 11 SG. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-1-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to support Post Fire Safe Shutdown. This could cause a loss of Cooling Water. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems. Components and Cables: Many Compliant Case: The ability to maintain Vital Auxiliaries should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 44- Revision 1
Modification identified in Table S-2, Item #13 will ensure hot shorts on cables for CV-31505 which supplies cooling water to D1 Emergency Diesel Generator are isolated when the control switch is placed in local. Modification identified in Table S-2, Item #14 will ensure the Diesel Driven Cooling Water Pumps remain available to provide cooling water to D1 Emergency Diesel Generator and AFW pumps. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-013-1-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Process Monitoring. Steam Generator Level, Pressurizer Level, RCS Pressure, RCS Hot Leg and Cold Leg temperature, and Source Range Neutron Flux This represents a loss of Process Monitoring indication in the Control Room and would require abandonment of the Control Room and Alternate Shutdown. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Process Monitoring. Compliant Case: One train of Process Monitoring indication should remain available in the Control Room.
Disposition Recovery Action Evaluate risk of recovery actions to locally read indication at Hot Shutdown Panels for RCS Pressure, Hot/Cold leg Temperature, Pressurizer Level, SG Level, and Source Range Neutron Flux. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-1-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of vital AC Power from Bus 15 and Bus 16. This could cause a loss of Vital AC Power. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital AC Power.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Components and Cables: Many Compliant Case: The ability to maintain Vital AC Power should remain available from the Control Room.
Dispostion
Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment F) to locally start D1 Emergency Diesel Generator and manually align to BUS 15 to restore Vital AC Power. Modification identified in Table S-2, Item #34 will install a synchronization relay that will prevent non-synchronous parallels in all fire areas with exception to areas the gear is located in for the normal, alternate, and diesel generator source breakers for BUS 15, BUS 16, BUS 25, and BUS 26. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-013-1-07
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of the 11 Component Cooling Water Pumps. Components and Cables: Many Compliant Case:
VFDR-013-2-01
Cooling to the RCP seals should remain unaffected by a fire in this area. Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited. This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Reactivity Control. This could cause a loss of Reactivity Control.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 46- Revision 1
The Nuclear Safety Performance Criteria is not met for Reactivity Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Reactivity Control. Components and Cables: Many Compliant Case: The ability to maintain Reactivity Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment C) to manually trip Unit 2 Turbine at the Front Standard, and close steam valves MV-32021 and MV-32022, Attachment B. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain RCS Inventory Control. This could cause a loss of Inventory Control. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Inventory Control. Components and Cables: Many Compliant Case: The ability to maintain RCS Inventory and Pressure Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in F5 Appendix B [Reference 15] (Attachments A, B, D and G) to isolateLetdown, Excess Letdown, Head vents, de-energize Sump B valves, de-energize the Containment Spray Pumps, locally trip the Reactor Coolant Pumps, and actions to re-align 22 Charging Pump suction to the RWST and restart a charging pump. Evaluate Recovery Actions to de-energize and manually close MV-32178 and MV-32179.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 47- Revision 1
Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after suction from the RWST is restored to inject borated water into the RCS. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and plant modifications credited.
VFDR-013-2-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Decay Heat Removal. This could cause a loss of Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a Lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Decay Heat Removal. Components and Cables: Many Compliant Case: The ability to maintain Decay Heat Removal should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in F5 Appendix B (Attachment I) to manually start the 22 TDAFWP and align to 21 Steam Generator. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-2-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to support Post Fire Safe Shutdown. This could cause a loss of Cooling Water. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems. Components and Cables: Many
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 48- Revision 1
Compliant Case: The ability to maintain Vital Auxiliaries should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #14 will ensure the Diesel Driven Cooling Water Pumps remain available to provide cooling water to D1 Emergency Diesel Generator and AFW pumps. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-013-2-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Process Monitoring. Steam Generator Level, Pressurizer Level, RCS Pressure, RCS Hot Leg and Cold Leg temperature, and Source Range Neutron Flux This represents a loss of Process Monitoring IndicationProcess Monitoring indication in the Control Room and would require abandonment of the Control Room and Alternate Shutdown. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Process Monitoring. Compliant Case: One train of Process Monitoring indication should remain available in the Control Room.
Disposition Recovery Action: Evaluate risk of recovery actions to locally read indication at Hot Shutdown Panels for RCS Pressure, Hot/Cold leg Temperature, Pressurizer Level, SG Level, and Source Range Neutron Flux. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-013-2-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of vital AC Power from Bus 25 and Bus 26. This could cause a loss of Vital AC Power. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 49- Revision 1
redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital AC Power. Components and Cables: Many Compliant Case: The ability to maintain Vital AC Power should remain available from the Control Room.
Disposition
VFDR-013-2-07
Disposition
Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment G) to locally start D5 Emergency Diesel Generator and manually align to BUS 25 to restore Vital AC Power. Modification identified in Table S-2, Item #34 will install a synchronization relay that will prevent non-synchronous parallels in all fire areas with exception to areas the gear is located in for the normal, alternate, bus tie, and diesel generator source breakers for BUS 15, BUS 16, BUS 25, and BUS 26. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited. This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of the 21 Component Cooling Water Pumps.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliaries
Components and Cables: None Compliant Case: Cooling to the RCP seals should remain unaffected by a fire in this area. Recovery Action(s): None Modification identified in Table S-1, Item #1 has installed new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 50- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
13 Detection 57 Ionization N N N Y N Area wide and under raised floor
Suppression - - - - - - -
Hose Station - - N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 51- Revision 1
Unit Fire Area Description 1 14 Working Material, Lunch Room Note: Fire Area 14 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 52- Revision 1
Unit Fire Area Description 1 15 Access Control Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 12 MDAFW Pump to 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Generator Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
15 Detection 17 Ionization N N N N N
Suppression WPS-20 Wet Pipe N N N N N
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 54- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 55- Revision 1
Unit Fire Area Description 2 16 Train B Event Monitoring Equipment Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG
Process Monitoring Train B could be affected, the following Train A Process Monitoring is credited RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Generator Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 56- Revision 1
Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
16 Detection 50 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 57- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 58- Revision 1
Unit Fire Area Description 2 17 Unit 2 Normal Switchgear Room & Control Rod Drive Room Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW or 12 MDAFW Pump to SG-11 or 12 Unit 2 - 21 MDAFW Pump to SG-21 or SG-22
Process Monitoring* RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 59- Revision 1
(LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room or use Alternate Rod Insertion. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 60- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
17 Detection 88 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 61- Revision 1
Unit Fire Area Description 1, 2 18 Relay and Cable Spreading Room Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach Note: One train may be affected but the redundant train is likely to remain available.
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
VFDR-018-1-03 VFDR-018-2-03
Process Monitoring Note: Unit 1 and 2, one train of process monitoring could be affected but the redundant train remains available. RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp
VFDR-018-1-05 VFDR-018-2-05
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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(LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Train A or B RCS Head Vents or Pressurizer Vents. Unit 2 - Charging System (Train A) or Safety Injection (Train B) Train A or B RCS Head Vents or Pressurizer Vents.
Reactivity Control Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
VFDR-018-1-01 VFDR-018-2-01
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B, or D1 Emergency Diesel Generator supplying Electrical Distribution Train A Unit 2 - Offsite Power or supplying Electrical Distribution Trains A or B, or D5 Emergency Diesel Generators supplying Electrical Distribution Train A
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Variances from Deterministic Requirements (VFDR)
VFDR-018-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling
water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Criteria is not met for Vital Auxiliary Systems. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems. Components and Cables: Many
Compliant Case: Train A and Train B Cooling Water Strainers (CV-31652, CV-31653, CV-31654, CV-31655, MTR-111C-21, MTR-111C-22, MTR-121C-21 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s):
Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix B. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-137. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-018-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Reactivity Control. This could cause a loss of Reactivity Control. The Nuclear Safety Performance Criteria is not met for Reactivity Control. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Reactivity Control. Components and Cables: Many
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Compliant Case: The ability to maintain Reactivity Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment C) to manually trip Unit 1 Turbine at the Front Standard and close MV-32006 and MV-32010 (Attachment B). This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-018-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain RCS Inventory Control. This could cause a loss of Inventory Control. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Inventory and Pressure Control. Components and Cables: Many Compliant Case: The ability to maintain RCS Inventory and Pressure Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in F5 Appendix B (Attachments A, B, E, and F) to isolate Letdown, de-energize Sump B valves, de-energize the Containment Spray Pump, locally trip the Reactor Coolant and actions to re-align 12 Charging Pump suction to the RWST and restart a charging pump. Evaluate Recovery Actions to de-energize and manually close MV-32084 and MV-32085. Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after suction from the RWST is restored to inject borated water into the RCS. Modification identified in Table S-2, Item #27 will provide a means to isolate Pressurizer PORVs, Excess Letdown, Head Vents, Pressurizer Vents, and Pressurizer Heaters prior to room evacuation.
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This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and plant modifications credited.
VFDR-018-1-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Decay Heat Removal. This could cause a loss of Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Decay Heat Removal. Components and Cables: Many Compliant Case: The ability to maintain Decay Heat Removal should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment I) to manually to start the 11 TDAFWP and align to 11 SG. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-018-1-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to support Post Fire Safe Shutdown. This could cause a loss of Cooling Water. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems. Components and Cables: Many Compliant Case: The ability to maintain Vital Auxiliaries should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Modification identified in Table S-2, Item #13 will ensure hot shorts on cables for CV-31505 which supplies cooling water to D1 Emergency Diesel Generator are isolated when the control switch is placed in local. Modification identified in Table S-2, Item #14 will ensure the Diesel Driven Cooling Water Pumps remain available to provide cooling water to D1 Emergency Diesel Generator and AFW pumps. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-018-1-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Process Monitoring. Steam Generator Level, Pressurizer Level, RCS Pressure, RCS Hot Leg and Cold Leg temperature, and Source Range Neutron Flux This represents a loss of Process Monitoring IndicationProcess Monitoring indication in the Control Room and would require abandonment of the Control Room and Alternate Shutdown. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Process Monitoring. Compliant Case: One train of Process Monitoring indication should remain available in the Control Room.
Disposition Recovery Action Evaluate risk of recovery actions to locally read indication at Hot Shutdown Panels for RCS Pressure, Hot/Cold leg Temperature, Pressurizer Level, SG Level, and Source Range Neutron Flux. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-018-1-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of vital AC Power from Bus 15 and Bus 16. This could cause a loss of Vital AC Power. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital AC Power.
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Components and Cables: Many Compliant Case: The ability to maintain Vital AC Power should remain available from the Control Room.
Disposition
Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment F) to locally start D1 Emergency Diesel Generator and manually align to BUS 15 to restore Vital AC Power. Modification identified in Table S-2, Item #34 will install a synchronization relay that will prevent non-synchronous parallels in all fire areas with exception to areas the gear is located in for the normal, alternate, bus tie, and diesel generator source breakers for BUS 15, BUS 16, BUS 25 and BUS 26. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-018-1-07
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of the 11 and 12 Component Cooling Water Pumps. This could impact the ability to cool the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. Components and Cables: None Compliant Case:
Disposition VFDR-018-2-01
Cooling to the RCP seals should remain unaffected by a fire in this area. Recovery Action(s): None Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited. This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Reactivity Control. This could cause a loss of Reactivity Control.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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The Nuclear Safety Performance Criteria is not met for Reactivity Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Reactivity Control. Components and Cables: Many Compliant Case: The ability to maintain Reactivity Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment B and C) to manually trip Unit 2 Turbine at the Front Standard, and close steam valves MV-32021 and MV-32022. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-018-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain RCS Inventory Control. This could cause a loss of Inventory Control. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Inventory Control. Components and Cables: Many Compliant Case: The ability to maintain RCS Inventory and Pressure Control should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in F5 Appendix B (Attachments A, B, D, and G) to isolate Letdown, de-energize Sump B Valves, de-energize Containment Spray Pump, locally trip the Reactor Coolant Pumps and actions to re-align 22 Charging
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Pump suction to the RWST and restart a charging pump. Evaluate Recovery Actions to de-energize and manually close MV-32178 and MV-32179. Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after suction from the RWST is restored to inject borated water into the RCS. Modification identified in Table S-2, Item #27 will provide a means to isolate Pressurizer PORVs, Excess Letdown, Head Vents, Pressurizer Vents, and Pressurizer Heaters prior to room evacuation. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and plant modifications credited.
VFDR-018-2-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to maintain Decay Heat Removal. This could cause a loss of Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a Lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Decay Heat Removal. Components and Cables: Many Compliant Case: The ability to maintain Decay Heat Removal should remain available from the Control Room.
Disposition Recovery Action(s): Evaluate Recovery Actions performed in procedure F5 Appendix B (Attachment D) to manually start the 22 TDAFWP and align to 21 Steam Generator. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-018-2-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of equipment required to support Post Fire Safe Shutdown. This could cause a loss of Cooling Water. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxiliary Systems. Components and Cables: Many Compliant Case: The ability to maintain Vital Auxiliaries should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #14 will ensure the Diesel Driven Cooling Water Pumps remain available to provide cooling water to D1 Emergency Diesel Generator and AFW pumps. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-018-2-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Process Monitoring. Steam Generator Level, Pressurizer Level, RCS Pressure, RCS Hot Leg and Cold Leg temperature, and Source Range Neutron Flux This represents a loss of Process Monitoring IndicationProcess Monitoring indication in the Control Room and would require abandonment of the Control Room and Alternate Shutdown. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Process Monitoring. Compliant Case: One train of Process Monitoring indication should remain available in the Control Room.
Disposition Recovery Action: Evaluate risk of recovery actions to locally read indication at Hot Shutdown Panels for RCS Pressure, Hot/Cold leg Temperature, Pressurizer Level, SG Level, and Source Range Neutron Flux. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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VFDR-018-2-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of vital AC Power from Bus 25 and Bus 26. This could cause a loss of Vital AC Power. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital AC Power. Components and Cables: Many Compliant Case: The ability to maintain Vital AC Power should remain available from the Control Room.
Disposition
VFDR-18-2-07
Recovery Action(s): Evaluate risk of recovery actions performed in procedure F5 Appendix B (Attachment G) to locally start D5 Emergency Diesel Generator and manually align to BUS 25 to restore Vital AC Power. Modification identified in Table S-2, Item #34 will install a synchronization relay that will prevent non-synchronous parallels in all fire areas with exception to areas the gear is located in for the normal, alternate, bus tie, and diesel generator source breakers for BUS 15, BUS 16, BUS 25, and BUS 26 This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited. This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of the 21 and 22 Component Cooling Water Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxilliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3. due to a lack of separation between redundant equipment required to perform the Nuclear Safety Performance Criteria for Vital Auxilliaries. Components and Cables: None Compliant Case: Cooling to the RCP seals should remain unaffected by a fire in this area.
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Disposition
Recovery Action (s): None Modification identified in Table S-1, Item #1 has installed new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a modification credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
18 Detection 12, 14 Ionization N N N Y N
Detection VEWSD N N N Y N Modification identified in Table S-2, Item #5 will install a Very Early Warning Smoke Detection System (Incipient)
Suppression CO2 N N N Y N Total Flooding
Hose Station - - N N N Y N Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic CO2 fire suppression system is installed in the fire area. The CO2 system was designed and installed in accordance with NFPA 12, Standard on Carbon Dioxide Extinguishing Systems, 1972. Chapter 1123. states “Some of the more important types of hazards and equipment that carbon dioxide systems may satisfactorily protect include:
1. Gaseous and liquid flammable materials. 2. Electrical hazards such as transformers, oil switches and circuit breakers, and rotating equipment. 3. Engines utilizing gasoline and other flammable fuels. 4. Ordinary combustibles such as paper, wood and textiles. 5. Hazardous solids”.
Fire Area 18 contains predominantly cable insulation, plastic, and ordinary combustibles, therefore, no damage to equipment relied on to achieve the Nuclear Safety Performance Criteria goals from the discharge of the system is expected.
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Firefighting water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
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Unit Fire Area Description 1 20 Unit 1, 4.16 kV Safeguards Switchgear (Bus 16) Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.3.2 – Deterministic Approach
Unit 1 - 11 TDAFW Pump to 11 SG Unit 2 - 21 MDAFW Pump to 21 SG
See VFDR020-1-07
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-427) Pressurizer Level (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
See VFDR020-1-03 See VFDR020-1-05
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
See VFDR020-1-02
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump to inject borated water from the RWST
Vital Auxiliaries Unit 1 - D1 supplying Electrical Distribution Train A Unit 2 - Offsite Power (2RY) supplying Electrical Distribution Train A
See VFDR020-1-01
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit 1 - CC Train A
Unit 2 - CC Train A
CL Train A or B
Reference Documents
Safe/Genesis V 4.0.2 EC 23581, Fire Risk Evaluation, Fire Area 20, Unit 1, 4.16kV Safeguards Switchgear (Bus 16), Rev. 1, March 2014 Licensing Actions
This Variance From Deterministic Requirements (VFDR) is due to fire damage that could cause a loss of power to Bus 16, a loss of power from the offsite power sources (CT11 and 1RY), and spurious operation of BKR 15-3 which could prevent Bus 15 from being powered by BKR 15-2 (D1 Emergency Diesel Generator).
The Nuclear Safety Performance Criteria is not met for Vital AC Power. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due to lack of separation between redundant trains of safeguards AC power. Components and Cables: Bus 15 Source from 1R XFMR, BKR-15-3 (1C-419) Bus 16 4.16KV Switchgear, Bus 16 (Many) Compliant Case: BKR-15-3, 1RY source to BUS-15, should remain available from the control room.
Disposition Recovery Action(s): No recovery actions credited.
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Modification identified in Table S-2, Item #6 will re-wire cable 1C-419 so that fire damage to cable 1C-419 will not spuriously operate, spurious opening, BKR 15-3. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-020-1-02
Disposition
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of power to PNL 113 which powers CV-31198, charging line to 11 Regenerative Heat Exchanger, which would cause CV-31198 to fail open. If CV-31198 failed open, it could cause a flow diversion away from the RCP seals. A fire in this area could also affect train B Safeguards Bus 16 power supply to the 12 Component Cooling Water Pump, could impact cable 1C-419 which could impact the Train A Safeguards Bus 15 power supply to the 11 Component cooling Water Pump which provides cooling to the RCP TBHX. Loss of all seal cooling to the RCP seals could lead to increased leakage from the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory Control. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due to lack of separation between redundant trains of RCP seal cooling. Components and Cables: Instrument Bus III (Blue) Panel 113, PNL-113 (1CX-99) Bus 16 4.16KV Switchgear, Bus 16 (Many) Bus 15 Source from 1R XFMR, BKR-15-3 (1C-419) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #6 will re-route cable 1CX-99 out of Fire Area 20 and into Fire Area 58 so PNL 113 will remain available to support safe shutdown. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-020-1-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of power to Train B Bus 16 and loss of power from 13 Inverter to Train A PNL 113 which powers control room indication for 1N51 Source Range Monitor, 1T-450A and 1T-450B for RCS Hot and Cold Leg Temperature Indication. This could cause a loss of indication for Source Range Neutron Flux Monitoring, and RCS Hot and Cold Leg Temperature Indication in the Control Room.
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The NFPA 805 Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due to lack of separation between redundant trains of process monitoring. Components and Cables: Instrument Bus III (Blue) Panel 113, PNL-113 (1CX-99) Instrument Bus IV (Yellow), PNL-114 (1CY-99) Compliant Case: Process Monitoring indication for RCS Hot and Cold Leg, and Source Range should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #6 will re-route cable 1CX-99 out of Fire Area 20 and into 58 so PNL 113 will remain available to support Safe Shutdown Process Monitoring. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-020-1-05 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of power to Train B Bus 16 and loss of power from 11 Inverter to Train A PNL 111 which powers control room indication for 1L-487, 11 SG Wide Range Level indication. This could cause a loss of indication for 11 Steam Generator Level Indication in the Control Room. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due to lack of separation between redundant trains of process monitoring. Components and Cables: Instrument Bus II (White) Panel 111, PNL-111 (1CW-99) Instrument Bus I (Red) Panel 112, PNL-112 (1CR-99) Compliant Case: Process Monitoring indication for 11 Steam Generator Level Indication should remain available from the Control Room.
Disposition Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Modification identified in Table S-2, Item #6 will re-route cable 1CW-99 out of Fire Area 20 and into 58 so PNL 111 will remain available to support Safe Shutdown Process Monitoring. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-020-1-07 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-16-3, 12 MDAFWP, and loss of power to BUS-121 which powers MCC-1A2 which powers MV-32381 and MV-32382. A loss of power to MV-32381 and MV-32382 would prevent remote closure of the valves and the inability to isolate spurious AFW flow. This could eventually lead to an over-fill of the Steam Generator which would challenge the Decay Heat Removal Nuclear Safety Performance Criteria. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due to lack of separation between redundant trains of decay heat removal due to over-fill of steam generator. Components and Cables: 12 MDAFW Pump, MTR 16-3 (16403-A, 16403-C, 16403-D, 16403-E) Bus 121 480V Switchgear, BUS-121 (16404-1, 16404-A, 16404-C) Compliant Case: MTR-16-3, 12 MDAFWP should not spuriously start and over-fill the 11 and 12 Steam Generators.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of over-filling the Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
20 Detection 11 Ionization N N N Y N
Suppression - - - - - - -
Hose Station - - N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 21 Unit 1, 4.16 kV Normal Switchgear (Bus 13, 14) Note: Fire Area 21 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 81- Revision 1
Unit Fire Area Description 1 22 480 V Safeguards Switchgear (Bus 121) Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW Pump to 11 SG Unit 2 - 21 MDAFW Pump to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A Unit 2 - Offsite Power supplying Electrical Distribution Train A
VFDR-22-0-01
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 82- Revision 1
Unit 1 - CC Train A
Unit 2 - CC Train A
CL Train A
Reference Documents
Safe/Genesis V 4.0.2 EC 23582, Fire Risk Evaluation, Fire Area 22, 480V Safeguards Switchgear (Bus 121), Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
VFDR-022-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 cooling water strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect that function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3, due a lack of separation between redundant trains of cooling water strainers. Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CB-374) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CB-374) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CB-374) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CB-374) 11 Cooling Water Strainer Motor, MTR-111C-21 (1CB-374) 12 Cooling Water Strainer Motor, MTR-121C-21 (1CB-374) 21 Cooling Water Strainer Motor, MTR-111C-22 (1CB-374) 22 Cooling Water Strainer Motor, MTR-121C-22 (1CB-374) AC Distribution Panel 136, PNL-136 (1CB-374)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 83- Revision 1
Compliant Case: Train A, 11 Cooling Water Strainer (CV-31652 and MTR-111C-21) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-136. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
22 Detection 43 Ionization N N N Y N
Suppression - - - - - - -
Hose Station - - N N N Y N Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 84- Revision 1
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 85- Revision 1
Unit Fire Area Description 2 23 Unit 2, 4.16 KV Normal Switchgear (Bus 23, 24)
Note: Fire Area 23 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 86- Revision 1
Unit Fire Area Description 1, 2 24 Oil Storage Area Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 87- Revision 1
(LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 88- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
24 Detection 4 Ionization, Flame, Heat
N N N N N
Suppression DA-2 Deluge N N N N N
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
There is an automatic fire suppression system in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 89- Revision 1
Unit Fire Area Description 1 25 Diesel Generator 1 Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 12 MDAFW Pump to 12 SG
Unit 2 - 22 TDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 90- Revision 1
to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Unit 1 - CC Train B
Unit 2 - CC Train A or B
CL Train B
VFDR-025-0-01
Reference Documents EC 23583, Fire Risk Evaluation, Fire Area 25, Diesel Generator 1 Room, Rev. 1, March 2014.
Safe/Genesis V 4.0.2 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-025-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious close of BKR 15-2 and parallel D1 to Bus 15 out of phase. This could cause a lockout of Bus 15 which powers PNL 136 which powers the Cooling Water Strainers backwash central control panel for 11, 12, 21, and 22 CL Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 91- Revision 1
This condition would challenge the Nuclear Safety Performance Criteria for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. This is a separation issue for Vital Auxiliaries. Components and Cables: BKR 15-2, (15402-G, 15402-K, 1CA-1140, 1CA-1142) Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 82. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
25 Detection 82 Ionization, Heat, Flame
N N Y N Y
Suppression PA-1 Pre-Action N N Y Y N
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will exit through the floor drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 92- Revision 1
The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 93- Revision 1
Unit Fire Area Description 1 26 Diesel Generator 2 Room
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 94- Revision 1
(LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A and B Unit 2 - Offsite Power supplying Electrical Distribution Trains A and B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 95- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
26 Detection 6 Ionization, Heat, Flame
N N N N N
Suppression PA-1 Pre-Action N N N Y N
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will exit through the floor drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 96- Revision 1
Unit Fire Area Description 1 27 Water Conditioning Equipment Area
Note: Fire Area 27 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 97- Revision 1
Unit Fire Area Description 1,2 28 Transformers
Fire Area 28 includes Fire Area(s): 28a Main Transformer (Unit 1) 28b Main Transformer (Unit 2) 28c 1R Transformer 28d 1M Transformer 28e 2M Transformer 28f 2RX/Y Transformer Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 98- Revision 1
(LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train A or B Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Train A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 99- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
28a Detection 58 HAD N N N Y N
Suppression DM-1 Deluge N N N N N Manual Actuation
Feature - - - - - - -
28b Detection 60 HAD N N N Y N
Suppression DM-5 Deluge N N N N N Manual actuation
Feature - - - - - - - 28c Detection 62 HAD N N N Y N
Suppression DM-3 Deluge N N N N N Manual activation
Feature - - - - - - - 28d Detection 59 HAD N N N Y N
Suppression DM-2 Deluge N N N N N Manual activation
Feature - - - - - - -
28e Detection 61 HAD N N N Y N
Suppression DM-4 Deluge N N N N N Manual activation Feature - - - - - - -
28f Detection 96 HAD N N N Y N
Suppression DM-6 Deluge N N N N N
Feature - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or
Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
A fire suppression system is installed in the fire area. The area around each transformer is diked, with drainage to nearby pits. The only electrical equipment is the transformers, which are mounted on pedestals. However, the main source of fire, and consequently either manual suppression system actuation or hose stream application, is the transformers and water would be applied to the fire. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 100- Revision 1
Fire Area Comments
Note: Fire Area 28 a – f is now combined into Fire Area 28.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 101- Revision 1
Unit Fire Area Description 1 29 Administration Building Elect & Piping Room #1
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 102- Revision 1
(LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection System (Train B) Unit 2 - Charging System (Train A) or Safety Injection System (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A and B Unit 2 - Offsite Power supplying Electrical Distribution Trains A and B
Unit 1 - CC Train B
Unit 2 - CC Train B
CL Train B
VFDR-029-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23584, Fire Risk Evaluation, Fire Area 29, Administration Building Elect & Piping Room # 1, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-029-0-01 This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 103- Revision 1
would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pump. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CA-529) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CA-529) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CA-529) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CA-529) 11 Cooling Water Strainer Motor, MTR-111C-21 (1CA-529) 12 Cooling Water Strainer Motor, MTR-121C-21 (1CA-529) 21 Cooling Water Strainer Motor, MTR-111C-22 (1CA-529) 22 Cooling Water Strainer Motor, MTR-121C-22 (1CA-529) Motor Control Center 1AB, Bus 1, (111C-4, 111C-5) Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 4. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
29 Detection 4 Ionization N N Y N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 29 extends over Fire Area 24.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 30 Administration Building Elect & Piping Room #2
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW Pump to 11 SG or 12 SG
Unit 2 - 21 MDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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(LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A and B Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Trains A and B Unit 1 - CC Train A Unit 2 - CC Train A CL Train A
VFDR-030-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23585, Fire Risk Evaluation, Fire Area 30, Administration Building Elect & Piping Room # 2, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-030-0-01 This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pump.
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The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CB-370) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CB-370) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CB-370) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CB-370) 11 Cooling Water Strainer Motor, MTR-111C-21 (1CB-370) 12 Cooling Water Strainer Motor, MTR-121C-21 (1CB-370) 21 Cooling Water Strainer Motor, MTR-111C-22 (1CB-370) 22 Cooling Water Strainer Motor, MTR-121C-22 (1CB-370) AC Distribution Panel 136, PNL-136 (1CB-370) Motor Control Center 1AB Bus 2, MCC-1AB2 (221C-4) Compliant Case: Train A, 11 Cooling Water Strainer (CV-31652 and MTR-111C-21) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 4. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-136. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
30 Detection 40 Ionization N N Y N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 30 extends over the Oil Storage Room (FA 24)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 31 “A” Train Hot Shutdown Panel & Air Compressor/Aux. Feedwater Pump Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 – 11 TDAFW Pump to 11 SG Unit 2 – 21 MDAFW Pump to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-427) Pressurizer Level (LOOP 2L-427) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
VFDR-031-2-02
Inventory and Pressure Control Unit 1 – Charging System (Train A) Unit 2 – Charging System (Train A)
Reactivity Control Unit 1 – Trip reactor from the Control Room. Use Charging to inject borated water from the RWST Unit 2 – Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 – Offsite Power (1RY) supplying Electrical Distribution Train A and portions VFDR-031-1-03
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of Train B Unit 2 – Offsite Power (CT 12) supplying Electrical Distribution Train A and portions of Train B Unit 1 – CC Train A Unit 2 – CC Train A CL Train A
VFDR-031-2-03
Reference Documents
Safe/Genesis V 4.0.2 EC 23586, Fire Risk Evaluation, Fire Area 31, A Train Hot Shutdown Panel & Air Compressor/Auxiliary Feedwater Room, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
FPEE-12-004, CA-01313808-01, AFW Pump Room Ducts without Fire Dampers
Summary The purpose of this evaluation is to assess the impact on fire safe shutdown capability of four exhaust ducts and one supply duct that are not provided with 3-hour rated fire dampers in the boundaries of the Auxiliary Feedwater Pump Rooms, Fire Areas 31 and 32. The lack of fire dampers in the ductwork penetrations of the common barrier between the AFW Pump Rooms and the 480V Normal Switchgear Rooms, and between the 480V Normal Switchgear Rooms to the Turbine Buildings, has no impact on safe shutdown capability and is acceptable as is.
EEEE Title CA-01040686 (Attachment), NFPA 72E Requirement for Non-Restorable Heat Detectors Evaluation
Summary The 2006 NFPA 72E, 1974 edition, code compliance review identified that nonrestorable spot type thermal detectors were not being tested in accordance with the code. Based on the type of fire expected in these fire areas, the generation of smoke will exceed the generation of heat during the incipient stage of a fire. Therefore, the area smoke detection will provide the early warning of a fire. The heat detectors are not credited in the USAR by reference to F5 Appendix K, but will provide a secondary level of automatic detection. Based on the lack of credit given to the heat detectors, the requirements of NFPA 72E are not mandatory and the functional testing for the rate of rise function is adequate. The additional testing of the fixed temperature function through destructive testing at a nationally recognized laboratory is not cost justified for the minor benefit. The lack of meeting the full code requirement for non-restorable heat detection testing will not have adverse effect on safe shutdown for any of the affected fire areas.
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is. Variances from Deterministic Requirements (VFDR)
VFDR-031-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function and loss of power to PNL-136 due to lack of coordination. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect that function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to the lack of separation between redundant trains of cooling water strainers. Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CB-370, 1CB-374) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CB-370, 1CB-374) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CB-370, 1CB-374) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CB-370, 1CB-374) 11 Cooling Water Strainer Motor, MTR-111C-21 (1CB-370, 1CB-374) 12 Cooling Water Strainer Motor, MTR-121C-21 (1CB-370, 1CB-374) 21 Cooling Water Strainer Motor, MTR-111C-22 (1CB-370, 1CB-374) 22 Cooling Water Strainer Motor, MTR-121C-22 (1CB-370, 1CB-374) AC Distribution Panel 136, PNL-136 (1CB-370, 1CB-374) Motor Control Center 1AB Bus 2, MCC-1AB2 (221C-4)
Compliant Case: Train A, 11 Cooling Water Strainer (CV-31652 and MTR-111C-21) should remain unaffected by a fire.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 2. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-136. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-031-0-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of both trains of AFW. An automatic wet pipe suppression system is installed in the general area of the AFW pump room, but does not provide full area wide coverage in the mezzanine area above the battery rooms. There are no fixed fire initiators in the mezzanine area
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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above the battery rooms. There are some Fire PRA cables in the mezzanine area. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of a full area suppression system with credit given to deterministic requirements of section 4.2.3.3 b and c.
Components and Cables:
None
Compliant Case:
The automatic wet pipe fire suppression system in the AFW pump room should provide full area wide coverage to meet deterministic requirements.
Disposition Recovery Action(s): No recovery actions credited. This fire area meets the Performance Based requirements of NFPA 805, Section 4.2.4 and only credits the automatic fire suppression system in the general area where it is installed.
VFDR-031-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 11 TDAFWP (Train A) discharge control valves MV-32238 and MV-32239 and damage 12 MDAFWP (Train B). A fire at the Train A Hot Shutdown Panel could cause spurious closure of MV-32238 or MV-32239 which would prevent 11 TDAFWP from supplying Aux Feedwater to 11 or 12 Steam Generator. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of Aux Feedwater. Components and Cables: 11 AFW to 11 SG MV, MV-32238 (1CA-115) 11 AFW to 12 SG MV, MV-32239 (1CA-116) Compliant Case: Train A, 11 TDAFWP, along with valves MV-32238, and MV-32239 should remain unaffected by a fire to provide AFW to a SG for Decay Heat Removal.
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Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #1 will relocate control switches CS-51003 and CS-51005 so that 11 TDAFWP discharge valves are not affected by a fire in FA 31. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-031-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-16-3, 12 MDAFWP, and loss of power to MCC-1A2 which powers MV-32381 and MV-32382. A loss of power to MV-32381 and MV-32382 would prevent remote closure of the valves and the inability to isolate spurious AFW flow. This could eventually lead to an over-fill of the Steam Generator which would challenge the Decay Heat Removal Nuclear Safety Performance Criteria. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of Aux Feedwater. Components and Cables: 12 MDAFW Pump, MTR 16-3 (16403-C, 1CB-31, 1CB-920) Motor Control Center 1A Bus 2, MCC 1A2 (121E-1) Compliant Case: The 12 MDAFWP should not spuriously start and over-fill the 11 and 12 Steam Generators.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of over-filling the Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-031-1-03 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power to Bus 16 and damage to power cables that could over-heat and cause secondary fires. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
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This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital Auxiliaries.
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition Recovery Action(s):
No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure protection of the over-current trip function of 4kV breakers and preclude secondary ignition of cables in other fire areas. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-031-2-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 21 MDAFWP (Train A) discharge control valves and damage to 22 TDAFWP (Train B). A fire at the Train A Hot Shutdown Panel could cause spurious closure of MV-32383 or MV-32384 which would prevent 21 MDAFWP from supplying Aux Feedwater to 21 or 22 Steam Generator. A fire at MCC-2A1 could cause spurious operation of MV-32383, MV-32384, MV-32026, or MV-32336. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of Aux Feedwater.
Components and Cables:
Motor Control Center 2A Bus 1, MCC 2A1 (211E-1) 21 MDAFW Pump to 21 SG, MV-32383 (2A1-5, 2A1-5A, 2CA-115, 2CA-116, 2CA-65) 21 MDAFW Pump to 22 SG, MV-32384 (2A1-5A, 2A1-6, 2CA-116, 2CA-117, 2CA-66) 21 MDAFW Pump suction from CL, MV-32026 (2A1-2, 2A1-2A, 2A1-4A, 2CA-30) 21 MDAFW Pump suction from CST, MV-32336 (2A1-4A, 2CA-30)
Compliant Case:
Train A, 21 MDAFWP, along with valves MV-32383, MV-32384, MV-32026, and MV-32336 should remain unaffected by a fire to provide AFW to a Steam Generator for Decay Heat Removal.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #2 will ensure that 21 MDAFWP (Train A) remains available to provide AFW to 21 Steam Generator. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a modification credited.
VFDR-031-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of Pressurizer Level Indication, Steam Generator Level Indication, and Source Range Flux Monitoring from the Control Room. Loss of Pressurizer and Steam Generator Level Indication in the Control Room could lead to SG over-fill and loss of Decay Heat Removal function. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains pressurizer level indication. Components and Cables: Pressurizer Level Indication, LOOP 2L-433 (1C-5118, 1CA-1106, 2CF-207, 2CF-214) Instrument Bus II (White) Panel 211, PNL-211 (2CW-1) Instrument Bus III (Blue) Panel 213, PNL-213 (2CX-1) Interruptable Panel 217, PNL-217 (2AC1-5, 2CV-35, 2CV-6) Motor Control Center 2AC Bus 2, MCC-2AC2 (221F-1) Instrument Bus I (RED) Panel 212, PNL-212 (2CR-1) Instrument Bus IV (YELLOW) Panel 214, PNL-214 (2CY-1)
Compliant Case:
Train A Process Monitoring indication should remain available from the Control Room without the need for Recovery Actions.
Disposition Recovery Action(s): Evaluate risk of recovery actions to re-power PNL-211 and PNL-213 from PNL-217 to restore Pressurizer Level (2L-427) , 21 SG Level (2L-487) level indication, and source Range Neutron Flux indication from the control room. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-217. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-031-2-03 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power to Bus 24 and damage to power cables that could over-heat and cause secondary fires. If the fire damaged DC control power,
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and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital Auxiliaries.
Components and Cables:
Bus 24 Feed Breaker to 204-206-209-402 Transformer, BKR-24-6 (24406-3, 24403-F, 24404-F, 2C-2552, 2C-866, 2DC-5)
Compliant Case:
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition Recovery Action(s):
No recovery actions credited.
Modification identified in Table S-2, Item #10 will ensure protection of the over-current trip function of 4kV breakers and preclude secondary ignition of cables in other fire areas.
This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
31 Detection 2 Ionization N N Y Y N
Suppression WPS-10 Wet Pipe N N Y Y N
Feature See Note ERFBS N N N Y N Cables 2AC1-5, 1CA-115, 2A1-4A, 2A1-5A, 2CA-30, 2CA-115, 2CA-116, 2CA-117, TB-2390 and 2SG-TA11
Hose Station N N N Y N Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
FA 31 extends above Fire Areas 35 and 36
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 32 "B" Train Hot Shutdown Panel & Air Compressor/Aux. Feedwater Pump Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Safety Injection (Train B) Unit 2 - Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 119- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power (1 RY) supplying Electrical Distribution Train B Unit 2 - Offsite Power supplying Electrical Distribution Train B
Safe/Genesis V 4.0.2 EC 23587, Fire Risk Evaluation, Fire Area 32, B Train Hot Shutdown Panel & Air Compressor/Auxiliary Feedwater Room, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
FPEE-12-004, CA-01313808-01, AFW Pump Room Ducts without Fire Dampers
Summary The purpose of this evaluation is to assess the impact on fire safe shutdown capability of four exhaust ducts and one supply duct that are not provided with 3-hour rated fire dampers in the boundaries of the Auxiliary Feedwater Pump Rooms, Fire Areas 31 and 32. The lack of fire dampers in the ductwork penetrations of the common barrier between the AFW Pump Rooms and the 480V Normal Switchgear Rooms, and between the 480V Normal Switchgear Rooms to the Turbine Buildings, has no impact on safe shutdown capability and is acceptable as is.
EEEE Title CA-01040686 (Attachment), NFPA 72E Requirement for Non-Restorable Heat Detectors Evaluation
Summary The 2006 NFPA 72E, 1974 edition, code compliance review identified that nonrestorable spot type thermal detectors were not being tested in accordance with the code. Based on the type of fire expected in these fire areas, the generation of smoke will exceed the generation of heat during the incipient stage of a fire. Therefore, the area smoke detection will provide the early warning of a fire. The heat detectors are not credited in the USAR by reference to F5 Appendix K, but will provide a secondary level of automatic detection. Based on the lack of credit given to the heat detectors, the requirements of NFPA 72E are not mandatory and the functional testing for the rate of rise function is adequate. The additional testing of the fixed temperature function through destructive testing at a nationally recognized laboratory is not cost justified for the minor benefit. The lack of meeting the full code requirement for non-restorable heat detection testing will not have adverse effect on safe shutdown for any of the affected fire areas.
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3,
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-032-0-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety
Performance Criteria is not met for Vital Auxiliaries.
This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers.
Components and Cables:
MCC 1AB1 (111C-4)
11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CA-529, 1CA-538)
12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CA-529, 1CA-538)
21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CA-529, 1CA-538)
22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CA-529, 1CA-538)
11 Cooling Water Strainer Motor, MTR-111C-21 (1CA-529, 1CA-538)
12 Cooling Water Strainer Motor, MTR-121C-21 (1CA-529, 1CA-538)
21 Cooling Water Strainer Motor, MTR-111C-22 (1CA-529, 1CA-538)
22 Cooling Water Strainer Motor, MTR-121C-22 (1CA-529, 1CA-538)
Compliant Case:
Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 2. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-032-0-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of both trains of AFW. An automatic wet pipe suppression system is installed in the general area of the AFW pump room, but does not provide full area wide coverage in the mezzanine area above the battery rooms. There are no fixed fire initiators in the mezzanine area above the battery rooms. There are some Fire PRA cables in the mezzanine area.
The Nuclear Safety Performance Criteria is not met for Decay Heat Removal.
This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of a full area suppression system with credit given to deterministic requirements of section 4.2.3.3 b and c.
Components and Cables:
None
Compliant Case:
The automatic wet pipe fire suppression system in the AFW pump room should provide full area wide coverage to meet deterministic requirements.
Disposition Recovery Action(s): No recovery actions credited. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with automatic fire suppression system in the general area credited.
VFDR-032-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 11 TDAFWP (Train A) and damage to circuits for 12 MDAFWP (Train B) suction and discharge valves. A fire at the Train B Hot Shutdown Panel could cause spurious operation of MV-32381 or MV-32382. A fire at MCC-1A2 could cause spurious operation of MV-32381, MV-32382, MV-32027, or MV-32335.
The Nuclear Safety Performance Criteria is not met for Decay Heat Removal.
This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
The NFPA 805 Nuclear Safety Performance Goal Criteria is not met for Decay Heat Removal.
Compliant Case:
Train B, 12 MDAFWP, along with valves MV-32381, MV-32382, MV-32027, and MV-32335 should remain unaffected by a fire to provide AFW to a SG for Decay Heat Removal.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #3 will ensure that 12 MDAFWP (Train B) remains available to provide AFW to 12 Steam Generator. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-032-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of power to BUS-15 and BUS-16 from the offsite power sources 1R, CT11 and D1 to BUS-15 and D2 to BUS-16. Loss of power to BUS-15 and BUS-16 could cause a station black-out on Unit 1 and would not meet Vital Auxiliaries Performance Criteria of NFPA 805. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of safeguard power. Components and Cables: Bus 16 4.16KV Switchgear, BUS 16 (1DCB-1) Bus 16 Source from 1R XFMR, BKR 16-2 (1C-332, 1C-333) Bus 16 Source from Bus CT 11, BKR 16-8 (15407-1, 15407-2, 16408-1)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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D2 Diesel Generator, 034-021 (1DCB-2, 1DCB-95) Bus 15 4.16KV Switchgear, BUS 15 (15406-B, 1DCA-1) Bus 15 Source from D1 Diesel Generator, BKR 15-2 (15402-G, 15402-K, 15402-1) Bus 15 Source from 1R XFMR, BKR 15-3 (1C-332, 1C-333, 15403-B) Bus 15 Source from Bus CT 11, BKR 15-7 (15407-1, 15407-2, 15407-A, 16408-1)
Compliant Case:
BUS-16 should remain energized from 1RY (BKR-16-2) or D2 (BKR-16-9) to provide vital AC power to support safe shutdown.
Disposition
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #6 will re-route cable 1C-333 out of Fire Area 32 so that the 1RY source will remain available to BUS-16. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for the Bus-15 source breakers. If a Bus 15 feeder breaker is faulted the source breakers will remain available to clear the fault and protect the 1RY source. 048 This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with plant modifications credited.
Disposition VFDR-032-1-03
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited. This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power to BUS-13, 14, 15, and 16, which would preclude operation of the over-current trip relay protection on the following 4 KV breakers: BKR-13-3, BKR-14-3, BKR-15-1, BKR-15-4, BKR-15-5, BKR-15-9, and BKR-16-3. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital and Non-Vital AC power.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements. Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-032-2-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 21 MDAFWP (Train A) and damage to circuits for 22 TDAFWP (Train B) suction and discharge valves. A fire at the Train B Hot Shutdown Panel could cause spurious operation of MV-32246 or MV-32247. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal.
This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between
redundant trains of AFW.
Components and Cables:
22 AFW to 21 SG MV, MV-32246 (2CB-164)
22 AFW to 22 SG MV, MV-32247 (2CB-163)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Compliant Case:
Train B, 22 TDAFWP and discharge valves MV-32246 and MV-32247 should remain unaffected by a fire to provide AFW to a SG for Decay Heat Removal.
Disposition Recovery Action: No recovery actions credited. Modification identified in Table S-2, Item #4 will ensure that 22 TDAFWP (Train B) remains available to provide AFW to 22 Steam Generator. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-032-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-25-10, 21 MDAFWP and loss of power MCC-2A1 which powers MV-32383 and MV-32384. A loss of power to MV-32383 and MV-32384 would prevent remote closure of the valves and the inability to isolate spurious AFW flow. This could eventually lead to an over-fill of the Steam Generator which would challenge the Decay Heat Removal Nuclear Safety Performance Criteria. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, lack of separation between redundant trains of decay heat removal due to over-fill of steam generator. Components and Cables 21 MDAFW Pump, MTR-25-10 (25410-D, 25410-E, 2CA-505, 2CA-506, 2CA-525, 2CA-778) Motor Control Center 2A Bus 1, MCC-2A1 (211E-1) Compliant Case: The 21 MDAFWP should not spuriously start and over-fill the 21 and 22 Steam Generators.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of over-filling the Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 126- Revision 1
VFDR-032-2-03 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power to BUS-25 which would preclude operation of the over-current trip relay protection on 4 KV breaker BKR-25-10. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital and Non-Vital AC power.
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure protection of the over-current trip function of 4kV breakers and preclude secondary ignition of cables in other fire areas. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
32 Detection 2 Ionization N N Y Y N Suppression WPS-10 Wet Pipe N N Y Y N
Feature ERFBS N N N Y N Cables and trays 16403-1, 16403-C, 1A2-6A, 1A2-7A, 1A2-8A, 1CB-52, 1CB-53, 1CB-55, 1CB-56, 1CR-99, 1CY-99, 2CB-163, 1CB-164, TB-1263 and 1SG-TA11 have supplemental barriers (ERFBS)
Feature RES N N N Y N Cable 1SG-LB22 has a radiant energy shield (RES), Marinite board Hose Station N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
FA 32 extends above Fire Areas 33 and 34
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 33 Battery Room 11
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW or 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train B and portions of Train A Unit 2 - Offsite Power supplying Electrical Distribution Train B and portions of Train A
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
33 Detection 1 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 130- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 32 extends above Fire Area 33.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 34 Battery Room 12 Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW Pump to 11 SG Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 132- Revision 1
inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
34 Detection 1 Ionization N N Y N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 32 extends above Fire Area 34.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 35 Battery Room 21 Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 135- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train B Unit 1 - CC Train A or B Unit 2 - CC Train B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
35 Detection 35 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 136- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 31 extends above Fire Area 35.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 36 Battery Room 22 Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG Unit 2 - 22 TDAFW to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A
Unit 1 CC Train A or B
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 139- Revision 1
the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Fire Area 31 extends above Fire Area 36.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 140- Revision 1
Unit Fire Area Description 1 37 Unit 1, 480 V Normal Switchgear Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 141- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 142- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
Unit 1 CC Train B
Unit 2 CC Train B
CL Train B
VFDR-037-0-01 VFDR-037-1-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23588, Fire Risk Evaluation, Fire Area 37, Unit 1 480V Normal Switchgear Room, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-037-0-01
This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3 due to a lack of
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 143- Revision 1
separation between redundant trains of cooling water strainers.
Components and Cables:
Motor Control Center 1AB Bus 1, MCC 1AB1, (111C-4)
11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CA-529)
12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CA-529)
21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CA-529)
22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CA-529)
11 Cooling Water Strainer Motor, MTR-111C-21 (1CA-529)
12 Cooling Water Strainer Motor, MTR-121C-21 (1CA-529)
21 Cooling Water Strainer Motor, MTR-111C-22 (1CA-529)
22 Cooling Water Strainer Motor, MTR-121C-22 (1CA-529)
Compliant Case:
Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 83. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 144- Revision 1
VFDR-037-1-01
This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause a loss of DC control power to BUS-13 which would preclude operation of the over-current trip relay protection on the 4KV breakers. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital Auxiliaries.
Components and Cables:
Bus 13 Feed Breaker to 103-105-109-301 Transformer, BKR-13-3 (13403-2, 13403-3, 13403-4, 13404-B, 13405-B, 13408-B, 1DC-3)
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 145- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
37 Detection 83 Ionization N N Y Y N
Suppression - - - - - - -
Hose Station - - N N N Y N Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 146- Revision 1
Unit Fire Area Description 2 38 Unit 2, 480 V Normal Switchgear Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW Pump to 11 SG or 12 SG
Unit 2 - 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 147- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 148- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
Unit 1 CC Train A
Unit 2 CC Train A
CL Train A
VFDR-038-0-01 VFDR-038-2-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23590, Fire Risk Evaluation, Fire Area 38, Unit 2 480V Normal Switchgear Room, Rev. 1, March 2014 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-038-0-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3 due to a lack of separation between redundant trains of cooling water strainers.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 149- Revision 1
Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (1CB-370) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (1CB-370) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (1CB-370) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (1CB-370) 11 Cooling Water Strainer, MTR-111C-21 (1CB-370) 12 Cooling Water Strainer, MTR-121C-21 (1CB-370) 21 Cooling Water Strainer, MTR-111C-22 (1CB-370) 22 Cooling Water Strainer, MTR-121C-22 (1CB-370) Motor Control Center 1AB Bus 2 (221C-4) AC Distribution Panel 136, PNL-136 (1CB-370) Compliant Case: Train A, 11 Cooling Water Strainer (CV-31652 and MTR-111C-21) should remain unaffected by a fire.
Disposition
Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 85. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-136. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-038-2-01
This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause a loss of DC control power to BUS-23 and BUS-24 which would preclude operation of the over-current trip relay protection on the 4KV breakers. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements.
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries.
This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital Auxiliaries.
Bus 24 Feed Breaker to 204-206-209-402 Transformer, BKR-24-6 (24406-2, 24406-3, 24403-F, 24404-F, 24405-E, 2DC-5)
Compliant Case:
Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
38 Detection 85 Ionization N N Y Y N
Suppression - - - - - - -
Hose Station - - N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 151- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 152- Revision 1
Unit Fire Area Description 1, 2 39 Radwaste Building
Note: Fire Area 39 is now combined into Fire Area 4.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 153- Revision 1
Unit Fire Area Description 1, 2 40 Maintenance Storage Shed
Note: Fire Area 40 is now combined into Fire Area 4.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 154- Revision 1
Unit Fire Area Description 1, 2 41 Screenhouse (General Area)
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 155- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 156- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
EEEE Title CA 01241917-01, Fire Protection Engineering Evaluation of Fire Barrier between FA 41A and FA 41
Summary The purpose of this Fire Protection Engineering Evaluation is to show that there is adequate separation of redundant trains of equipment required to achieve safe shutdown in the event of a fire in the Screenhouse, Fire Areas 41, 41A, and 41B. There are no significant fire hazards in FA 41 or 41B in the vicinity of the ventilation openings above doors 257 and 258. The redundant cooling water pumps and cables are also well separated from the ventilation openings. There is fire detection and an automatic fire suppression system on both sides of the openings. These defense-in depth features provide the justification that the ventilation openings above doors 257 and 258 are acceptable as is.
EEEE Title CA-01311046-01, Fire Doors 257, 258, 259 & 260
Summary
The purpose of this evaluation is to assess Doors 257, 258, 259 and 260, which do not have an identified fire rating. Doors 257 and 258 also have transoms above the doors that are not fire rated. The doors are in the boundaries between Fire Area 41 and Fire Area 41A on the 695ft elevation of the Screenhouse. Fire Area 41 is open to Fire Area 41B on the 670ft elevation. Fire Area 41A contains safe shutdown capabilities that are redundant to those in Fire Area 41B. Fire Doors 257 and 258, inclusive of the 14ga steel plate transoms, and Fire Doors 259 and 260 provide adequate protection to prevent fire spread that could adversely impact redundant safe shutdown capability. The bases for this conclusion include the following: Fire Doors 257, 258, 259, and 260 are of substantial construction, constructed by the same vendor that built many of the fire doors installed during original plant construction. The doors could be qualified as 1-1/2hr rated fire doors based on the referenced drawings. The transoms above Fire Doors 257 and 258 are 14ga plate steel, securely attached to the openings above the doors. While not fire rated, the transoms will resist the passage of flame, smoke and hot gases. Automatic detection and pre-action sprinklers are provided on both sides of the subject doors in Fire Areas 41A and 41. Automatic detection and pre-action sprinklers are provided throughout Fire Area 41A and Fire Area 41B, which would control the size of postulated fires in the areas. The automatic detection systems provided throughout Fire Area 41B and Fire Area 41A, and on the south side of Fire Area 41, would also result in prompt fire brigade response. Fire brigade response would be effective in controlling and extinguishing postulated fires prior to the fire impacting on redundant fire safe shutdown capability.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 157- Revision 1
VFDR-041-0-01 Dispostion
Variances from Deterministic Requirements (VFDR) This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of coordination of PNL-136 and PNL-137 which supply power to the cooling water strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. This is a separation issue for Vital Auxiliaries. Components and Cables: AC Distribution Panel 136, PNL-136 (1AB1-18) AC Distribution Panel 137, PNL-137 (1AB2-17) Compliant Case: One strainer in each Cooling Water Header (11 or 12 in Train A Header) and (21 or 22 in Train B Header) should remain unaffected by a fire in this area. The Train A and Train B Cooling Water Header need to remain functional because the isolation valves could be affected by the fire. Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-136 and PNL-137. This VFDR has been evaluated and it has been determined to meet the acceptance criteria of NFPA 805 Section 4.2.3 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 158- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
41 Detection 75 Ionization N N Y Y N
Suppression PA-9 Pre-Action N N Y N N
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
There is an automatic fire suppression system in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 159- Revision 1
Unit Fire Area Description 1, 2 41A Screenhouse (DDCLP Rooms)
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 160- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 161- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Trains A or B Unit 2 - Offsite Power supplying Electrical Distribution Trains A or B
Unit 1 CC Train A or B
Unit 2 CC Train A or B
CL Train A or B
VFDR-041A-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23591, Fire Risk Evaluation, Fire Area 41A, Screenhouse (DDCLP Rooms), Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title CA 01241917-01, Fire Protection Engineering Evaluation of Fire Barrier between FA 41A and FA 41
Summary The purpose of this Fire Protection Engineering Evaluation is to show that there is adequate separation of redundant trains of equipment required to achieve safe shutdown in the event of a fire in the Screenhouse, Fire Areas 41, 41A, and 41B. There are no significant fire hazards in FA 41 or 41B in the vicinity of the ventilation openings above doors 257 and 258. The redundant cooling water pumps and cables are also well separated from the ventilation openings. There is fire detection and an automatic fire suppression system on both sides of the openings. These defense-in depth features provide the justification that the ventilation openings above doors 257 and 258 are acceptable as is.
EEEE Title CA-01311046-01, Fire Doors 257, 258, 259 & 260
Summary The purpose of this evaluation is to assess Doors 257, 258, 259 and 260, which do not have an identified fire rating. Doors 257 and 258 also have transoms above the doors that are not fire rated. The doors are in the boundaries between Fire Area 41 and Fire Area 41A on the 695ft elevation of the Screenhouse. Fire Area 41 is open to Fire Area 41B on the 670ft elevation. Fire Area 41A contains safe shutdown capabilities that are redundant to those in Fire Area 41B. Fire Doors 257 and 258, inclusive of the 14ga steel plate transoms, and Fire Doors 259 and 260 provide adequate protection to prevent fire spread that could adversely impact redundant safe shutdown capability. The bases for this conclusion include the following: Fire Doors 257, 258, 259, and 260 are of substantial construction, constructed by the same vendor that built many of the fire doors installed during original plant construction. The doors could be qualified as 1-1/2hr rated fire doors based on the referenced drawings. The transoms above Fire Doors 257 and 258 are 14ga plate steel, securely attached to the openings above the doors. While not fire rated, the transoms will resist the passage of flame, smoke and hot gases. Automatic detection and pre-action sprinklers are provided on both sides of the subject doors in Fire Areas 41A and 41. Automatic detection and pre-action sprinklers are provided throughout Fire Area 41A and Fire Area 41B, which would control the size of postulated fires in the areas. The automatic detection systems provided throughout Fire Area 41B and Fire Area 41A, and on the south side of Fire
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Area 41, would also result in prompt fire brigade response. Fire brigade response would be effective in controlling and extinguishing postulated fires prior to the fire impacting on redundant fire safe shutdown capability.
Variances from Deterministic Requirements (VFDR)
VFDR-041A-0-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. This condition would challenge the Nuclear Safety Performance Criteria for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. This is a separation issue for Vital Auxiliaries. Components and Cables: 11 Cooling Water Strainer Backwash Control Valve, CV-31652 (Many) 12 Cooling Water Strainer Backwash Control Valve, CV-31653 (Many) 21 Cooling Water Strainer Backwash Control Valve, CV-31654 (Many) 22 Cooling Water Strainer Backwash Control Valve, CV-31655 (Many) 11 Cooling Water Strainer, MTR-111C-21 (Many) 12 Cooling Water Strainer, MTR-121C-21 (Many) 21 Cooling Water Strainer, MTR-111C-22 (Many) 22 Cooling Water Strainer, MTR-121C-22 (Many) AC Distribution Panel 136, PNL-136 (Many) AC Distribution Panel 137, PNL-137 (Many) Motor Control Center 1AB Bus 1, MCC-1AB1 (111C-5) Motor Control Center 1AB Bus 2, MCC-1AB2 (221C-4) Compliant Case: One strainer in each Cooling Water Header (11 or 12 in Train A Header) and (21 or 22 in Train B Header) should remain unaffected by a fire in this area. The Train A and Train B Cooling Water Header need to remain functional because the isolation valves could be affected by the fire.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 75. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
41A Detection 75 Ionization, Heat N N Y N N
Suppression PA-9 Wet Pipe N N Y N N
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 41B Screenhouse Basement Below Grade
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 165- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B)
Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Train A or B Unit 1 - If CC Train A is not available use CC Train B Unit 2 - If CC Train A is not available use CC Train B If CL Train A is not available use CL Train B
VFDR FA41B-0-02 VFDR-FA41B-0-03 VFDR-FA41B-0-04
Reference Documents
Safe/Genesis V 4.0.2 EC 23592, Fire Risk Evaluation, Fire Area 41B, Screenhouse Basement Below Grade, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title CA 01241917-01, Fire Protection Engineering Evaluation of Fire Barrier between FA 41A and FA 41
Summary The purpose of this Fire Protection Engineering Evaluation is to show that there is adequate separation of redundant trains of equipment required to achieve safe shutdown in the event of a fire in the Screenhouse, Fire Areas 41, 41A, and 41B. There are no significant fire hazards in FA 41 or 41B in the vicinity of the ventilation openings above doors 257 and 258. The redundant cooling water pumps and cables are also well separated from the ventilation openings. There is fire detection and an automatic fire suppression system on both sides of the openings. These defense-in depth features provide the justification that the ventilation openings above doors 257 and 258 are acceptable as is.
Variances from Deterministic Requirements (VFDR)
VFDR-41B-0-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to lack of separation between redundant trains of cooling water cables. This condition would challenge the Nuclear Safety Performance Criteria for Vital Auxiliaries. Components and Cables: 11 Cooling Water Strainer Backwash Valve, CV-31652, (1CA-529, 1CB-370) 22 Cooling Water Strainer Backwash Valve, CV-31655 (1CA-529, 1CB-370) 12 Cooling Water Strainer Backwash Valve, CV-31653 (1CA-529, 1CB-370) 21 Cooling Water Strainer Backwash Valve, CV-31654 (1CA-529, 1CB-370)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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11 Cooling Water Strainer Motor, MTR-111C-21 (1CA-529, 1CB-370) 22 Cooling Water Strainer Motor, MTR-121C-22 (1CA-529, 1CB-370) 12 CL Strainer Motor, MTR-121C-21 (1CA-529, 1CB-370) 21 CL Strainer Motor, MTR-111C-22 (1CA-529, 1CB-370) AC Distribution Panel 136, PNL-136 (1CA-529, 1CB-370) AC Distribution Panel 137, PNL-137 (1CB-370) Motor Control Center 1AB Bus 2, MCC-1AB2 (221C-4) Motor Control Center 1AB Bus 1, MCC-1AB1 (111C-5) Compliant Case: Train A, 12 Cooling Water Strainer (CV-31653 and MTR-121C-21) or Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 74. Modification identified in Table S-2, Item #20 will correct fuse/breaker coordination for PNL-137. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification credited.
VFDR-41B-0-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power to Bus 23 and damage to power cables that could over-heat and cause secondary fires. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3.4.b due to a lack of OCT protection for the Power and Control cables for the Screenwash pump. Components and Cables: 121 Screenwash Pump MTR 23-1 (23401-2, 1C-1550, 1C-1552, 1C-2280, 1C-2285, 1C-4661, 2C-1359) 21 Cooling Water pump, BKR-23-4 (1C-2285, 2C-1359, 23404-2) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 identifies a modification to resolve VFDR-041B-0-03 by protecting the over-current trip function to protect cables from overheating and secondary fires. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-41B-0-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause failure of redundant cooling water pumps due to deviations in the fire detection and automatic fire suppression systems credited. This could cause a loss of cooling water to support safe shutdown equipment. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. Components and Cables: None Compliant Case: The fire detection and automatic fire suppression systems should meet NFPA requirements.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #9 identifies a modification to resolve VFDR-041B-0-04 by installing a missing Sprinkler (#229), and a sprinkler above the 122 Diesel Driven Fire Pump. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
41B Detection 74 Ionization, Heat
Y N Y Y N
Suppression PA-9 Pre-Action Y N Y N N
Feature ERFBS N N N Y N Cable 221C-4 has a Darmatt 1-hour cable wrap Hose Station N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 46 Cooling Tower Equipment House and Transformers
Note Fire Area 46 includes: Fire Area 46A Cooling Tower Transformers Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Vital Auxiliaries Unit 1 - Offsite Power (1RY) supplying Electrical Distribution Train A and B Unit 2 - Offsite Power (2RY) supplying Electrical Distribution Train A and B
None Variances from Deterministic Requirements (VFDR)
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
46 Detection 71 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 46A Cooling Tower Transformers
Note: Fire Area 46A is now combined into Fire Area 46.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1,2 58 Auxiliary Building Ground Floor Units 1 and 2
Fire Area 58 includes Fire Area(s): 73 Auxiliary Building Ground Floor Unit 2 Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
VFDR-058-1-07 VFDR-058-2-07 VFDR-058-2-011
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Inventory and Pressure Control Unit 1 - If Charging System (Train A) is not available use Safety Injection (Train B) Unit 2 - If Charging System (Train A) is not available use Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. If Charging Pump is not available use Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. If Charging Pump is not available use Safety Injection Pump (Train B) to inject borated water from the RWST
VFDR-058-1-02 VFDR-058-2-02 VFDR-058-1-012
Vital Auxiliaries Unit 1 - Offsite Power (1RY) supplying Electrical Distribution Train A or B. Use train A if B is not available Unit 2 - D5 Emergency Diesel Generator supplying Electrical Distribution Train A or Offsite Power (2RY) supplying electrical distribution Train B Use train A if B is not available Unit 1 – if CC Train B pump is not available use CC Train A pump to supply Train B flow path Unit 2 – if CC Train B pump is not available use CC Train A pump to supply Train B flow path CL Train A or B
Safe/Genesis V 4.0.2 EC 23593, Fire Risk Evaluation, Fire Area 58, Auxiliary Building Ground Floor Unit 1, Rev. 1, March 2014. Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title AR 1266236-01, Basis for the use of (2) 1 1/2 Hour Rated Fire Doors in a (3) Hour Fire Rated Barrier
Summary The purpose of this evaluation is to address the adequacy of Class B (1.5 hour) fire doors in Appendix R-required fire barriers. This
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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evaluation, through the use of a bounding condition, finds that the lesser-rated doors, where used in PINGP, are acceptable given the combustible loading and available suppression systems of the areas. No challenges to barrier integrity or safe shutdown will occur as a result of the current barrier configuration, as bound by this evaluation. The doors listed are considered acceptable without modification or further administrative control.
EEEE Title CA-01244458-02, Fire Doors 94 & 95
Summary The purpose of this evaluation is to assess the unrated penetrations through the transoms and the unrated ¼-in checker steel plates and tray/conduit penetrations above Doors 94 and 95 for impact on the fire area boundaries separating Fire Areas 73 and 58 from Fire Area 4. This evaluation also assesses the adequacy of the assemblies to provide adequate protection for the 3-hour barrier in which they are located. The bases for this conclusion include the following: Postulated fires north of Unit 1 Door 95 and Unit 2 Door 94 would not adversely impact on redundant safe shutdown capability in Fire Area 58 and Fire Area 73 based on approved exemption requests for lack of automatic suppression that rely on enclosing Division B safe shutdown cable in one hour rated fire barriers, enclosing Division A safe shutdown cable trays in the vicinity of MCCs and at specified coordinates in one hour rated fire barriers, low combustible loading, automatic detection, and fire brigade response. Postulated fire spread south through the unrated penetrations and ¼-in checker steel plate from Fire Area 58 or Fire Area 73 to Fire Area 4 will not impact on safe shutdown capability since the only cables of concern for safe shutdown in the event of a fire in Fire Area 4 are 1CT-1, 16408-1, and 15407-3. As such, a fire in one corridor in Fire Area 4 will not result in a simultaneous fire in the opposite corridor, and the unsealed penetrations and ¼-in checker steel plates above Unit 1 Door 95 and Unit 2 Door 94 will not be challenged by the same fire simultaneously.
Variances from Deterministic Requirements (VFDR)
VFDR-058-0-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of automatic cooling water strainer backwash function for 11, 12, 21, and 22 Cooling Water Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to AFWP. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of cooling water strainers. Components and Cables: 11 CL Strainer Backwash CV, CV-31652 (1CA-538) 11 CL Strainer Motor, MTR 111C-21 (1CA-538) 12 CL Strainer Backwash CV, CV-31653 (1CA-538) 12 CL Strainer Motor, MTR 121C-21 (1CA-538) 21 CL Strainer Backwash CV, CV-31654 (1CA-538) 21 CL Strainer Motor, MTR 111C-22 (1CA-538) 22 CL Strainer Backwash CV, CV-31655 (1CA-538) 22 CL Strainer Motor, MTR 121C-22 (1CA-538)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655, and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery action to manually backwash the cooling water strainer described in procedure F5 Appendix D, Zone 8. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery action credited.
VFDR-058-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of 11 and 12 Component Cooling (CC) water Pumps. Train B CC is credited to cool the RCP TBHX and the 12 Safety Injection Pump. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of suppression with a one hour barrier and detection in the area. Components and Cables; 11 CC Pump, MTR 15-5 (15405-1,15045-A,15405-B,15405-G,1CA-184) 12 CC Pump, MTR 16-5(16405-A , 1CB-181, 1CB-71, 16405-1) Compliant Case: The Train B, 12 Component Cooling Water Pump should remain unaffected by a fire. The 12 CC and its support systems should be available to supply Component Cooling water for the plant.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that for fires that affect the 12 CC pump, Train A 11 CC Pump should remain unaffected by a fire. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-058-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of 12 Charging Pump (Train A) and the 12 SI Pump (Train B) which are credited to inject borated water into the RCS to support Inventory Control and Reactivity Control. This could result in the loss of the ability to inject borated water into the RCS to maintain Reactivity Control and RCS Inventory. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Inventory Control. Components and Cables 12 Charging Pump, MTR 111J-1 ( 1CA-1264, 1CA-1266, 1CA-754, 1CA-91, 1CA-92, 1K1-21, 1K1-21A, 1K1-29, 1K1-30, 1K1-3B), 12 SI Pump, MTR 16-7 (16407-1, 16407-B) 11 RWST to Charging Pump Suction MV, MV-32060 (1K1-14, 1K1-14A, 1K1-14B) 11 VCT Outlet to Charging Pump Suction MV, MV-32061 (1K1-14A, 1K1-3A, 1K1-3B, 1K1-3) Motor Control Center 1K Bus 1, MCC 1K1 (111J-1) Compliant Case: The 12 Safety Injection Pump (MTR-16-7) should remain unaffected by a fire in this area.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of fire affecting both injection pumps is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-058-1-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 11 CC pump and spurious opening of MV-32093 and MV-32094. If only one (11 or 12) CC pump is running and both MV-32093 and MV-32094 open, there could be excessive flow through the CC pump potentially creating a run-out condition. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of Component Cooling Pumps. Components and Cables 11 CC Pump, MTR 15-5 (15405-1, 15405-A, 15405-B, 15405-G, 1CA-184, 1CA-24) 12 CC Pump, MTR 16-5 (16405-1, 16405-A, 1CB-181, 1CB-71) 11 RHR HX CC Inlet Valve, MV-32093 (15404-A, 1K1-9A, 1K1-9B) 12 RHR HX CC Inlet Valve, MV-32094 (1K2-5A, 1K2-5B) 11 CC HX Outlet Valve, MV-32120 (1K1-4, 1K1-4A, 1K1-4B) 12 CC HX Outlet Valve, MV-32121 (1KA2-7, 1KA2-7A, 1KA2-8B) Motor Control Center 1K Bus 1, MCC 1K1 (111J-1) Motor Control Center 1K Bus 2, MCC 1K2 (121J-1, 121J-2)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Motor Control Center 1KA Bus 2, MCC 1K1 (121B-1) Compliant Case: If only the 12 CC pump is available, MV-32121 should be able to close or MV-32093 should not spuriously open. If only the 11 CC pump is available, MV-32120 should be able to close or MV-32094 should not spuriously open.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of losing Component Cooling due to run-out is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-058-1-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of DC control power for and damage to power cables for the following 4KV motor breakers: BKR-13-3, BKR-15-1, BKR-15-4, BKR-15-5, BKR-15-9, BKR-16-1, BKR-16-5, BKR-16-6, and BKR-16-7. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of vital buses. Components and Cables: Bus 13 Feed Breaker to 103-105-109-301 Transformer, BKR 13-3 (13403-3, 13408-B, 1C-1293, 1C-1295, 1C-4637) 11 SI Pump, MTR 15-1 (15401-1, 15401-B, 15401-C, 15401-E) 11RHR Pump, MTR 15-4, (15404-1, 15404-C, 15404-E, 1CA-753, 1CA-98) 11 CC Pump, MTR 15-5 (15405-1, 15405-A,, 15405-G, 1CA-184) 11 CS Pump, MTR 15-9 (15409-1, 15409-B, 1CA-97) 12 RHR Pump, MTR 16-6 (16406-1, 1CB-36, 1CB-564) 12 CS Pump, MTR 16-1 (16401-1, 16401-B, 1CB-29) 12 SI, MTR 16-7 (16407-1, 16407-B) 12 CC Pump, MTR 16-5 (16405-1, 16405-A,, 1CB-71) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Disposition
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-058-1-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of the 11 Containment Spray Pump and spurious opening of the discharge valve MV-32103. This could cause a drain-down of the RWST which is the credited makeup source for injecting borated water into the RCS. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3.due to lack of suppression with a one hour barrier and detection in the area. Components and Cables 11 CS Pump, MTR 15-9 (15409-B, 15409-D, 15409-E, or 1CA-97) 11 CS discharge valve, MV-32103 (1K1-10A) Compliant Case: No flow diversion is caused by spurious operation of the Containment Spray System, so that RWST Inventory Control is maintained.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of spurious operation of the Containment Spray System is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the e acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-058-1-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of the 12 Containment Spray Pump and spurious opening of the discharge valve MV-32105. This could cause a drain-down of the RWST which is the credited makeup source for injecting borated water into the RCS. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of suppression with a one hour barrier and detection in the area.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Components and Cables 12 CS Pump, MTR 16-1 (16401-B, or 1CB-29) 12 CS discharge valve, MV-32105 (1KA2-3A) Compliant Case: MV-32105 should not spuriously open concurrently with a spurious start of the 12 CS pump and deplete the RWST.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of spurious operation of the Containment Spray System is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-058-1-07
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious isolation of 12 MDAFWP flow to 12 Steam Generator. This could prevent adequate AFW flow to the credited Steam Generator. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of separation with one hour barrier and detection in the area. Components and Cables: 12 MDAFW Pump Discharge to 11 Steam Generator Valve, MV-32381 (1CB-52) 12 MDAFW Pump Discharge to 12 Steam Generator Valve, MV-32382 (1CB-52) 12 MDAFW Pump, MTR 16-3 (1CB-31) Compliant Case: The 12 MDAFWP, and discharge valves should remain unaffected by a fire to provide AFW flow to 12 Steam Generator.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of spurious FW valve operation is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-058-1-08
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of MV-32202 or MV-32203 which could isolate the Safety Injection (SI) Pump minimum flow recirculation path back to the RWST. If the (11 or 12) Safety Injection Pump started with the RCS at normal operating pressure and the min flow recirc path isolated,
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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the SI Pump would be dead headed and could be damaged. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation with one hour barrier and detection in the area. Components and Cables: 11 RWST to Charging Pump Suction Valve, MV-32060 (1K1-14,1K1-14A, 1K1-14B) SI Recirculation Valve SI test to 11 RWST isolation MV Train B, MV-32203 (1KA2-11C) SI Recirculation Valve SI test to 11 RWST isolation MV Train A, MV-32202 (1K1-15B) 13 Charging Pump, MTR-121J-2 (1CB-567, 1K2-7, 1K2-7A, 1CB-25, 1CB-24, 1CB-27, 1CB-1050) Compliant Case: The SI pump min flow recirc valves (MV-32202 or MV-32203) should not spuriously close and dead-head the SI Pumps.
Disposition
Recovery Action(s): Evaluate risk of recovery actions to manually open VC-1-1, MV-32060 Bypass Charging Pump Suction, and manually close VC-3-8, 11 VCT Outlet Manual Valve Isolation, supply valves to charging pump and re-start MTR-121J-2 (13 Charging Pump). Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after suction from the RWST is restored to inject borated water into the RCS. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions and a plant modification.
VFDR-058-1-11
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of CT11 and 1R source to BUS-15 and BUS-16 and the D1 source (034-011) to BUS-15 and the D2 source (034-021) to BUS-16. The power sources to safeguards BUS-15 and BUS-16 could be impacted by a fire in this area. BUS-15 and/or BUS-16 are required to power safe shutdown equipment. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of suppression with a one hour barrier and detection in the area. Components and Cables: Bus 15 Source from D1 Diesel Generator, BKR-15-2 (15402-1, 15402-G, 15402-K, 1CA-1140) Bus 15 Source from 1R XFMR, BKR-15-3 (1C-332, 1C-333, 15403-B) Bus 15 Source from CT 11, BKR-15-7 (15407-3, 15407-A, 16408-1) Bus 16 Source from 1R XFMR, BKR-16-2 (1C-332, 1C-333) Bus 16 Source from CT 11, BKR-16-8 (15407-3, 16408-1)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Bus 16 Source from D2 Diesel Generator, BKR-16-9 (16409-1, 1CB-135) D2 Diesel Generator, 034-021 (1CB-116, 1CB-117, 1CB-121, 1CB-130, 1CB-133, 1CB-135, 1CB-140, 1CB-526, 1DCB-2) CT-11 4.16KV CLG TWR Switchgear, BUS-CT11 (1CT-1) Compliant Case: BUS-16 should remain powered from either the 1RY or D2 source. BUS-15 should remain powered from the 1RY source for fire scenarios that impact Train B safe shutdown equipment in this area.
Disposition
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #6 will ensure BUS-16 will remain powered from either the 1RY or D2 source. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-058-1-12
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of power to PNL-171 and PNL-181 which power CV-31098 and CV-31099 Main Steam Isolation Valves and prevent remote closure of the MSIVs. Loss of the ability to close the MSIV could cause excessive steam flow and excessive cool down of the RCS, which could increase Reactivity. The Nuclear Safety Performance Criteria is not met for Reactivity Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of suppression with a one hour barrier and detection in the area. Components and Cables: DC Distribution Panel 181, PNL-181 (1DCB-102) DC Distribution Panel 171, PNL-171 (1DCA-102) Compliant Case: PNL-171 (Train A power) or PNL-181 (Train B power) for closing CV-31098 and CV-31099 should remain un-affected by a fire to isolate steam flow to prevent excessive cool down of the RCS.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes risk is low for a fire to cause concurrent damage to Train A cables and Train B cable 1DCB-102 for PNL-181.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-058-2-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of 21 and 22 Component Cooling (CC) water Pumps. Train B CC is credited to cool the RCP TBHX and the 22 Safety Injection Pump. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of suppression with a one hour barrier and detection in the area. Components and Cables 21 CC Pump, MTR 25-13 (25413-1, 25413-C, 25413-D, 25413-E, 25413-G, 2CA-4) 22 CC, MTR 26-5 (26405-1, 26405-D, 26405-E, 2CB-7) Compliant Case: MTR-26-5, 22 Component Cooling Water Pump (Train B) should remain unaffected by a fire.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that for fires that affect the 22 CC pump, that Train A 21 CC Pump remains unaffected by a fire. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation.
VFDR-058-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of 22 Charging Pump (Train A) and the 22 SI Pump (Train B) which are credited to inject borated water into the RCS to support Inventory Control and Reactivity Control. This could result in the loss of the ability to inject borated water into the RCS to maintain Reactivity Control and RCS Inventory. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of Inventory Control. Components and Cables 22 Charging Pump, MTR 211J-1 (2CA-148, 2CA-162, 2CA-199, 2CA-45, 2CA-624, 2CA-626, 2K1-41, 2K1-42, 2K1-5, 2K1-5A, 2K1-7B) Motor Control Center 2K Bus 1, MCC 2K1 (211J-1) 22 SI Pump, MTR 26-10 (26410-1, 26410-C)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Bus 25 Feed Breaker to 11A Transformer, BKR 25-15 (112A-1, 112A-2, 112A-3, 25415-1, 25410-E, 2CA-750) Bustie Breaker Bus 25/Bus 15, BKR 25-17 (25417-1, 25417-2, 25410-E, 25417-C, 2CA-749) Bus 25 Source from D5 Diesel Generator, BKR 25-2 (25402-E, 2CA-751, 2CA-757) Bus 25 Source from CT-12, BKR 25-5 (25405-C, 2CA-751) Bus 25 Source from 2RY XFMR, BKR 25-16 (25416-1, 25416-2, 25416-C, 2CA-730, 2CA-751) 21 RWST to Charging Pump Suction MV, MV-32062 (2K1-6, 2K1-6A, 2K1-6B) 21 VCT Outlet to 21 Charging Pump Suction Header MV, MV-32063 (2K1-7, 2K1-7A, 2K1-7B, 2K1-6A) Compliant Case: The 22 Safety Injection Pump (MTR-26-10) should remain unaffected by a fire in this area.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a 1 hour fire barrier or 20 feet of separation is adequate to prevent fire damage to both Trains of injection pumps. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for BKR 25-15 and 25-17. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation and a plant modification credited.
VFDR-058-2-03
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 21 CC pump and spurious opening of MV-32128 and MV-32129. If only one (21 or 22) CC pump is running and both MV-32128 and MV-32129 open, there could be excessive flow through the CC pump potentially creating a run-out. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of Component Cooling Pumps. Components and Cables 21 CC Pump, MTR 25-13 (25413-1, 25413-C, 25413-D, 25413-E, 25413-G, 2CA-4) 22 CC Pump, MTR 26-5 (26405-1, 26405-D, 26405-E, 2CB-7) 21 RHR HX CC Inlet Valve, MV-32128 (2K1-3A, 2K1-3B, 25407-D) 22 RHR HX CC Inlet Valve, MV-32129 (2K2-1A, 2K2-1B) 21 CC HX Outlet Valve, MV-32122 (2K1-4, 2K1-4A, 2K1-4B) 22 CC HX Outlet Valve, MV-32123 (2KA2-2, 2KA2-2A, 2KA2-2B) Motor Control Center MCC 2K2 (221J-1) Motor Control Center MCC 2KA2 (221B-1) Motor Control Center MCC 2K1 (211J-1)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Compliant Case: If only the 22 CC pump is available, MV-32128 should not spuriously open.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of spuriously opening MV-32128 and MV-32129 is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation.
VFDR-058-2-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of DC control power for and damage to power cables for the following 4KV breakers: BKR-24-6, BKR-25-7, BKR-25-8, BKR-25-9, BKR-25-10, BKR-25-13, BKR-26-5, BKR-26-9, BKR-26-10, and BKR-26-11. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of Vital Auxiliaries. Components and Cables Bus 24 Feed Breaker to 204-206-209-402 Transformer, BKR-24-6 (24403-F, 24404-F, 24406-3, 2C-2552, 2C-866) 21 RHR Pump, MTR 25-7, (25407-1, 25407-C, 2CA-8, 25410-E) 21 SI Pump, MTR 25-8 (25408-1, 25408-B, 25408-C, 25410-E), 21 CS Pump, MTR 25-9 (25409-1, 25409-D, 2CA-7, 25410-E), 21 AFW Pump, MTR 25-10 (25410-1, 25410-E, 2CA-778), 21 CC Pump, MTR 25-13 (25413-1, 25413-D, 25413-E, 2CA-4, 25410-E) 22 CC Pmp, MTR 26-5 (26405-1, 26405-D, 2CB-7) 22 CS, MTR 26-9 (26409-1, 26409-E,, 2CB-315) 22 SI Pump, MTR 26-10 (26410-1, 26410-C) 22 RHR Pump, MTR 26-11 (26411-1, 2CB-9) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition
Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-058-2-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of the 21 Containment Spray Pump and spurious opening of the discharge valve MV-32114. This could cause a drain-down of the RWST which is the credited makeup source for injecting borated water into the RCS. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of suppression with a one hour barrier and detection in the area. Components and Cables: 21 CS Pump, MTR 25-9 (25409-C, 25409-D, 25409-E, 2CA-7) 21 CS Pump Discharge Valve, MV-32114 (2K1-13B) Compliant Case: MV-32114 should not spuriously open concurrently with a spurious start of the 21 CS pump and deplete the RWST.
Disposition
Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that MV-32114 should not spuriously open concurrently with a spurious start of the 21 CS pump and deplete the RWST. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-058-2-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of the 22 Containment Spray Pump and spurious opening of the discharge valve MV-32116. This could cause a drain-down of the RWST which is the credited makeup source for injecting borated water into the RCS. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of suppression with a one hour barrier and detection in the area.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Components and Cables 22 CS Pump, MTR 26-9 (26409-E, 2CB-315) 22 CS Pump Discharge Valve, MV-32116 (2KA2-8C) Compliant Case: MV-32116 should not spuriously open concurrently with a spurious start of the 22 CS pump and deplete the RWST.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the spurious opening of MV-32116 is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-058-2-07
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of MV-32019 and MV-32020 which could isolate the steam supply to the 22 Turbine Driven Auxiliary Feedwater Pump. This could cause a loss of AFW flow to the Steam Generator to provide Decay Heat Removal. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to the lack of separation between redundant trains of decay heat removal. Components and Cables 21 SG steam supply to 22 TDAFW Pump, MV-32019 (2K1-40A) 22 SG steam supply to 22 TDAFW Pump, MV-32020 (2K2-13A) 21 MDAFW Pump, MTR 25-10 (25410-1, 25410-C, 25410-D, 25410-E, 2CA-778) Instrument Bus II (White) Panel 211, PNL-211 (2CW-1) Distirbution Panel 2EMA, PNL-2EMA (2CA-601) Compliant Case: MV-32020 should not spuriously close in this area and isolate steam to the 22 Turbine Driven Aux Feedwater Pump.
Disposition Recovery Action(s): No recovery actions credited. Credit Fire Risk Evaluation and spatial separation between MV-32019 and MV-32020 cables so that if a fire causes spurious closure of MV-32020, then MV-32019 and 1L-487, 21 SG level indication would remain available to support DHR.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-058-2-08
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of MV-32204 or MV-32205 which could isolate the Safety Injection Pump minimum flow recirculation path back to the RWST. If the (21 or 22) Safety Injection Pump started with the RCS at normal operating pressure and the min flow recirc path isolated, the SI Pump would be dead headed and could be damaged. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. Components and Cables: 21 RWST to Charging Pump Suction Valve, MV-32062 (2K1-6A, 2K1-6B, 2K1-6) SI Recirculation Valve SI test to 21 RWST isolation MV Train A, MV-32204 (2K1-16B wrapped) SI Recirculation Valve SI test to 21 RWST isolation MV Train B, MV-32205 (2KA2-20B) 23 Charging Pump, MTR-221J-2 (2CB-178, 2K2-5, 2K2-5A, 2K2-16, 2CB-193, 2CB-574, 2CB-181, 2K2-17) Compliant Case: The SI pump min flow recirc. valves (MV-32204 or MV-32205) should not spuriously close and dead-head the SI Pumps.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually open 2VC-1-1, MV-32062 Bypass Charging Pump Suction, manually close 2VC-3-8, 21 VCT Outlet Manual Valve Isolation, supply valves to charging pump and re-start MTR-221J-2 (23 Charging Pump). Modification identified in Table S-2, Item #15 will provide suction protection to the charging pumps so the charging pump can be restarted after local action to open suction from the RWST is restored to inject borated water into the RCS. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a recovery action and a plant modification credited.
VFDR-058-2-11
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-25-10, 21 MDAFWP and loss of power MCC-2A1 which powers MV-32383 and MV-32384. A loss of power to MV-32383 and MV-32384 would prevent remote closure of the valves and the inability to isolate spurious AFW flow. This could eventually lead to an over-fill of the Steam Generator which would challenge the Decay Heat Removal Nuclear Safety Performance Criteria. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements (VFDR) of NFPA 805 Section 4.2.3, lack of separation between redundant trains of decay heat removal due to over-fill of steam generator. Components and Cables:
21 Motor Driven AFW Pump, MTR-25-10 (25410-C, 25410-D, 25410-E, 2CA-778)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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21 MDAFW Pump Discharge to 21 SG MV, MV-32383 (2CA-116) 21 MDAFW Pump Discharge to 22 SG MV, MV-32384 (2CA-116) Motor Control Center 2A Bus 1, MCC-2A1 (211E-1) Compliant Case: One train of AFW should remain unaffected by a fire.
Disposition Recovery Action(s): The Fire Risk Evaluation concludes that the risk of over-filling the 21 and 22 Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
58 Detection 8 Ionization, Heat
N N Y Y N
Detection 40 Ionization N N Y Y N
Detection 108 Ionization, Heat
N N Y Y N
Suppression SWP-2 Wet Pipe N N N N N Stairwell System Suppression SWP-4 Wet Pipe N N N N N Stairwell System
Suppression WPS-11 Wet Pipe N N N N N Elevator Machine Room Feature ERFBS N N N Y N 1AG-LA12, 1AG-LA30, 1AG-TA1, 1AG-TA2, 1AG-TA3, 1AG-TA4,
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 192- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
There is an automatic fire suppression system in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Fire Area 58 now includes the Unit 2 portion of the elevation, which was Fire Area 73 prior to the transition to NFPA 805.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 59 Auxiliary Building Mezzanine Level Units 1 and 2
Fire Area 59 includes Fire Area(s) 74 Auxiliary Building Mezzanine Level Unit 2 Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 21 MDAFW Pump to 22 SG
VFDR-059-2-010
VFDR-059-2-011
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 194- Revision 1
Process Monitoring If Train B Process Monitoring is not available, use Train A RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
VFDR-059-1-010
VFDR-059-2-08
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 195- Revision 1
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Train A Unit 1 - CC Train A Unit 2 - CC Train A CL Train A and B
VFDR-059-1-011 VFDR-059-2-012
Reference Documents
Safe/Genesis V 4.0.2 EC 23604, Fire Risk Evaluation, Fire Area 59, Auxiliary Building Mezzanine Level, Rev. 1, March 2014 Licensing Actions
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
ENG-ME-048, Appendix R RCS Inventory Control with a SI Pump
Summary During postulated fire scenarios it is expected that at least one charging pump would remain operational. However, in the Auxiliary Building a fire could be postulated to disable all three charging pumps for one unit. If this were to occur, PINGP assumes that the SI System would be used to control RCS inventory. This evaluation demonstrated that the fire protection features provide a level of protection for the existing hazards in the areas.
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
VFDR-059-1-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31245 (11 Reactor Coolant Pump (RCP) Thermal Barrier Heat Exchanger (TBHX)) and CV-31335 (11 RCP Seal Injection). This could cause a loss of all RCP seal cooling to 11 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of cooling to RCP seals. Components and Cables: 11 RCP TBHX, CV-31245 (1C-2221) 11 RCP Seal Injection CV-31335 (1C-1169) Compliant Case: CV-31245 should not spuriously close due to a fire in this area
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 197- Revision 1
This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-1-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31246 (12 Reactor Coolant Pump (RCP) Thermal Barrier Heat Exchanger (TBHX)) and CV-31336 (12 RCP Seal Injection). This could cause a loss of all RCP seal cooling to 12 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of cooling to RCP seals. Components and Cables: 12 RCP TBHX, CV-31246 (1C-4638) 12 RCP Seal Injection, CV-31336 (1C-1171) Compliant Case: CV-31246 should not spuriously close due to a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #18 will install new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-1-03
This VFDR involves a fire in FA 59 which spuriously opens CV-31231 (1 PRZR PORV B CV) and MV-32195 (1 PRZR PORV ISOLATION A MV). If the Power Operated Relief Valve (PORV) spuriously opens and the block valve cannot close, a loss of RCS Inventory could occur. CV-31231 is separated from MV-32195 by a one hour fire barrier or 20 feet with negligible intervening combustibles with automatic fire detection, but an automatic fire suppression system is not installed in the area This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious opening of CV-31231 (1 PRZR PORV B) and loss of ability to close MV-32195 (1PRZR PORV A Isolation Valve). This could a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of PORV isolation.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Components and Cables: 1 PRZR PORV B Control Valve, CV-31231 (1CB-928) 1 PRZR PORV A Isolation MV, MV-32195 (1LA1-11, 1LA1-11A, 1LA1-11B) Bus 112 Source from 112M XFMR, BKR-112M (1CA-1306) 112M Transformer, 112M/XFMR (15406-1) Compliant Case: CV-31231 should not spuriously open and cause a loss of RCS Inventory and Pressure Control.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that spurious operation of CV-31231 is low and therefore will not spuriously operate. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-059-1-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious opening of CV-31232 (1 PRZR PORV A) and loss of ability to close MV-32196 (1PRZR PORV B Isolation Valve). This could a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of PORV isolation. Components and Cables: 1 PRZR PORV A Control Valve, CV-31232 (1CA-1133) 1 PRZR PORV B Isolation Valve, MV-32196 (1LA2-12, 1LA2-12A, 1LA2-12B) Motor Control Center 1LA Bus 2, MCC-1LA2 (122L-1) Bus 122 Source from 122M XFMR, BKR-122M (1CB-1026) 122M Transformer, 122M/XFMR (16411-1) Compliant Case: The Fire Risk Evaluation concludes that spurious operation of CV-31232 is low and therefore will not spuriously operate
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 199- Revision 1
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a single fire will not simultaneously affect both the PORV and the block valve This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with credited fire risk evaluation.
VFDR-059-1-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of CV-31226, Unit 1 Letdown Isolation Train A, and spurious opening of CV-31255, Unit 1 Letdown Isolation Train B control valve. This could a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of letdown isolation. Components and Cables: Letdown Isolation Train A, CV-31226 (1CA-291) Letdown Isolation Train B, CV-31255 (1CB-283) Compliant Case: CV-31226 or CV-31255 should remain free of fire damage to isolate the normal letdown path.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a single fire will not simultaneously affect both Train A and Train B letdown isolation valves. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-1-06
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of CV-31330, 11 Excess Letdown Heat Exchanger Inlet Isolation Control Valve, and spurious opening of CV-31210 , 11 Excess Letdown Heat Exchanger Outlet Flow CV. This could a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 200- Revision 1
This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of letdown isolation. Components and Cables: Excess Letdown HX Inlet Isolation Valve, CV-31330 (1C-1128) Excess Letdown HX Outlet Flow Control Valve, CV-31210 (1CF-144) Compliant Case: CV-31330 should not spuriously open due to a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a single fire will not simultaneously affect both excess letdown isolation valves. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-1-08 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of MV-
32164, 1RCS Loop A Hot Leg RHR Supply (Inside) and MV-32165, 1RCS Loop A Hot Leg RHR Supply (Outside). MV-32164 has its breaker de-energized to preclude spurious opening. If both valves spuriously open, an Interfacing System Loss Of Coolant Accident (ISLOCA) could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant RHR suction valves. Components and Cables: 1RCS Loop A Hot Leg RHR Supply (Inside), MV-32164 (1LA1-2) 1RCS Loop A Hot Leg RHR Supply (Outside), MV-32165 (1LA1-3, 1LA-3A, 1LA1-3B) Compliance Case MV-32164 should not spuriously open due to three phase proper polarity hot shorts in this area.
Disposition Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 201- Revision 1
The Fire Risk Evaluation concludes that the risk of this VFDR is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-1-09 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of MV-
32230, 1RCS Loop B Hot Leg RHR Supply (Inside) and MV-32231, 1RCS Loop B Hot Leg RHR Supply (Outside). MV-32230 has its breaker de-energized to preclude spurious opening. If both valves spuriously open, an Interfacing System Loss Of Coolant Accident (ISLOCA) could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant RHR suction valves. Components and Cables: 1RCS Loop B Hot Leg RHR Supply (Inside), MV-32230 (1LA2-4) 1RCS Loop B Hot Leg RHR Supply (Outside), MV-32231 (1LA2-11, 1LA2-11A) Compliance Case MV-32230 should not spuriously open due to three phase proper polarity hot shorts in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of this VFDR is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-1-10
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Train A and Train B Process Monitoring indication. Train B cables are wrapped with a one hour fire barrier, except the following required cables for Breaker 16-10: 26401-1, 26401-2, and 2CB-679. If the fore mentioned Breaker 16-10 cables sustain fire damage and Train A Process Monitoring cables are concurrently damaged both Trains of Process monitoring will be lost. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of process monitoring.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 202- Revision 1
Components and Cables: Distribution Panel 1EMA, PNL 1EMA (1CA-1228) Pressurizer Level Cold Cal., LOOP 1L-433 (1CA-1106, 1CF-236) 11 SG Wide Range Level, LOOP 1L-487 (1CX-125) U1 Excore Detection Train A 1N51, LOOP 1N51 (1CNX-3, 1CNX-4) U1 Loop A RCS Wide Range Press, LOOP 1P-709 (1CX-119) U1 RCS Loop A Hot Leg Temp, LOOP 1T-450A (1CX-131) U1 RCS Loop A Cold Leg Temp, LOOP 1T-450B (1CX-133) Distribution Panel 1EMB, PNL 1EMB (1CB-981, 2CV-38) Pressurizer Level Red Channel, LOOP 1L-426RP (1CR-34) 12 SG Wide Range Level, LOOP 1L-488 (1CR-128) U1 Excore Detection Train B 1N52, LOOP 1N52 (1CNY-3) U1 Loop B RCS Wide Range Press, LOOP 1P-710 (1CR-122) U1 RCS Loop B Hot Leg Temp, LOOP 1T-451A (1CR-132) U1 RCS Loop B Cold Leg Temp, LOOP 1T-451B (1CR-134) Bustie Bus 16 and Bus 26, BKR 16-10 (26401-1, 26401-2, 2CB-679) Bustie Bus 26 and Bus 16, BKR 26-1 (26401-1, 26401-2, 2CB-679, 26401-C) 11 SG Wide Range LVL XMTR, 1LT-460, 1LR-460 (1CB-1001) Compliant Case: Train B Process Monitoring should remain available in the control room.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that there is a low risk that both trains of Process Monitoring will be impacted by a single fire in this area. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-1-11 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to 4 kV power cables and DC control power which could cause a spurious close of BKR 16-6 which could fault BUS-16. If the fire damaged DC control power, and then damaged 4kV power cables, the faulted power cable could cause the source breaker to BUS-16 to trip. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of Vital Auxiliaries.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 203- Revision 1
Components and Cables: 12 RHR Pump Breaker, BKR 16-6 (1CB-564, 16406-1) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition
Recovery Action(s): No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-2-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31248 (22 RCP TBHX) and CV-31427 (22 RCP Seal Injection). This could cause a loss of all RCP seal cooling to 22 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of RCP seal cooling. Components and Cables: 22 RCP TBHX, CV-31248 (2C-2556) 22 RCP Seal Injection, CV-31427 (2C-1455) Compliant Case: CV-31248 or CV-31427 should remain free of fire damage to provide cooling to the RCP seals.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-1, Item #1 has installed new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 204- Revision 1
This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-2-02
This VFDR involves a fire in FA 59, which spuriously opens CV-31233 (2 PRZR PORV B CV) and MV-32197 (2 PRZR PORV ISOLATION A MV). CV-31233 is separated from MV-32197 by a one hour fire barrier or 20 feet with negligible intervening combustibles with automatic fire detection, but an automatic fire suppression system is not installed in the area. If the PORV spuriously opens and the block valve cannot be closed, a loss of RCS inventory could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a VFDR of NFPA 805 Section 4.2.3.4 due to lack of separation between redundant trains of PORV isolation Components and Cables: 2 PRZR PORV B, CV-31233 (2CB-472) 2 PRZR PORV ISOLATION A, MV-32197 (2LA1-23, 2LA1-23A, 2LA1-23B) Compliant Case: CV-31233 should not spuriously open due to a fire in the area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of a single fire affecting both the PORV and the block valve is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a fire risk evaluation credited.
VFDR-059-2-03
This VFDR involves a fire in FA 59, which spuriously opens CV-31234 (2 PRZR PORV A CV) and MV-32198 (2 PRZR PORV ISOLATION B MV). CV-31234 is separated from MV-32198 by a one hour fire barrier or 20 feet with negligible intervening combustibles with automatic fire detection, but an automatic fire suppression system is not installed in the area. If the PORV spuriously opens and the block valve cannot be closed, a loss of RCS inventory could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a VFDR of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of PORV isolation.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 205- Revision 1
Components and Cables: 2 PRZR PORV A, CV-31234 (2CA-522) 2 PRZR PORV ISOLATION B, MV-32198 (2LA2-20, 2LA2-20A, 2LA-20B) Motor Control Center 2LA Bus 2, MCC-2LA2 (222L-1) Compliant Case: CV-31234 should not spuriously open due to a fire in the area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a single fire should not affect both the PORV and the block valve. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-04
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of CV-31230 (U2 Letdown Isolation Train A) and CV-31279 (U2 Letdown Isolation Train B). CV-31230 is separated from CV-31279 by a one hour fire barrier or 20 feet free of intervening combustibles. This could cause a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of letdown isolation. Components and Cables: Letdown Isolation Train A, CV-31230 (2CA-359) Letdown Isolation Train B, CV-31279 (2CB-350) Compliant Case: CV-31230 should remain unaffected by a fire to isolate Normal Letdown flow.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that a single fire should not affect both trains of letdown isolation. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 206- Revision 1
VFDR-059-2-05
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of CV-31422 (U2 Excess Letdown HX Inlet Isolation) and CV-31222 (U2 Excess Letdown HX Outlet Flow Control Valve). This could cause a loss of RCS Inventory. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a VFDR of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of excess letdown isolation. Components and Cables: Excess Letdown HX Inlet Isolation Valve, CV-31422 (2C-1455) Excess Letdown HX Outlet Flow Control Valve, CV-31222 (2CF-96) Compliant Case: CV-31222 should remain unaffected by a fire to isolate Excess Letdown flow.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that one train of excess letdown isolation will remain available. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-06 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of MV-32192, 2RCS Loop A Hot Leg RHR Supply (Inside) and MV-32193, 2RCS Loop A Hot Leg RHR Supply (Outside). MV-32192 has its breaker de-energized to preclude spurious opening. If both valves spuriously open, an Interfacing System Loss Of Coolant Accident (ISLOCA) could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 2.4.2.2 due to a lack of separation between redundant RHR suction valves. Components and Cables: 2RCS Loop A Hot Leg RHR Supply (Inside), MV-32192 (2LA1-10) 2RCS Loop A Hot Leg RHR Supply (Outside), MV-32193 (2LA1-14, 2LA1-14A, 2LA1-14B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 207- Revision 1
Compliant Case: MV-32192 should not spuriously open due to three phase proper polarity hot shorts in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of this VFDR is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-07
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of MV-32232, 2RCS Loop B Hot Leg RHR Supply (Inside) and MV-32233, 2RCS Loop B Hot Leg RHR Supply (Outside). MV-32232 has its breaker de-energized to preclude spurious opening. If both valves spuriously open, an Interfacing System Loss Of Coolant Accident (ISLOCA) could occur. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant RHR suction valves Components and Cables: 2RCS Loop B Hot Leg RHR Supply (Inside), MV-32232 (2LA2-10) 2RCS Loop B Hot Leg RHR Supply (Outside), MV-32233 (2LA2-8, 2LA2-8A, 2LA2-8B) Compliant Case: MV-32232 should not spuriously open due to three phase proper polarity hot shorts in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of this VFDR is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-08
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause loss of Train A and Train B Process Monitoring indication. Train B cables are wrapped with a one hour fire barrier, except cable 221F-1 which powers MCC
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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2AC1 which powers the 22 Battery Charger and PNL-212, PNL-214, and PNL-2EMB. Loss of power to 22 Battery Charger will eventually result in loss of power to Instrument Panels PNL-212, PNL-214, and PNL-2EMB when 22 Battery is depleted. The Instrument Panels are required power supplies for Train B Process Monitoring. The Nuclear Safety Performance Criteria is not met for Process Monitoring. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to lack of separation between redundant trains of process monitoring. Components and Cables: Train A Pressurizer Level Cold Cal., LOOP 2L-433 (1CA-1106, 2CF-207) 21 SG Wide Range Level, LOOP 2L-487 (2CX-70) U2 Excore Detection Train A 2N51, LOOP 2N51 (2CNX-3, 2CNX-4) U2 Loop A RCS Wide Range Press, LOOP 2P-709 (2CX-64) U2 RCS Loop A Hot Leg Temp, LOOP 2T-450A (2CX-76) U1 RCS Loop A Cold Leg Temp, LOOP 2T-450B (2CX-78) Train B Pressurizer Level Red Channel, LOOP 2L-426RP (2CR-9) 21 SG Wide Range Level Recorder, 2LR-460 (2CB-551) 22 SG Wide Range Level, LOOP 2L-488 (2CR-74) U2 Excore Detection Train B 2N52, LOOP 2N52 (2CNY-3) U2 Loop B RCS Wide Range Press, LOOP 2P-710 (2CR-68) U2 RCS Loop B Hot Leg Temp, LOOP 2T-451A (2CR-78) U2 RCS Loop B Cold Leg Temp, LOOP 2T-451B (2CR-80) Motor Control Center 2AC Bus 2, MCC 2AC2 (221F-1) Bus 26 4.16KV Switchgear, BUS-26 (Many) Compliant Case: Train A Process Monitoring indication from the Control Room should remain unaffected by a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that there is a low risk that both trains of Process Monitoring will not be impacted by a single fire in this area. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
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VFDR-059-2-10
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of the ability to trip MTR 21-3 (21 Main Feedwater Pump), and close MV-32028 (21 Main Feedwater Isolation Valve), and close CV-31135 (21 Main Feedwater Regulating Valve). This could cause an over-fill of the 21 Steam Generator. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of process monitoring. Components and Cables: 21 Main Feedwater Pump, MTR 21-3 (21403-G) 21 Main Feedwater Isolation Valve, MV-32028 (2K1-27, 2K1-27A, 2K1-28) 21 Main Feedwater Regulating Valve, CV-31135 (2CX-13, 2CR-12, 2CY-11, 2CW-13, 2CA-417, 2CB-115, 2CA-364, 2CA-366, 2CB-322, 2CB-325) 21 Main Feedwater Bypass Control Valve, CV-31371 (2CX-13, 2CR-12, 2CY-11, 2CW-13, 2CA-417, 2CB-115, 2CA-363, 2CA-365, 2CB-323, 2CB-324) 22 Main Feedwater Pump, MTR 22-3 (21403-K) Compliant Case: The ability to isolate Main Feedwater to prevent over-filling the 21 Steam Generators should not be affected by a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes risk is low for a fire to cause concurrent damage to redundant cables for the Main Feedwater regulating valves so that one train is unaffected by a fire in this area. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-11
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of the ability to trip MTR 22-3 (22 Main Feedwater Pump), and close MV-32029 (22 Main Feedwater Isolation Valve), close CV-31136 (22 Main Feedwater Regulating Valve). This could cause an over-fill of the 22 Steam Generator. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of process monitoring. Components and Cables: 21 Main Feedwater Pump, MTR 21-3 (21403-G)
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22 Main Feedwater Pump, MTR 22-3 (21403-K) 22 Main Feedwater Isolation Valve, MV-32029 (2KA2-23, 2KA2-23A) 22 Main Feedwater Regulating Valve, CV-31136 (2CX-13, 2CR-12, 2CY-11, 2CW-13, 2CA-417, 2CB-115, 2CA-361, 2CA-367, 2CB-321, 2CB-326) 22 Main Feedwater Bypass Control Valve, CV-31372 (2CX-13, 2CR-12, 2CY-11, 2CW-13, 2CA-417, 2CB-115, 2CA-362, 2CA-368, 2CB-320, 2CB-327) Motor Control Center 2KA Bus 2, MCC 2KA2 (221B-1) Compliant Case: The ability to isolate Main Feedwater to prevent over-filling the 22 Steam Generators should not be affected by a fire in this area.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes risk is low for a fire to cause concurrent damage to redundant cables for the Main Feedwater regulating valves so that one train is unaffected by a fire in this area. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
VFDR-059-2-12
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to DC control power to Bus 26 tripping circuits and subsequent damage to AC power cables resulting and could fail Bus 26. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between 4KV Breakers. Components and Cables: 22 CC, MTR 26-5 (26405-1, 26405-D) 22 CS, MTR 26-9 (26409-1, 26409-E) 22 SI, MTR 26-10 (26410-1, 26410-B, 26410-C) 22 RHR, MTR 26-11 (26411-1, 26411-C) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements. Coordination of BUS-26 should remain unaffected by a fire in this area.
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Disposition Recovery Action(s): No Recovery Actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables to preclude secondary fires on power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-2-13
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31247 (21 RCP TBHX) and CV-31426 (21 RCP Seal Injection) which could cause a loss of all seal cooling to the Reactor Coolant Pump. This could cause a loss of all RCP seal cooling to 21 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of RCP seal cooling. Components and Cables: 21 RCP TBHX, CV-31247 (2C-2553) 21 RCP Seal Injection, CV-31426 (2C-1455) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area.
Disposition Recovery Action(s): No recovery actions credited. Modification identified in Table S-1, Item #1 has installed new Reactor Coolant Pump Seals that will not be susceptible to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-059-2-16 This Variance From Deterministic Requirements is due to fire damage to cable(s) which results in a spurious "P" signal and the inability to trip BKR 26-9. Damage to cables for 2PT-945, 2PT-946, 2PT-947, 2PT-948, 2PT-949, and 2PT-950, Unit 2 Containment Pressure Transmitters. This could cause a flow diversion of the RWST to Containment. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3, due to lack of separation between redundant trains of containment pressure transmitters.
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Components and Cables: 2 CNTMT PRESS NUM 1 (CHAN I-RED) P XMTR, 2PT-945 (2CR-20) 2 CNTMT PRESS NUM 2 (CHAN II-WHI) P XMTR, 2PT-946 (2CW-19) 2 CNTMT PRESS NUM 3 (CHAN IV-YEL) P XMTR, 2PT-947 (2CY-45) 2 CNTMT PRESS NUM 5 (CHAN III-BLU) P XMTR, 2PT-948 (2CX-21) 2 CNTMT PRESS NUM 4 (CHAN II-WHI) P XMTR, 2PT-949 (2CW-20) 2 CNTMT PRESS NUM 6 (CHAN IV-YEL) P XMTR, 2PT-950 (2CY-18) 22 CONTAINMENT SPRAY PUMP BREAKER, BKR 26-9 (26409-C, 26409-E , 26409-F) Compliant Case: The Containment Spray Pump (MTR-26-9) should not spuriously start and deplete the RWST.
Disposition
Recovery Action(s): Evaluate risk of recovery actions in procedure F5 Appendix D, Zone 46, to manually trip BKR-26-9 at BUS-26 to prevent RWST drain down. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-059-2-17
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious opening of the pressurizer spray valves and loss of the ability to trip the RCP and loss of charging pump speed control. Spurious opening of the pressurizer spray valves would cause a reduction in RCS pressure. The Nuclear Safety Performance Criteria is not met for RCS Inventory Control and Pressure Control. Components and Cables: 21 RCP LOOP A PRZR SPRAY CV, CV-31228 (2CF-79) 22 RCP LOOP B PRZR SPRAY CV, CV-31229 (2CF-77) 21 REGEN HX AUX SPRAY TO 21 PRZR CV, CV-31421 (2C-1495) 22 CHARGING PUMP, MTR-211J-1 (2CA-45) 21 RC PMP, MTR-21-2 (21403-G, 21402-C) 22 RC PMP, MTR-22-2 (22402-D) Compliant Case: CV-31228, CV-31229, and CV-31421 should not spuriously open causing pressurizer spray to initiate.
Disposition
Recovery Action(s): No recovery actions credited.
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The Fire Risk Evaluation concludes that the risk of spurious pressurizer spray with a concurrent loss of charging is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
59 Detection 19 Ionization N N Y Y N
Detection 46 Ionization N N Y Y N
Detection 108 Ionization N N Y Y N Suppression PA-3 Pre-Action N N N N N
Suppression PA-4 Pre-Action N N N N N Suppression PA-6 Pre-Action N N N N N
Suppression PA-7 Pre-Action N N N N N
Suppression WPS-12
Wet Pipe N N Y N N
Suppression WPS-19
Wet Pipe N N Y N N
Feature ERFBS N N Y Y N 1AM-TA12, 1AM-TA13, 1AM-TA16, 1AM-TB26I, 1AM TR1, 1CA-1133, 1CB-928, 1CNY-3, 1LA1-1A, 1LA1-3B, 2CF-74
Hose Station N N N Y N
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
There are automatic fire suppression systems in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways or through grating to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
Fire Area 59 now includes the Unit 2 portion of the elevation, which was Fire Area 74 prior to the transition to NFPA 805.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 60 Auxiliary Building Operating Level Unit 1
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW or 21 MDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
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Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
Summary The purpose of this evaluation is to justify that sliding fire Doors 278 and 279 are acceptable as installed. The doors will provide adequate protection against the spread of fire in their current configuration Based on: minimal combustible loading on either side of both doors, the 3' concrete curb which will provide an additional level of protection to prevent flame propagation under the minor door to floor gaps, automatic smoke detection in the walkways, manual firefighting equipment located nearby, the lack of safe shutdown equipment in the area and therefore will not impact the ability to achieve and maintain safe shutdown.
EEEE Title FPEE-12-002; CA-01327430-1, Steam Line Pipe Penetrations without Penetration Seals PENF-1526, PENF-1528, PENF-1689, & PENF-
1692
Summary The purpose of this evaluation is to assess the impact on fire safe shutdown capability of four main steam line penetrations that are not provided with 3-hour fire rated penetration seals in G wall separating the Turbine Operating Deck from the 735ft elevation of the Auxiliary Building. The types, quantities, and continuity of combustible materials in Fire Area 60 or Fire Area 75 would not result in a sufficient heat release rate or fire to spread to, or spread through, the main steam line penetrations into Fire Area 8. The Turbine Stop Valves in Fire Area 8 are between 50 ft and 70 ft from the main steam line penetrations through G-wall. The only postulated fires that could result in damage to the Turbine Stop Valves and be large enough to potentially spread to the main steam line penetrations would involve either a turbine bearing oil fire or a catastrophic failure of the turbine oil system. The turbine bearings are protected by an automatically-actuated preaction sprinkler system that will respond to postulated turbine bearing oil fires and result in prompt fire brigade response. Postulated catastrophic turbine oil system failures could result in very severe fires; however, the area where oil piping runs and oil can spread are protected by automatic wet pipe sprinkler systems. Postulated turbine bearing oil fires and postulated catastrophic turbine oil system
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failure fires would result in significant heat release rates and smoke production; however, the large volume of the turbine building combined with the existing roof exhaust fans and smoke hatches that are fitted with automatic releases would release smoke and hot gas to the environment and delay the effects of such fires from banking down to the level of the main steam line penetrations located 50ft below the roof. The main steam line penetrations have limited annular gaps, 10in, for passage of fire effects to Fire Area 60 and Fire Area 75. There is very limited continuity of combustible materials in Fire Areas 60 and Fire Area 75 for fire to spread from Fire Area 8 to the vicinity of the MSlVs and solenoid valves. Postulated fires in Fire Area 8, Fire Area 60, and Fire Area 75 would not adversely impact redundant safe shutdown capability consisting of the Turbine Stop Valves in Fire Area 8 and the MSIVs and solenoid valves in Fire Area 60 and Fire Area 75.
EEEE Title Basis for the use of (2) 1 1/2 Hour Rated Fire Doors in a (3) Hour Fire Rated Barrier
Summary The purpose of this evaluation is to address the adequacy of Class B (1.5 hour) fire doors in Appendix R-required fire barriers. This evaluation, through the use of a bounding condition, finds that the lesser-rated doors, where used in PINGP, are acceptable given the combustible loading and available suppression systems of the areas. No challenges to barrier integrity or safe shutdown will occur as a result of the current barrier configuration, as bound by this evaluation. The doors listed are considered acceptable without modification or further administrative control.
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
60 Detection 28 Ionization, Heat N N Y N N
Detection 108 Ionization, Heat N N N N N
Suppression SWP-4 Wet Pipe N N N N N Stairwell System
Feature - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Fire Suppression Effects on Nuclear Safety Performance Criteria
There is an automatic fire suppression system in the fire area. In most areas, curbs, drains, and the mounting of equipment above the floor level minimizes the potential for flooding damage. Water will drain out doors or via stairways to lower elevations, such that the standing water would not affect safety-related equipment. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 61 Auxiliary Building Anti "C" Clothing Area
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
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Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
61 Detection 28 Ionization N N N N N
Suppression WPS-27 Wet Pipe N N N N N
Suppression WPS-28 Wet Pipe N N N N N
Feature - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 61A Auxiliary Building Hatch Area
Note: Fire Area 61A is now combined into Fire Area 4.
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Unit Fire Area Description 1, 2 62 Spent Fuel Pool Area
Note: Fire Area 62 is now combined into Fire Area 4.
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Unit Fire Area Description 1, 2 63 Filter Room
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B); Unit 2 - Charging System (Train A) or Safety Injection (Train B);
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
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Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
63 Detection - - - - - - -
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
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Unit Fire Area Description 1 64 Auxiliary Building Low Level Decay Area Unit 1
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 231- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power (1R) supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
64 Detection 8 Ionization N N N N N
Suppression - - - - N - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1 65 Spent Fuel Pool Heat Exchanger & Pumps
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 11 TDAFW or 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 235- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
65 Detection - - - - - - -
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 66 D3 Lunch Room
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Safety Injection (Train B) Unit 2 - Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train B Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Train B Unit 1 - CC Train B Unit 2 - CC Train B CL Train B
VFDR-066-0-01 VFDR-066-2-02 VFDR-066-2-03
Reference Documents
Safe/Genesis V 4.0.2 EC 23594, Fire Risk Evaluation, Fire Area 66, D3 Lunch Room, Rev. 1, March 2014 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
VFDR-066-0-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause faults on cables 25417-1 or 25417-2 or 15412-1 and spurious closure of BKR 15-8 or BKR-15-12 which could fault Bus 15 and cause a loss of power to PNL 136 which powers the CL strainer backwash control panel for 11, 12, 21, and 22 CL Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect that function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3, due to Lack of separation between redundant trains of cooling water strainers. Components and Cables: Bus Tie Breaker Bus 15/Bus 25, BKR-15-8 (25417-1, 25417-2 and 2CA-749) Bus Feed Breaker to 21A Transformer, BKR-15-12 (2CA-749, 15412-1) 22 CL Strainer Backwash CB, CV-31655 (none) 22 CL Strainer Motor, MTR-121C-22 (none) Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 239- Revision 1
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 39. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-066-2-02 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a loss of DC control power
to trip breakers on BUS-25 and damage to power cables could allow the cables to over-heat and start on fire. If the fire damaged DC control power, and then damaged 4kV power cables, the excessive current could cause load power cables to over-heat and develop secondary fires in other fire areas which violates common enclosure requirements. The Nuclear Safety Performance Criteria is not met for Vital AC Power. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant train vital buses. Components and Cables: 21 SI Pump Breaker, BKR 25-8 (25408-1, 25408-B, 25408-C, 25410-E), 21 RHR Pump Breaker, BKR 25-7 (25407-1, 25407-C, 25410-E) 21 CC Pump Breaker, BKR 25-13 (25413-1, 25413-D, 25413-E, 25410-E) 21 CS Spray Pump Breaker, BKR 25-9 (25409-1, 25409-D, 25410-E) 21 MDAFW Pump Breaker, BKR 25-10 (25410-1, 25410-E) Compliant Case: Cable over-current protection should be maintained to protect cables from over-heating and causing secondary fires to meet common enclosure requirements.
Disposition Recovery Action(s):
No recovery actions credited. Modification identified in Table S-2, Item #10 will ensure over-current protection is provided for power cables. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with a plant modification credited.
VFDR-066-2-03 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-25-
10, 21 MDAFWP and loss of power MCC-2A1 which powers MV-32383 and MV32384 and would prevent remote closure to
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 240- Revision 1
isolate spurious AFW flow. This could eventually lead to an over-fill of the Steam Generator which would challenge the Decay Heat Removal Nuclear Safety Performance Criteria. The Nuclear Safety Performance Criteria is not met for Decay Heat Removal. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to the lack of separation between redundant trains of decay heat removal due to over-fill of steam generator. Components and Cables: 21 MDAFW Pump, MTR-25-10 (25410-C, 25410-D, 25410-E) Motor Control Center 2A Bus 1, MCC 2A1 (211E-1) Compliant Case: The 21 MDAFWP should not spuriously start and over-fill the 21 and 22 Steam Generators.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes the risk of over-filling the Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
66 Detection 39 Ionization N N N N N
Suppression WPS-22 Wet Pipe N N N N N
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
The trench in Fire Area 66 communicates into Fire Area 70 beneath Door 91. The trench has been sealed with concrete and grout and includes a sealed pipe sleeve.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 1, 2 67 Resin Disposal Building
Note: Fire Area 67 is now combined into Fire Area 4.
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Unit Fire Area Description 1 69 Turbine Building Ground Floor & Mezzanine Floors Unit 1
Regulatory Basis Note: Fire Area 69 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 70 Turbine Building Ground Floor & Mezzanine Floors Unit 2
Note: Fire Area 70 is now combined into Fire Area 8.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 71 Containment and Containment Annulus Unit 2
Fire Area 71 includes Fire Area(s): 72 Containment Annulus Unit 2 Regulatory Basis Unit 1 4.2.3.2– Deterministic Approach Unit 2 4.2.3.4 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Note: Unit 2, one SG could be affected but the redundant SG remains available.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Note: Unit 2, one train of process monitoring could be affected but the redundant train remains available.
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
VFDR-071-2-01
VFDR-071-2-02
Reactivity Control Trip reactor from the Control Room. Use Charging Pump (Train A) (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 247- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
Unit 1 CC Train A or B
Unit 2 CC Train A or B
CL Train A or B
Reference Documents
Safe/Genesis V 4.0.2 Licensing Actions
Appendix R Exemption, Containment, RCP oil collection system not in strict compliance (III.O criteria), Units 1 and 2, Fire Areas 1 and 71 Reference Attachment K – Existing Licensing Action Transition for details
EEEE Title FPEE 2011-003, Fire Protection Engineering Evaluation of Appendix R Compliance with Section III.G.2.D Since Containment Annulus
Pre-Action Sprinkler System PA-3, PA-4, PA-6, and PA-7 May Not Actuate
Summary The purpose of this evaluation is to justify the treatment of the containment annulus in the same way the containment is treated in 10 CFR 50 Appendix R. Revision 1 to this FPEE evaluates the functional requirements for the existing pre-action fire suppression systems inside the containment annulus. FA 68 and 72 are in compliance with Appendix R, Section III.G.2.d, because redundant trains of equipment required for safe shutdown are separated by greater than 20 feet free of intervening combustibles. Although the Annulus is not inside the containment pressure boundary, it is inside the Reactor Containment Building and it qualifies to be treated like an area inside containment because access to the area is restricted in the same way access is restricted to containment during power operation. Reference 4.15 supports this position that the annulus is inside the Reactor Containment Building. Since FA 68 and FA 72 can be treated like areas inside containment, the partial area fire detection and automatic fire suppression in the Annulus do not need to be credited to meet the requirements of Appendix R, Section III.G.2. In addition, the functional requirements and surveillances required to ensure operability are not required for the cable tray sprinkler systems in the annulus of either unit, FA 68 and FA 72.
Variances from Deterministic Requirements (VFDR)
VFDR-071-2-01 This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31247 (21 RCP TBHX) and CV-31426 (21 RCP Seal Injection) which could cause a loss of all seal cooling to the Reactor Coolant Pump. This could cause a loss of all RCP seal cooling to 21 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3 due to a lack of separation between redundant trains of RCP seal cooling. Components and Cables: 21 RCP TBHX CC, CV-31247 (2C-2555, 2C-2560, 2C-472, 2C-473, 2C-474, 2C-475) 21 RCP seal water outlet isolation CV, CV-31426 (2C-1433, 2C-1434, 2C-1435) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area.
Disposition Recovery Action(s): No recovery actions credited. Table S-1, Item #1 has installed new Reactor Coolant Pump (RCP) seals that are not subject to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it has been determined to meet the acceptance criteria of NFPA 805 Section 4.2.3 with a plant modification credited.
VFDR-071-2-02
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause spurious closure of CV-31248 (22 RCP TBHX) and CV-31427 (22 RCP Seal Injection) which cause a loss of all seal cooling to the Reactor Coolant Pump. This could cause a loss of all RCP seal cooling to 22 RCP, which could result in increased leakage through the RCP seals. The Nuclear Safety Performance Criteria is not met for Inventory and Pressure Control. This condition represents a variance from the deterministic requirements of NFPA 805, section 4.2.3 due to a lack of separation between redundant trains of RCP seal cooling. Components and Cables: 22 RCP TBHX CC, CV-31248 (2C-2558, 2C-2559, 2C-477, 2C-478, 2C-479, 2C-497) 22 RCP seal water outlet isolation CV, CV-31427 (2C-1438, 2C-1439, 2C-1440) Compliant Case: RCP seal cooling from either seal injection or TBHX should remain unaffected by a fire in this fire area. .
Disposition Recovery Action(s): No recovery actions credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 249- Revision 1
Table S-1, Item #1 has installed new Reactor Coolant Pump (RCP) seals that are not subject to excessive leakage upon loss of all seal cooling. This VFDR has been evaluated and it has been determined to meet the acceptance criteria of NFPA 805 Section 4.2.3 with a plant modification credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area
Category ID Type Required?
Notes S L E R D
71 Detection 42 Ionization N N N Y N
Detection 52 Ionization N N N Y N Detection 54 Ionization N N N Y N
Detection 56 Ionization N N N Y N Feature See
Note ERFBS N N N Y N Cable 2CF-74 has 3M Interam wrap
Feature N/A Lube Oil
Collection System
N Y N N N
72 Detection 47 Ionization,
Flame N N N N N
Suppression PA-5 Pre-Action N N N N N
Suppression PA-6 Pre-Action N N N N N
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 250- Revision 1
The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Fire Area 71 now includes the Unit 2 Annulus, which was Fire Area 72 prior to the transition to NFPA 805.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 72 Containment Annulus Unit 2 Note: Fire Area 72 is now combined into Fire Area 71.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 73 Auxiliary Building Ground Floor Unit 2 Note: Fire Area 73 is now combined into Fire Area 58.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
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Unit Fire Area Description 2 74 Auxiliary Building Mezzanine Floor Unit 2 Note: Fire Area 74 is now combined into Fire Area 59.
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Unit Fire Area Description 2 75 Auxiliary Building Operating Level Unit 2
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG
Unit 2 -21 MDAFW Pump to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 1L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 255- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
EEEE Title AR 1266236-01, Basis for the use of (2) 1 1/2 Hour Rated Fire Doors in a (3) Hour Fire Rated Barrier
Summary The purpose of this evaluation is to address the adequacy of Class B (1.5 hour) fire doors in Appendix R-required fire barriers. This evaluation, through the use of a bounding condition, finds that the lesser-rated doors, where used in PINGP, are acceptable given the combustible loading and available suppression systems of the areas. No challenges to barrier integrity or safe shutdown will occur as a result of the current barrier configuration, as bound by this evaluation. The doors listed are considered acceptable without modification or further administrative control.
EEEE Title FPEE-11-001, Gaps on Doors 278 & 279
Summary The purpose of this evaluation is to justify that sliding fire Doors 278 and 279 are acceptable as installed. The doors will provide adequate protection against the spread of fire in their current configuration Based on: minimal combustible loading on either side of both doors, the 3' concrete curb which will provide an additional level of protection to prevent flame propagation under the minor door to floor gaps, automatic smoke detection in the walkways, manual firefighting equipment located nearby, the lack of safe shutdown equipment in the area and therefore will not impact the ability to achieve and maintain safe shutdown.
EEEE Title FPEE-12-002; CA-01327430-1, Steam Line Pipe Penetrations without Penetration Seals PENF-1526, PENF-1528, PENF-1689, & PENF-1692
Summary The purpose of this evaluation is to assess the impact on fire safe shutdown capability of four main steam and four feedwater line penetrations, also referred to in this evaluation as pipe penetrations, that are not provided with 3-hour fire rated penetration seals in G-wall separating the Turbine Operating Deck from the 735ft elevation of the Auxiliary Building. PENF-1526, PENF-1528, PENF 1530, and PENF-1533 contain pipe penetrations through G-wall between the Unit 1 side of the Turbine Operating Deck, Fire Area 8, and Fire Area
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 256- Revision 1
60 on the 735ft elevation of the Auxiliary Building. PENF-1686, PENF-1687, PENF-1689, and PENF-1692 contain pipe penetrations through G-wall between the Unit 2 side of the Turbine Operating Deck, Fire Area 8, and Fire Area 75 on the 735R elevation of the Auxiliary Building. Boot seals are provided on the Auxiliary Building side of all eight pipe penetrations, but there is no other seal material in the rest of each penetration through G-wall. The unprotected penetrations represent a path of potential fire spread between the Turbine Operating Deck, Fire Area 8, and Fire Area 60 and Fire Area 75 on the 735ft elevation of the Auxiliary Building. Postulated fires in Fire Area 8, Fire Area 60, and Fire Area 75 would not adversely impact redundant safe shutdown capability consisting of the Turbine Stop Valves in Fire Area 8 and the MSlVs and solenoid valves in Fire Area 60 and Fire Area 75. The bases for this conclusion include the following: The types, quantities, and continuity of combustible materials in Fire Area 60 or Fire Area 75 would not result in a sufficient heat release rate for fire to spread to, or spread through, the pipe penetrations into Fire Area 8. The Turbine Stop Valves in Fire Area 8 are between 50ff and 70ft from the pipe penetrations through G-wall. The only postulated fires that could result in damage to the Turbine Stop Valves and be large enough to potentially spread to the pipe penetrations would involve either a turbine bearing oil fire or a catastrophic failure of the turbine oil system. The turbine bearings are protected by an automatically-actuated preaction sprinkler system, with that portion protecting the exciter manually-actuated, that will respond to postulated turbine bearing oil fires and result in prompt fire brigade response. Postulated catastrophic turbine oil system failures could result in very severe fires; however, the areas where oil piping runs and oil can spread are protected by automatic wet pipe sprinkler systems. Postulated turbine bearing oil fires and postulated catastrophic turbine oil system failure fires would result in significant heat release rates and smoke production; however, the large volume of the turbine building combined with the existing smoke hatches that are fitted with automatic releases would release smoke and hot gas tithe environment and delay the effects of such fires from banking down to the level of the pipe penetrations located 50 feet below the roof. The main steam and feedwater line penetrations have limited annular gaps, 10in and 7in respectively, for passage of fire to Fire Area 60 and Fire Area 75. There is very limited continuity of combustible materials in Fire Areas 60 and Fire Area 75 for fire to spread from Fire Area 8 to the vicinity of the MSlVs and solenoid valves.
Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
75 Detection 51 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 257- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 258- Revision 1
Unit Fire Area Description 2 76 Vent & Fan Room Unit 2
Note: Fire Area 76 is now combined into Fire Area 2.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 259- Revision 1
Unit Fire Area Description 2 77 Auxiliary Building Low Level Decay Area Unit 2
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 260- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 261- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
77 Detection - - - - - - -
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 262- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Fire Area 77 does not have a corresponding detection zone in procedure F5 Appendix A.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 263- Revision 1
Unit Fire Area Description 2 78 Waste Gas Compressor Area
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 264- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 265- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 266- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 267- Revision 1
Unit Fire Area Description 1 79 480 V Safeguard Switchgear Room (Bus 112)
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 268- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 269- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
79 Detection 26 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 270- Revision 1
the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 271- Revision 1
Unit Fire Area Description 1 80 480 V Safeguard Switchgear Room (Bus 111)
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Safety Injection (Train B) Unit 2 - Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 272- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train B Unit 2 - Offsite Power (CT 12) supplying Electrical Distribution Train B
Unit 1 CC Train B
Unit 2 CC Train B
CL Train B
VFDR-080-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23595, Fire Risk Evaluation, Fire Area 80, 480V Safeguards Switchgear Room (Bus 111), Rev. 1, March 2014 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-080-0-01 This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause loss of power to BUS-111 which powers MCC-1AB1 which powers PNL 136 which powers the Cooling Water Strainers backwash central control panel for 11, 12, 21, and 22 CL Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of cooling water strainers.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 273- Revision 1
Components and Cables: 22 CL Strainer Backwash CB, CV-31655 (none) 22 CL Strainer Motor, MTR-121C-22 (none) Bus 111 480V Switchgear, BUS-111 (1DCA-100) Compliance Case Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 43. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
80 Detection 43 Ionization N N Y N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 274- Revision 1
The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Mezzanine of FA 80 extends over hallway; is adjacent to FA 23.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 275- Revision 1
Unit Fire Area Description 1 81 4.16 kV Safeguard Switchgear Room (Bus 15)
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot: Unit 1 - 12 MDAFW Pump to 12 SG Unit 2 - 22 TDAFW Pump to 22 SG
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 -Safety Injection (Train B) Unit 2 -Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 276- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train B Unit 1 - Offsite Power (CT 12) supplying Electrical Distribution Train B
Unit 1 CC Train B
Unit 2 CC Train B
CL Train B
VFDR-081-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23596, Fire Risk Evaluation, Fire Area 81, 4.16KV Safeguards Switchgear Room (Bus 15), Rev. 1, March 2014 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
EEEE Title FPEE 01151461, Evaluation of Non-Compliance with Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent
Zone, Fire and Security Door Inspection
Summary The purpose of this evaluation is to analyze the fire doors that exceed the gap criteria of Preventive Maintenance Procedure PM 3122-3, Shield Building Category 1 Vent Zone, Fire and Security Door Inspection. Twelve of the fourteen doors are acceptable in their current configuration. Door 43 should be replaced due to the skin separation from the door frame. Door 429 should receive field maintenance to adjust the hinges to remove the gap. If this is not successful, the door should be replaced. Doors 34, 54, and 63 should be re-inspected. The field verification could not determine gaps exceeding the acceptance criteria for these doors. If the re-inspection does determine the gaps are non-compliant, the evaluation of each door as contained in this evaluation is applicable and the doors are acceptable as-is.
Variances from Deterministic Requirements (VFDR)
VFDR-081-0-01
This Variance from Deterministic Requirements (VFDR) is due to fire damage to cable(s) that could cause loss of power to BUS-15 which powers BUS-111 which powers MCC-1AB1 which powers PNL 136 which powers the Cooling Water Strainers backwash central control panel for 11, 12, 21, and 22 CL Strainers. Loss of the cooling water strainer backwash function could eventually lead to reduced cooling water flow and additional pressure loss across the strainers which would affect the function of cooling water to provide cooling to credited loads and backup supply to Aux Feedwater Pumps.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 277- Revision 1
The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This condition would challenge the Nuclear Safety Performance Criteria for Vital Auxiliaries.
This represents a variance from the deterministic requirements of NFPA 805, Section 4.2.3 due to lack of separation between redundant trains of cooling water strainers. Components and Cables: Bus 15 4.16 KV Switchgear, BUS 15 (Many) 22 CL Strainer Backwash CB, CV-31655 (none) 22 CL Strainer Motor, MTR-121C-22 (none) Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 43. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
81 Detection 11 Ionization N N Y Y N
Suppression - - - - - - -
Hose Station - - N N N Y N
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 278- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Mezzanine of FA 81 extends over hallway; is adjacent to FA 21.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 279- Revision 1
Unit Fire Area Description 1 82 480 V Safeguard Switchgear Room (Bus 122)
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 280- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 281- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
82 Detection 50 Ionization N N N N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 282- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 283- Revision 1
Unit Fire Area Description 2 83 Operator's Lounge
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 284- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 285- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
Fire Area 83 is split into two parts (Secondary Alarm Station and Operator’s Lounge).
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 286- Revision 1
Unit Fire Area Description 1, 2 84 Counting Room and Labs
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 287- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 288- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
84 Detection - - - - - - -
Suppression WPS-19 Wet Pipe N N N N N
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 289- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 290- Revision 1
Unit Fire Area Description 1, 2 85 Hold-up Tank Area/Demineralizer Area
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 291- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B); Unit 2 - Charging System (Train A) or Safety Injection (Train B);
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 292- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
EEEE Title CA-01040686 (Attachment), NFPA 72E Requirement for Non-Restorable Heat Detectors Evaluation
Summary The 2006 NFPA 72E, 1974 edition, code compliance review identified that nonrestorable spot type thermal detectors were not being tested in accordance with the code. Based on the type of fire expected in these fire areas, the generation of smoke will exceed the generation of heat during the incipient stage of a fire. Therefore, the area smoke detection will provide the early warning of a fire. The heat detectors are not credited in the USAR by reference to F5 Appendix K, but will provide a secondary level of automatic detection. Based on the lack of credit given to the heat detectors, the requirements of NFPA 72E are not mandatory and the functional testing for the rate of rise function is adequate. The additional testing of the fixed temperature function through destructive testing at a nationally recognized laboratory is not cost justified for the minor benefit. The lack of meeting the full code requirement for non-restorable heat detection testing will not have adverse effect on safe shutdown for any of the affected fire areas.
EEEE Title FPEE-12-006, CA-01311038-03, Fire Area 85 Boundaries and F5 Appendix K Barriers
Summary The purpose of this evaluation is to assess two distinct but related issues. One is the impact on fire safe shutdown capability of the fire area boundaries surrounding Fire Area 85. The second is the relocation of the F5 Appendix K barrier on the 715ft elevation. Based on the identified types and locations of combustible materials, the similarity fire safe shutdown impact, and the fire protection features provided, there is reasonable assurance that fire will not spread between Fire Area 85 and Fire Areas 60 and 75 on the 735ft elevation and between Fire Area 85 and Fire Areas 59 and 74 on the 715ft elevation and adversely impact fire safe shutdown capability. The only redundant safe shutdown capability is associated with CV-31742 in Fire Area 85 and CV-31743 in Fire Area 74 which are Train B and Train A, respectively, of the Unit 2 Reactor Building Instrument Air isolation control valves. Based on the large spatial separation between cabling for the components, there is no impact on fire safe shutdown capability should a fire spread between Fire Area 74 and Fire Area 85. As such, the F5 Appendix K fire barrier between the areas can be deleted.
Variances from Deterministic Requirements (VFDR)
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 293- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
85 Detection 8 Ionization, Thermal
N N Y N N
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 294- Revision 1
Unit Fire Area Description 1, 2 86 Intake Screenhouse Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 295- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 296- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Variances from Deterministic Requirements (VFDR)
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
86 Detection - - - - - - -
Suppression - - - - - - -
Feature - - - - - - - Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 297- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 298- Revision 1
Unit Fire Area Description 2 92 Water Chiller Room Unit 2
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 299- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 300- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
EEEE Title AR 1266236-01, Class B (1.5 hour) fire doors in Appendix R-required fire barriers
Summary This evaluation demonstrates that the Class B doors are compliant to the requirements of the SER. The maximum fire exposure duration to a 1.5 hour fire door protecting an Appendix R area, as bound by this evaluation, is 31 minutes. Applying a “one-half barrier rating” acceptance criteria, the 1.5 hour doors in Appendix R barriers are acceptable due to the maximum exposure being less than one-half of the 1.5 hour rating.
EEEE Title FPEE 01086132-0, Condition/Fire Protection Evaluation Adequacy of the fire boundary between Fire Area 76 (Vent and Fan Room, Unit
2, 755' Auxiliary Building) and Fire Area 92 (Water Chiller Room, Unit 2, 755' Auxiliary Building) without an installed three-hour fire damper
Summary The Purpose of this evaluation is to address the adequacy of the fire boundary between Fire Area 76 (Vent and Fan Room, Unit 2, 755' Auxiliary Building) and Fire Area 92 (Water Chiller Room, Unit 2, 755' Auxiliary Building) without an installed three-hour fire damper in a duct transversing the barrier. Although the duct work is not in a fire-tested configuration, the construction of the duct itself provides a one-hour measure of fire protection. Additionally, a lack of combustibles and ignition sources in the vicinity of the duct, as well as control of transient combustibles, minimizes the risk of fire. Should a fire occur, area-wide detection would quickly alert operators to the presence of products of combustion. The existing duct configuration (without damper) provides adequate protection from a fire in Fire Area 76 propagating into Fire Area 92.
Variances from Deterministic Requirements (VFDR)
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 301- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
92 Detection 31 Ionization N N Y N N
Suppression - - - - - - -
Feature - - - - - - -
Legend:
Required?
S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is not installed in the fire area. Water from other sources will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 302- Revision 1
Unit Fire Area Description 1, 2 93 Drum Storage/Low Level Rad Waste Note: Fire Area 93 is now combined into Fire Area 4. Unit Fire Area Description 1, 2 94 Service Building/Computer Room
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 303- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 304- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
94 Detection 94 Ionization N N N N N
Suppression DPS-2 Dry Pipe N N N N N Truck Aisle system
Suppression SWP-31 Wet Pipe N N N N N Stairwell system
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 305- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 306- Revision 1
Unit Fire Area Description 2 97 D5 Diesel Generator Building
Note: Fire Area 97 includes Fire Areas: 99, 101, 103, 105, 107, 109, 111, 113, 115, 117, 119, 123, 125, and 127
Regulatory Basis Unit 1 4.2.4.2 - Performance Based Approach Unit 2 4.2.4.2 - Performance Based Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 12 MDAFW Pump to 12 SG
Unit 2 - 22 TDAFW Pump to 22 SG
VFDR-097-2-01
Process Monitoring RCS Pressure (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Safety Injection (Train B) Unit 2 - Safety Injection (Train B)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 307- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Safety Injection Pump (Train B) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power (CT 11) supplying Electrical Distribution Train B Unit 2 - D6 supplying Electrical Distribution Train B Unit 1 - CC Train B Unit 2 - CC Train B CL Train B
VFDR-097-0-01
Reference Documents
Safe/Genesis V 4.0.2 EC 23597, Fire Risk Evaluation, Fire Area 97, D5 Building, Rev. 1, March 2014 Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room, Rev. 31 Licensing Actions
Summary This evaluation will determine the acceptability of the 4-foot section on the underside of the bottom flange of a steel beam in the D5 Cable Spreading Room that is not coated with fireproof material. The condition will be evaluated against the licensing basis. Based on the existing detection, extinguishers, combustible loading, ignition source control, and fire brigade response the evaluation demonstrates that the lack of fire proofing material is acceptable based on defense in depth SSCs which are adequate for the identified hazard, and is therefore acceptable.
Variances from Deterministic Requirements (VFDR)
VFDR-97-0-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause damage to Bus 25 and cables that could cause a spurious closure of BKR 25-17 and BKR 15-8 which could fault Bus 15 and cause a loss of power to PNL 136 which powers the CL strainer backwash control panel for 11, 12, 21, and 22 CL Strainers. The Nuclear Safety Performance Criteria is not met for Vital Auxiliaries. This represents a variance from the deterministic requirements of NFPA 805 Section 4.2.3 due to a lack of separation between redundant trains of cooling water strainers.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 308- Revision 1
Components and Cables: BKR-15-8 (2CA-749, 25417-1, 25417-2) 22 CL Strainer Backwash CB, CV-31655 (none) 22 CL Strainer Motor, MTR-121C-22 (none) Motor Control Center 2A Bus 1, MCC-2A1 (211E-1) Compliant Case: Train B, 22 Cooling Water Strainer (CV-31655 and MTR-121C-22) should remain unaffected by a fire in this area.
Disposition Recovery Action(s): Evaluate risk of recovery actions to manually backwash the cooling water strainers described in procedure F5 Appendix D, Zone 97. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with recovery actions credited.
VFDR-97-2-01
This Variance From Deterministic Requirements is due to fire damage to cable(s) that could cause a spurious start of MTR-25-10, 21 Motor Driven Aux Feedwater Pump, and a loss of power to MCC 2A1 which is required power to close MV-32383 and MV-32384. The NFPA 805 Nuclear Safety Performance Goal is not met for Decay Heat Removal. This represents a variance from the deterministic requirements (VFDR) of NFPA 805 Section 4.2.3.4.b due to a lack of separation between redundant trains of decay heat removal due to over-fill of steam generator. Components and Cables: 21 MDAFW Pump, MTR 25-10 (25410-A, 25410-B, 25410-C, 25410-D, 25410-E) Compliant Case: The 21 MDAFWP should not spuriously start and over-fill the 21 and 22 Steam Generators.
Disposition Recovery Action(s): No recovery actions credited. The Fire Risk Evaluation concludes that the risk of over-filling the 21 and 22 Steam Generators is low. This VFDR has been evaluated and it was determined that the risk, safety margin and defense-in-depth meet the acceptance criteria of NFPA 805 Section 4.2.4 with fire risk evaluation credited.
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 309- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
97 Detection 97 Ionization, Flame, Thermal
N N Y Y N Ionization at elevation 718’ and at elevation 735’
Suppression PA-12 Pre-Action N N N Y N Diesel fuel, large combustible loading.
Suppression D5 Stairwell
Wet Pipe N N N N N Stairwell system
Suppression WPS-32 Wet Pipe N N N Y N Fuel Oil/Lube Oil storage room
Feature Area 113/115
Barriers - - - - Y Lube Oil/Fuel Oil Day Tank rooms. Large combustible loading
Hose Station N N N Y N Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 310- Revision 1
Unit Fire Area Description 2 98 D6 Diesel Generator Building Note Fire Area 98 includes fire Areas: 102, 104, 106, 108, 110, 112, 114, 116, 118, 120, 122, 124, 126, and 128 Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW Pump to 11 SG
Unit 2 - 21 MDAFW Pump to 21 SG
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 2L-433) Pressurizer Level Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 2N51) U2 Excore Detection Train A 2N51 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) Unit 2 - Charging System (Train A)
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 311- Revision 1
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) to inject borated water from the RWST
Vital Auxiliaries Unit 1 - Offsite Power (1RY) supplying Electrical Distribution Train A Unit 2 - D5 supplying Electrical Distribution Train A
EEEE Title FPEE 10-006, AR 1179070-03, Evaluate Impact of Inconsistent Closing of Door 428 between D6 EDG Control Room & D6 Future Battery
Room
Summary This evaluation is being performed to evaluate the door between the D6 Emergency Diesel Generator (EDG) Control Room (Fire Area 104) and the D6 Future Battery Room (Fire Area 106) on the 695' elevation of the D51D6 Building (also known as the SBO Building). The door between the two rooms, Door 428, does not consistently close and latch due to ventilation pressure against the door. This evaluation will determine the impact of the door on safe shutdown of the plant. Administrative controls will be established to ensure that the room is only used for spare breaker storage. The fire loading in Fire Area 106, the D6 Future Battery Room, will be revised in the FHA and the Combustible Loading Calculation (Reference 4.2 and 4.6) to reflect the spare breakers and FME covers. The FHA will also document that Door 428 is not required for safe shutdown. The D5 (Train A) portion of the building is adequately separated from the D6 (Train B) portion of the building; as such, a fire in either portion will not have an impact on redundant safe shutdown capability. The separation of redundant safe shutdown capability, combined with the existing negligible combustible loading, minimal ignition sources, and the early warning smoke detection in conjunction with fire brigade response provide assurance that a fire in these fire areas will not result in damage to both trains of electrical components. This door is acceptable in its current configuration and it will not impact the ability to achieve and maintain safe shutdown.
Variances from Deterministic Requirements (VFDR)
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 312- Revision 1
Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
98 Detection 97 Ionization, Flame. Thermal
N N Y N N Diesel fuel, large combustible loading
Suppression PA-13 Pre-Action N N N N N Diesel fuel. Large combustible loading
Suppression WPS-33 Wet Pipe N N N Y N Fuel Oil/Lube Oil Storage Room
Feature Area 114/116
Barriers - - - - Y Lube Oil/Fuel Oil Day Tank rooms. Large combustible loading
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged.
Fire Area Comments
None
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 313- Revision 1
Unit Fire Area Description 2 100 #21 D5/D6 Fuel Oil Unit: 2 Fire Area: 100 Receiving Tank (South of D6 Room)
Regulatory Basis Unit 1 4.2.3.2 – Deterministic Approach Unit 2 4.2.3.2 – Deterministic Approach
Performance Goal Method of Accomplishment Comments
Decay Heat Removal (HSB) Hot:
Unit 1 - 11 TDAFW or 12 MDAFW Pump to 11 SG or 12 SG
Unit 2 - 22 TDAFW or 21 MDAFW Pump to 21 SG or 22 SG
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 314- Revision 1
Process Monitoring RCS Pressure (LOOP 1P-709) U1 Loop A RCS Wide Range Press (LOOP 1P-710) U1 Loop B RCS Wide Range Press (LOOP 2P-709) U2 Loop A RCS Wide Range Press (LOOP 2P-710) U2 Loop B RCS Wide Range Press Pressurizer Level (LOOP 1L-433) Pressurizer Level (LOOP 1L-426-RP) Pressurizer Level Red Channel (LOOP 2L-433) Pressurizer Level (LOOP 2L-426-RP) Pressurizer Level Red Channel Ex-core Neutron Monitoring (Source Range) (LOOP 1N51) U1 Excore Detection Train A 1N51 (LOOP 1N52) U1 Excore Detection Train B 1N52 (LOOP 2N51) U2 Excore Detection Train A 2N51 (LOOP 2N52) U2 Excore Detection Train B 2N52 RCS Temperature (LOOP 1T-450A) U1 RCS Loop A Hot Leg Temp (LOOP 1T-450B) U1 RCS Loop A Cold Leg Temp (LOOP 1T-451A) U1 RCS Loop B Hot Leg Temp (LOOP 1T-451B) U1 RCS Loop B Cold Leg Temp (LOOP 2T-450A) U2 RCS Loop A Hot Leg Temp (LOOP 2T-450B) U2 RCS Loop A Cold Leg Temp (LOOP 2T-451A) U2 RCS Loop B Hot Leg Temp (LOOP 2T-451B) U2 RCS Loop B Cold Leg Temp Steam Gen. Wide Range Level (LOOP 1L-487) 11 SG Wide Range Level (LOOP 1L-488) 12 SG Wide Range Level (LOOP 2L-487) 21 SG Wide Range Level (LOOP 2L-488) 22 SG Wide Range Level
Inventory and Pressure Control Unit 1 - Charging System (Train A) or Safety Injection (Train B) Unit 2 - Charging System (Train A) or Safety Injection (Train B)
Reactivity Control Unit 1 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Unit 2 - Trip reactor from the Control Room. Use Charging Pump (Train A) or Safety Injection Pump (Train B) to inject borated water from the RWST
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 315- Revision 1
Vital Auxiliaries Unit 1 - Offsite Power supplying Electrical Distribution Train A or B Unit 2 - Offsite Power supplying Electrical Distribution Train A or B
None Required Fire Protection Systems and Features
REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES
Fire Area Category ID Type Required?
Notes S L E R D
100 Detection 97 Thermal N N N N N
Suppression DA-6 Deluge N N N N N Deluge system around the circumference of the tank.
Feature - - - - - - -
Legend:
Required? S - Required for Chapter 4 Separation Criteria L - Required for NRC-Approved Licensing Action E - Required for Existing Engineering Equivalency Evaluation R - Required for Risk Significance D - Required to Maintain an Adequate Balance of Defense-in-Depth or Fire Risk Evaluation
Northern States Power - Minnesota Attachment C – Table B-3 Fire Area Transition
PINGP Page C- 316- Revision 1
Fire Suppression Effects on Nuclear Safety Performance Criteria
An automatic fire suppression system is installed in the fire area. Water will drain out doors and drains which will minimize the potential for flooding damage; such that the standing water would not affect safety related electrical equipment which is mounted on pedestals above the floor level minimizing the potential for flooding damage. Therefore, fire suppression activities are not expected to adversely affect achievement of the nuclear safety performance criteria. The PINGP fire brigade is trained to discharge water in a judicious manner and instructed to direct hose streams and portable extinguishers at the base of the fire to limit the amount of overspray beyond the immediate fire area. For this reason, fire brigade activities are not expected to fail components not already considered damaged. Fire Area Comments
None
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-1 – Revision 1
D. NEI 04-02 Non-Power Operational Modes Transition
11 Pages Attached
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-2 – Revision 1
NFPA 805, Section 1.3.1 Nuclear Safety Goal The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. FAQ 07-0040 (Revision 4) Implementing Guidance F.1 - Review Existing Outage Management Processes Define Higher Risk Evolutions (HREs), if not already defined in plant outage management procedures. The HRE definition should consider the following:
• Time to boil
• Reactor coolant system and fuel pool inventory
• Decay heat removal capability
In accordance with NUMARC 91-06
• Activities that may impact Key Safety Functions (KSFs) should be limited and strictly controlled during HREs or infrequently performed evolutions.
Review Prairie Island Nuclear Generating Plant (PINGP) Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805,” defines Higher Risk Evolutions (HRE) as: “Outage activities, plant configurations, or conditions during shutdown where the plant is more susceptible to an event causing the loss of a key safety function.” Outage Management Procedure, 5AWI 15.6.0, “Outage Scheduling and Outage Management,” implements the PINGP philosophy of outage risk management for Modes 4 through 6, and when the reactor is defueled. Procedure 5AWI 15.6.1, ”Shutdown Safety Assessment,” identifies the KSFs that need to be maintained and provides guidelines for maintaining them. The procedure identifies special requirements for reduced inventory and mid-loop conditions. These conditions are based on short times to boil, limited methods available for decay heat removal (e.g. only the Residual Heat Removal (RHR) system available), and low Reactor Coolant System (RCS) inventory. These conditions are also consistent with FAQ 07-0040 Revision 4 (ML082200528) guidance which considers these conditions to be generally the period of highest risk. The Non-Power Operations (NPO) assessment for PINGP consists of the following “higher risk evolutions” when the Plant Operating States (POSs) meet the conditions identified immediately below, thus constituting a “higher risk condition”:
• Fuel is in the reactor vessel, AND
• Thermal margin is low with time to core boil less than or equal to 40 minutes, OR
• The plant is in a reduced inventory condition (i.e. water level is 36 inches below the reactor vessel flange)
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-3 – Revision 1
Reference Documents NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance Based Fire Protection Program Under 10 CFR 50.48(c),” (Revision 2). FAQ 07-0040, “Non-Power Operations Clarifications,” (Revision 4, ML082070249). EC-20612, “Non-Power/NSCA Operations Review for NFPA 805.” 5AWI 15.6.0, “Outage Scheduling and Outage Management,” (Revision 12). 5AWI 15.6.1, “Shutdown Safety Assessment,” (Revision 26). NUMARC 91-06, “Guidelines for Industry Actions to Assess Shutdown Management,” dated December 1991. National Fire Protection Association (NFPA) Standard 805-2001, “Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.”
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-4 – Revision 1
F.2 - Identify Components and Cables The identification of systems and components to be included in this NPO Review begins with the identification of the POSs that need to be considered. Identify the various operational states that a plant goes through during NPO, and which ones are the most risk significant. Review The PINGP NPO Transition Review is documented in PINGP Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805.” The following POSs were considered for this review: POS 1: This POS starts when the RHR system is put into service. The RCS is closed such that a steam generator could be used for decay heat removal, if the secondary side of a steam generator is filled. The RCS may have a bubble in the pressurizer. The POS ends when the RCS is vented such that the steam generators cannot sustain core heat removal. The POS typically includes Mode 4 (Hot Shutdown) and portions of Mode 5 (Cold Shutdown). For the purposes of the NPO assessment this POS has been identified with two variations (configurations POS 1A and POS 1B): one with steam generators available for heat removal, and the other where the steam generators are no longer available. POS 1A: In this configuration steam generators are available along with the RHR System.
There is sufficient redundancy and diversity to remove core decay heat such that risk to core damage is significantly low and does not warrant further review under this NPO assessment. Therefore, this POS configuration will not be considered for additional engineering analysis for the NFPA 805 transition for PINGP.
POS 1B: In this configuration the steam generators are no longer capable of being used to
remove core decay heat and the RHR system is the sole means of maintaining RCS temperature. For the PINGP evaluation this POS considered that the RCS has been cooled to the point where the steam generators are no longer capable of steaming and removing decay heat. At this point the RCS has not yet been vented, and may be in the process of being taken out of solid plant conditions to remove steam and non-condensable gases from the pressurizer. Once this short duration solid plant operation is completed, the RCS will be vented, and the plant will be in POS 2. This POS configuration has been considered in the PINGP Review.
POS 2: This POS begins when the RCS has been vented such that the steam generators cannot sustain core heat removal, and an adequate vent path exists to preclude the RCS from re-pressurizing to a point where the RHR system would need to be isolated and made unavailable. This operational state will include portions of Mode 5 (Cold Shutdown) and Mode 6 (Refueling). This POS includes reduced inventory operations and mid-loop operations with a vented RCS, and has been considered in the PINGP NPO assessment. POS 3: This POS represents the shutdown condition when the refueling cavity water level is at or above the minimum level required for movement of irradiated fuel assemblies with containment as defined by the PINGP Technical Specifications. This POS occurs during Mode 6, and has been considered in the PINGP NPO assessment. PINGP Procedure 5AWI 15.6.1 identifies the KSFs that are included in this NPO assessment. Based upon the POS defined above, it was determined that not all of the KSFs from 5AWI
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-5 – Revision 1
15.6.1 need to be included in this NPO assessment. The KSFs identified in 5AWI 15.6.1, for each unit, are:
• Decay Heat Removal – RCS
• Decay Heat Removal – SFP
• Inventory Control
• Power Availability (4160 Volts, 480 Volts, 120 Volt Instrument Buses, 120 Volt UPS Loads, DC)
• Reactivity Control
• Containment
The NPO Model includes the Component Cooling water and the Cooling Water System as a supporting function to the other KSFs. These systems are not specifically identified as unique KSFs in the NPO Model. The systems are included as a supporting function to other NPO systems / functions and/or equipment. The initial identification of plant equipment required for NPO (i.e., NPO equipment) was performed from a review of the NPO flowpaths / systems / functions and equipment identified in the PINGP Operations Procedures. The NPO equipment was identified primarily from system and paths identified in 5AWI 15.6.1. Additional components were added as needed to support these paths and as necessary to provide success paths for fires occurring during the “Higher Risk Evolutions” (e.g. reduced inventory). The identification of NPO equipment also included review of the PINGP Piping and Instrumentation Drawings to select: (1) electrically operated plant equipment whose active function would be required to support the associated NPO flowpaths / systems / functions, and (2) electrically operated plant equipment whose spurious operation could be adverse to the successful performance of the associated NPO flowpaths / systems / functions. The functional attribute(s) required of each NPO component to support the associated NPO flowpaths / systems / functions were identified (i.e. Valve required to be open, to be closed, to remain operable, Motor Control Center required to be energized, Instrument Loop required “available” to provide reliable indication, etc.). Most of the plant equipment that was determined to be required for the NPO Model was already included in the NFPA 805 at-power Nuclear Safety Performance Criteria (NSPC) Model and/or the Fire PRA Model. Furthermore, most of these components were determined to have been analyzed consistently with the functional attributes required for the associated NPO flowpath / system / function (i.e., valve required to be operable for Fire PRA – same valve required to be operable for NPO). As such, most of the existing circuit analysis/cable selection for the NFPA 805 at-power NSPC Model and the Fire PRA Model was determined to be adequate for use in the NPO Model. Circuit analysis/cable selection was performed for each “NPO only” component based on the functional attribute(s) required of each NPO component. The circuit analysis/cable selection identified the plant cable(s) required to remain free of fire damage in order for the NPO component to be credited as “available” in the subsequent NPO Analysis. The cable to equipment relationships were incorporated into the SAFE database as CABLE LOGIC.
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-6 – Revision 1
Each “NPO only” component and cable was assigned plant fire zone locations consistent with those already defined for the NFPA 805 at-power NSPC Model and the Fire PRA Model. All of the location information was entered into the SAFE database. PINGP aligns with FAQ 07-0040 implementing guidance, F.2, Identify Components and Cables. Reference Documents NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance Based Fire Protection Program Under 10 CFR 50.48(c),” (Revision 2). NEI 00-01, “Guidance for Post Fire Safe Shutdown Circuit Analysis,” (Revision 1). FAQ 07-0040, “Non-Power Operations Clarifications,” (Revision 4, ML082070249). EC-20612, “Non-Power/NSCA Operations Review for NFPA 805.” 5AWI 15.6.1, “Shutdown Safety Assessment,” (Revision 26). National Fire Protection Association (NFPA) Standard 805-2001, “Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.”
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-7 – Revision 1
F.3 - Perform Fire Area Assessments (Identify pinch points) Identify locations where:
• Fires may cause damage to the equipment (and cabling) credited above, or
• KSFs are achieved solely by crediting recovery actions, e.g., alignment of gravity feed.
Fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling) thereby eliminating a pinch point. To implement this guidance perform the following tasks:
• Determine if a single fire in the fire area can cause loss of success paths for a KSF.
• Conservatively, assume the entire contents of a fire area are lost. Document the loss of success paths. Specifically identify those areas that cause loss of all success paths for a KSF.
• If fire modeling is used to limit the damage in a fire area, document that fire modeling is credited and ensure the basis for acceptability of that model (location, type, and quantity of combustible, etc.) is documented. These critical design inputs should be maintained during outage modes. Fire modeling treatment should include an assessment of safety margin to account for uncertainties/accuracy of the fire model used.
Review A deterministic fire separation analysis (i.e., assuming full area burn) was performed as documented in PINGP Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805,” to identify pinch points (i.e., areas where redundant equipment and cables credited for a given KSF fail due to fire damage). There is a total of fifty-eight (58) fire areas at the PINGP.
• Twenty-six (26) fire areas were found to have an adequate number of KSF success paths to survive the entire contents loss of the fire area.
• Thirty-two (32) fire areas were found to have pinch points resulting in the potential loss of one or more KSFs success paths.
Fire modeling was not utilized to eliminate identification of pinch point fire areas as part of the implementation process for the step F.3 guidance from FAQ 07-0040. PINGP aligns with FAQ 07-0040 implementing guidance, F.3, Perform Fire Area Assessments (Identify pinch points). Reference Documents NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance Based Fire Protection Program Under 10 CFR 50.48(c),” (Revision 2). FAQ 07-0040, “Non-Power Operations Clarifications,” (Revision 4, ML082070249). EC-20612, “Non-Power/NSCA Operations Review for NFPA 805.”
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-8 – Revision 1
NUMARC 91-06, “Guidelines for Industry Actions to Assess Shutdown Management,” dated December 1991. National Fire Protection Association (NFPA) Standard 805-2001, “Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants.”
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-9 – Revision 1
F.4 - Manage Risks Associated with Fire-Induced Vulnerabilities During the Outage The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. During those NPO evolutions where risk is relatively low, the normal fire protection program defense-in-depth actions are credited for addressing the risk impact of those fires that potentially impact one or more trains of equipment that provide a KSF required during non-power operations. The following actions are considered to be adequate to address minor losses of system capability or redundancy during those NPO evolutions where risk is relatively low:
• Control of Ignition Sources
o Hot Work (cutting, welding and/or grinding)
o Temporary Electrical Installations
o Electric portable space heaters
• Control of Combustibles
o Transient fire hazards
o Modifications
o Flammable and Combustible liquids and gases
• Compensatory Actions for fire protection system impairments
o Openings in fire barriers
o Inoperable fire detectors or detection systems
o Inoperable fire suppression systems
o Housekeeping
As required by NFPA 805 Chapter 3, the Fire Protection Program defense-in-depth administrative programs described above are in place during all NPO modes. During those NPO evolutions that are defined as HREs: Additional fire protection defense in depth measures will be taken during HREs by:
• Managing risk in fire areas that contain known pinch points (all success paths for a KSF subject to damage by a fire).
• Managing risk in fire areas where the pinch points may arise because of equipment taken out of service.
NUMARC 91-06 discusses the development of outage plans and schedules. A key element of that process is to ensure the KSFs perform as needed during the various outage evolutions. During outage planning, the NPO Fire Area Assessment should be reviewed to identify areas of single-point KSF vulnerability during higher risk evolutions to develop any needed contingency
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-10 – Revision 1
plans/actions. For those areas consider combinations of the following options to reduce fire risk, depending upon the significance of the potential damage:
• Prohibition or limitation of hot work in fire areas during periods of increased vulnerability.
• Verification of operable detection and /or suppression in the vulnerable areas.
• Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability.
• Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position).
• Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability.
• Use of recovery actions to mitigate potential losses of key safety functions.
• Identification and monitoring in-situ ignition sources for “fire precursors” (e.g., equipment temperatures).
• Reschedule the work to a period with lower risk or higher DID.
In addition, for KSF Equipment removed from service during the HREs the impact should be evaluated based on KSF equipment status and the NPO Fire Area Assessment to develop needed contingency plans/actions. Review A KSF pinch point analysis was performed for all PINGP fire areas in accordance with NFPA 805 and NRC FAQ 07-0040 Rev. 4 guidance. For fire areas where the pinch point analysis identified areas of single-point KSF vulnerability and higher risk, combinations of the following options to reduce fire risk were considered, depending upon the significance of the potential damage:
• Prohibition or limitation of hot work in fire areas during periods of increased vulnerability.
• Verification of operable fire detection and /or suppression in the vulnerable areas.
• Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability.
• Plant configuration changes (e.g., removing power from equipment once it is placed in its desired position).
• Provision of additional fire patrols at periodic intervals or other appropriate compensatory measures (such as surveillance cameras) during increased vulnerability.
• Use of recovery actions to mitigate potential losses of key safety functions.
• Identification and monitoring in-situ ignition sources for “fire precursors” (e.g., equipment temperatures).
• Reschedule the work to a period with lower risk or higher DID.
Note: Operator actions taken to mitigate the loss of a KSF are credited in the NPO analysis contained within PINGP Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805,” contemporaneous with this LAR submittal.
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-11 – Revision 1
PINGP procedure 5AWI 3.13.0, “Fire Protection Program,” will be revised to contain an overview of the NPO requirements, the commitments for implementation of the NPO risk reduction actions required by PINGP Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805,” and a road map to identify the site specific implementing procedures used to implement the NPO requirements. Additional fire protection procedures that will be revised to implement NPO requirements include the following (See Attachment S):
• 5AWI 3.13.3, “Hot Work,” contains controls to establish fire watches for the hot work activities including all plant operating states within the NPO.
• F5 Appendix K, “Fire Protection Systems Functional Requirements” contains the compensatory actions to be implemented should a fire protection system required to be operable during HRE periods be found to be impaired.
• EM 3.4.1, “Review of Proposed Changes to the Fire Protection Program” contains guidance to ensure that changes to the fire protection program are reviewed for impact to the NPO requirements and risk reduction actions.
• 5AWI 15.6.1, “Shutdown Safety Assessment,” contains discussion on HREs, NFPA 805 and the NPO requirements as part of risk management.
Guidance is contained within the outage control procedures to ensure that upon entry into the NPO plant operating states the outage roving fire watches are established. No specific requirements are necessary for the hot work controls because they are in place in all plant operating states. Additional guidance and controls are in place to ensure the HRE risk reduction tools are implemented prior to entry into a plant HRE. Guidance is also in place to monitor the plant state (T-Boil Times) to determine when the HRE is exited. PINGP outage procedures that will be revised to implement NPO guidance include the following (see Attachment S):
• 1C1.6, “Shutdown Operations – Unit 1,” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
• 2C1.6, “Shutdown Operations – Unit 2,” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
• 1C4.1, “RCS Inventory Control Pre-refueling,” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
• 2C4.1, “RCS Inventory Control Pre-refueling,” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
• 1C4.2, “RCS Inventory Control – Post Refueling” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
Northern States Power Company Attachment D – Non-Power Operational Modes Transition
PINGP Page D-12 – Revision 1
• 2C4.2, “RCS Inventory Control – Post Refueling” revise to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
• H24.1, “Assessment and Management of Risk Associated With Maintenance Activities,” contains guidance to consider potential system unavailability as a result of a fire when developing a Key Safety Function Availability Checklist for a plant configuration change.
Reference Documents NEI 04-02, “Guidance for Implementing a Risk-Informed, Performance Based Fire Protection Program Under 10 CFR 50.48(c).” (Revision 2). FAQ 07-0040, “Non-Power Operations Clarifications,” (Revision 4, ML082200528). PINGP 5AWI 3.13.0, “Fire Protection Program,” (Revision 21). 5AWI 3.13.3, “Hot Work,” (Revision 3). F5 Appendix K, “Fire Protection Systems Functional Requirements” (Revision 15). EM 3.4.1, “Review of Proposed Changes to the Fire Protection Program” (Revision 2). 5AWI 15.6.1, “Shutdown Safety Assessment,” (Revision 26). 1C1.6, “Shutdown Operations – Unit 1,” (Revision 24). 2C1.6, “Shutdown Operations – Unit 2,” (Revision 25). 1C4.1, “RCS Inventory Control Pre-refueling,” (Revision 26). 2C4.1, “RCS Inventory Control Pre-refueling,” (Revision 31). 1C4.2, “RCS Inventory Control – Post Refueling,” (Revision 29). 2C4.2, “RCS Inventory Control – Post Refueling,” (Revision 30). H24.1, “Assessment and Management of Risk Associated With Maintenance Activities,” (Revision 15). PINGP Engineering Evaluation EC-20612, “Non-Power/NSCA Operations Review for NFPA 805.”
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-1 – Revision 1
E. NEI 04-02 Radioactive Release Transition
36 Pages Attached
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-2 – Revision 1
Fire Area Compartmentation and Screening The PINGP Radiation Release evaluation is performed and documented on a Fire Area basis. This screening step is provided to evaluate each fire area and determine if the fire area is within scope of this evaluation based on the potential for radiological release in the event of a fire within the fire area. A fire area is either screened in when it affects radiological release or is screened out when it cannot affect radiological release. Attachment E identifies the results of the fire area screening.
Attachment E is a combination of Table E-1, Radioactive Release Compartment Review, and Table E-2, Radioactive Release Transition Engineered Controls Review, from the LAR Template. The requisite information and content from LAR Template Tables E-1 and E-2 is provided in Attachment E.
Attachment E provides reference to Attachment S for the implementation items that will result in compliance with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205. The implementation items consist of:
• Revisions to the fire fighting strategies to identify potential cross-contamination issues for each applicable fire area and fire detection zone.
• Revisions to the fire fighting strategies and fire brigade lesson plans to provide additional instructions on the control of the spread of contamination as a result of fire fighting activities.
• Revisions to the fire fighting strategies to address control of contaminated smoke and water runoff in areas without installed, or with non-functioning, filtered ventilation controls or filtered drainage using a combination of filtered ventilation in adjacent areas, portable filtered ventilation equipment and booms to contain water spread based on input from radiation protection personnel to the fire brigade personnel.
• A combination of containerization and administrative controls will be used to limit the amount of exposed contaminated combustible materials in areas without filtered ventilation or where the spread of contaminated water to adjacent radiologically controlled areas, radiologically clean areas or to the exterior are potential concerns.
Refer to Section 4.4, Radiological Release Performance Criteria, and Attachment S for details.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-3 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
1 Containment Unit 1 Containment Annulus Unit 1 (Previously identified as Fire Area 68)
Det. Zone 10
• U1 Rx Bldg 697’ Det. Zone 20
• U1 Rx Bldg 711’ Det. Zone 29
• U1 Rx Bldg 733’ Det. Zone 32
• U1 Rx Bldg 755’ Det. Zone 21 U1 Cntmt Annulus
Yes Floor drains to Containment Sump, then pumped to Aux Bldg Aerated Drains System. Liquid is treated, filtered, processed and sampled before release. Liquids drain to sumps in floor and transferred to Aux Bldg Aerated Drain System. Liquid is treated, filtered, processed and sampled before release. Under accident conditions, liquids transferred to Containment Sump.
Containment Internal Cleanup Subsystem recirculates and filters air for Modes 1, 2, 3, 4. Containment Purge and In-Service Purge Systems filter air prior to exhaust through U1 Shield Building Stack for Modes 5 and 6. Shield Building Ventilation System recirculates air in annulus and maintains negative pressure. In exhaust mode, Shield Building Ventilation System filters air before exhausting through Shield Building vent stack.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-4 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
2 Ventilation Fan Room Det. Zone 30
• Unit 1 Aux Bldg 755’ SFP & Fuel Receipt Area
Det. Zone 108
• Aux Bldg Elevator
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
3 Water Chiller Room, Unit 1
Det. Zone 31
• 121 & 122 Cont Rm Chiller Aux Bldg 755′
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Chiller rooms are part of the Control Room Habitability Envelope and have internal air handling unit during operations and a closed loop cleanup system during an accident.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-5 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
4 Fuel Handling Area Det. Zone 8
• U1 695′ Aux Bldg
Det. Zone 33
• Fuel Loading & Spent Fuel Area
Det. Zone 30
• Unit 1 Aux Bldg 755′ SFP & Fuel Receipt Area
Waterflow: 9
• Unit 1 695’ Aux Bldg
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
No ventilation. Potential transfer of contaminated smoke to adjacent areas and to exterior.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Revised fire strategies (to be completed as identified in Attachment S) will incorporate mitigative actions to filter potentially contaminated smoke based on radiological conditions identified during the conduct of fire fighting activities. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
6 Old Admin Bldg HVAC Equip Area
Det. Zone 55
• OAB HVAC Equipment 750’
Det. Zone 66
• OAB 735’ & 750’
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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PINGP Page E-7 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
7 Old Admin Bldg Office Area
Det. Zone 23
• OAB I&C, Tech Support, 735’ & 750’
Det. Zone 66 OAB 735’ & HVAC Penthouse 750’ Det. Zone 91
• OAB West Elevator, Stairwell
Det. Zone 105
• OAB East Elevator
Det. Zone 106 OAB West Elevator
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-8 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
8 Turbine Deck Det. Zone 24
• U1 Turb Gen Brg Prot Pnl
Det. Zone 49
• Turbine Bearing Protection Unit 2 El. 735′
Det. Zone 107
• Turb/Service Building Elevator
Yes Floor drains to Turbine Building Sump, part of Waste Liquid Treatment System. Liquid sampled prior to release.
Turbine Building Ventilation System, which is not filtered.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
The only contaminated area within Fire Area 8 is the location of the spare RCP motor on the 735ft elevation of the Unit 2 Turbine Deck. An area of approximately 10ft by 10ft around the spare RCP is marked and barricaded as a contaminated area. The potential for contaminated smoke or contaminated water due to firefighting activities in this location is considered negligible. However, revised fire strategies will incorporate mitigating actions to utilize booms to contain the flow of potentially contaminated water to balance of the Turbine Deck, and to utilize portable exhaust equipment with HEPA filters to filter potentially contaminated smoke based on radiation protection input to fire brigade on radiological conditions. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
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PINGP Page E-9 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
9 Maintenance Shops Det. Zone 107
• Turb/Service Building Elevator
Det. Zone 27
• Machine Shops, 735’
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
10 ’A’ Train Event Monitoring Room
Det. Zone 26
• 480V Swgr 112 Bus Rm & Trn A Em Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
11 Unit 1 Normal SWGR & Control Rod Drive Room
Det. Zone 87
• U1 Rod Drive Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
12 OSC Room Det. Zone 25
• Records Room & Hot Instr Shop
Waterflow: 18
• Lndry, Chem & Inst Labs or Records Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
13 Control Room Det. Zone 57
• Control Room U1 & U2
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
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Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
14 Working Material & Lunch Room
Det. Zone 15
• U1 Turb Bldg 715′, Locker Room, Lunch Room, Serv Bldg Stairway, Fan Rm, & Work Area
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
15 Access Control Det. Zone 17
• Access Control Waterflow: 18
• Lndry, Chem & Inst Labs or Records Room
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release. However, there are no barriers to water spread between the potentially contaminated and clean portions of Access Control.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Revised fire strategies will incorporate mitigative actions to utilize booms to prevent flow of potentially contaminated water between the potentially contaminated and clean portions of Access Control. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
16 Train B Event Monitoring Equipment Room
Det. Zone 50
• 480V Swgr 122 Bus Rm & Trn A Em Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
17 Unit 2 Normal SWGR & Control Rod Drive Room
Det. Zone 88
• Rod Control Room Unit 2
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
18 Relay & Cable Spreading Room Units 1 & 2
Det. Zone 12
• Relay & Cable Spreading Rm Units 1 & 2
Det. Zone 14
• Old Computer Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
20 Unit 1 4.16 kV Safeguards SWGR (Bus 16)
Det. Zone 11
• Bus Rms 15 & 16 Unit 1 715′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
21 Unit 1 4.16 kV Normal SWGR (Bus 13, 14)
Det. Zone 84
• Bus Rms 13 & 14 Unit 1 715′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
22 480 V Safeguards SWGR (Bus 121)
Det. Zone 43
• 480V Bus 111 and 121 715’
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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PINGP Page E-12 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
23 Unit 2 4.16 kV Normal SWGR (Bus 23, 24)
Det. Zone 86
• Normal Switchgear Bus 23 & 24
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
24 Oil Storage Area Det. Zone 4
• U1 695′ Turb
Bldg Waterflow: 5
• U1 695′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
25 Diesel Generator Room 1
Det. Zone 82
• D1 Diesel Gen Rm
Waterflow: 7
• D1 & D2 Rooms
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
26 Diesel Generator Room 2
Det. Zone 6
• D2 Diesel Gen Rm
Waterflow: 7
• D1 & D2 Rooms
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
27 Water Conditioning Equipment Area
Det. Zone 4
• U1 695′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
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Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
28a 1GT Transformer Det. Zone 58
• 1GT Waterflow: 76
• 1GT
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
28b 2GT Transformer Det. Zone 60
• 2GT Waterflow: 78
• 2GT
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
28c 1R Transformer Det. Zone 62
• 1R Waterflow: 80
• 1R
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
28d 1M Transformer Det. Zone 59
• 1M Waterflow: 77
• 1M
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
28e 2M Transformer Det. Zone 61
• 2M Waterflow: 79
• 2M
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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PINGP Page E-14 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
28f 2RX/Y Transformer Det. Zone 96
• 2RX/Y Waterflow: 95
• 2RX/Y
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
29 Admin Bldg Elec. & Piping Room 1
Det. Zone 4
• U1 695′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
30 Admin Bldg Elec. & Piping Room 2
Det. Zone 4
• U1 695′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
31 ”A” Train Hot Shutdown Panel & Air Compressor / Auxiliary Feedwater Pump Room
Det. Zone 2
• Air Cmprsr & Aux Feed Pmp Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
32 ”B” Train Hot Shutdown Panel & Air Compressor / Auxiliary Feedwater Pump Room
Det. Zone 2
• Air Cmprsr & Aux Feed Pmp Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
33 Battery Room 11 Det. Zone 1
• 11 & 12 Batt Rms Turb Bldg
695′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
34 Battery Room 12 Det. Zone 1
• 11 & 12 Batt Rms Turb Bldg
695′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
35 Battery Room 21 Det. Zone 35
• 21 & 22 Battery Rooms
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
36 Battery Room 22 Det. Zone 35
• 21 & 22 Battery Rooms
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
37 480V Normal SWGR Room Unit 1
Det. Zone 83
• Bus Rms 150 & 160 Unit 1 695′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
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Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
38 480V Normal SWGR Room Unit 2
Det. Zone 85
• Normal Switchgear Bus 250 & 260
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
39 Rad Waste Building Det. Zone 34
• Rad Waste & Resin Disposal El. 695′
Det. Zone 81
• Rad Waste & Resin Disposal El. 695′ & El. 715′
Yes Floor drains to Radwaste Building Sumps, collected in tanks, filtered, processed and sampled prior to release.
Radwaste Building Ventilation filters air prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
40 Maintenance Storage Shed
Det. Zone - None
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-17 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
41 Screenhouse (General Area)
None No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
41A Screenhouse (DDCWP Rooms)
Det. Zone 74
• Plant Screen
House El. 670′
Det. Zone 75
• Plant Screen
House El. 695′
Waterflow: 63
• Screen House
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
41B Screenhouse (Basement)
Det. Zone 75
• Plant Screen
House El. 695′
Waterflow: 63
• Screen House
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Cooling Tower Pump House (CTPH)
Det. Zone 71
• CTPH, CTEH, Warehouse 1, Construction / Fab Shop
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Fuel Oil and Transfer House
Det. Zone 72
• Fuel Oil & Transfer House
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-18 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
46 Cooling Tower Equipment House (CTEH)
Det. Zone 71
• CTPH, CTEH, Warehouse 1, Construction / Fab Shop
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
46A Cooling Tower Transformers
None No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- 121/122 CLG Tower Control House
Det. Zone 68
• Cooling Tower Control House 121/122
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Waste Neutralizing Tank Pump House / Warehouse 2
Det. Zone 92
• Neut Tk House & Whse #2
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- 123/124 CLG Tower Control House
Det. Zone 69
• Cooling Tower Control House 123/124
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-19 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
-- Main Warehouse 1 and Fab Shop
Det. Zone 71
• CTPH, CTEH, Warehouse 1, Construction / Fab Shop
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
57 Hydrogen House Det. Zone 4
• U1 695′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
58 Aux Building Ground Floor Unit 1 & Unit 2 (Unit 2 was previously identified as Fire Area 73)
Det. Zone 8
• U1 695′ Aux Bldg
Waterflow: 9
• U1 695’ Aux Bldg
Det. Zone 40
• U2 695’ Aux Bldg
Det. Zone 108
• Aux Bldg Elevator
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-20 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
59 Aux Building Mezzanine Floor Unit 1 & Unit 2 (Unit 2 was previously identified as Fire Area 74)
Det. Zone 19
• U1 715′ Aux Bldg
Waterflow: 22
• U1 715’ Aux Bldg Penetrations
Waterflow: 41
• Auxiliary Building Unit 2 & D-3 Storage Room
Det. Zone 46
• Auxiliary Building Unit 2
El. 715′
Waterflow: 48 Annulus Penetration Unit 2 Det. Zone 108
• Aux Bldg Elevator
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-21 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
60 Aux Building Operating Level Unit 1
Det. Zone 28
• Unit 1 Aux Bldg 735′
Det. Zone 108
• Aux Bldg Elevator
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
61 Aux Bldg Anti “C” Clothing Area
Det. Zone 28
• Unit 1 Aux Bldg 735′
Waterflow: 103
• E/W Demin Removal Area
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Laundry Room Filters and Exhaust Fan can be operated locally to prevent gaseous effluent from escaping.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Revised fire strategies (to be completed as identified in Attachment S) will incorporate mitigative actions to filter potentially contaminated smoke based on radiological conditions identified during the conduct of fire fighting activities. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-22 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
61A Aux. Bldg Hatch Area Det. Zone 30
• Unit 1 Aux Bldg 755′ SFP & Fuel Receipt Area
Det. Zone 53
• Auxiliary Building Unit 2 & West Side Fuel Handling Area El. 755′
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
No ventilation. Potential transfer of contaminated smoke to adjacent areas and to exterior.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Revised fire strategies (to be completed as identified in Attachment S) will incorporate mitigative actions to filter potentially contaminated smoke based on radiological conditions identified during the conduct of fire fighting activities. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
62 Spent Fuel Pool Area Det. Zone 30
• Unit 1 Aux Bldg 755′ SFP & Fuel Receipt Area
Yes No drains. Floor drains in adjacent areas route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Spent Fuel Pool Normal Ventilation System filters air prior to release. Spent Fuel Pool Special Ventilation System filters air before exhausting through Shield Building vent stack.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-23 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
63 Filter Room Det. Zone - None
• Filter Room (Refer to FA 60 and FA 75)
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Make-Up Fans from the control room to prevent gaseous effluent from escaping. The filter room exhaust fan can be turned off locally. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-24 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
64 Aux Bldg Low Level Decay Area Unit 1
Det. Zone 8
• U1 695′ Aux Bldg
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Make-Up Fans from the control room and locally stop the Filter Room Exhaust Fan to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-25 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
65 Spent Fuel Pool Heat Exchangers & Pumps
Det. Zone – None
• SFP Hx Room (Refer to U1 715’ Aux Bldg, Det. Zone 19 )
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
66 D3 Lunch Room Det. Zone 39
• D-3 Storage/Diesel Room Detection
Waterflow: 41
• Auxiliary Building Unit 2 & D-3 Storage Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
67 Resin Disposal Area Det. Zone 34
• Rad Waste & Resin Disposal El. 695′
Det. Zone 81
• Rad Waste & Resin Disposal El. 695′ & El. 715′
Yes Floor drains to Radwaste Building Sumps, collected in tanks, filtered, processed and sampled prior to release.
Radwaste Building Ventilation System filters air prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-26 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
69 Turbine Bldg Ground & Mezzanine Floors Unit 1
Det. Zone 3
• FW Pmp Turb
Bldg 695′
Det. Zone 4
• U1 695′ Turb
Bldg Det. Zone 15
• U1 Turb Bldg
715′, Locker
Room, Lunch Room, Serv Bldg Stairway, Fan Rm, & Work Area
Det. Zone 107
• Turb/Service Building Elevator
Waterflow: 5
• U1 695′ Turb
Bldg Waterflow: 16
• U1 715′ Turb
Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-27 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
70 Turbine Bldg Ground & Mezzanine Floors Unit 2
Det. Zone 36
• Feed Water Pump Area Unit 2
Det. Zone 37
• Turbine Building
Unit 2 El. 695′
Det. Zone 44
• Turbine Building
Unit 2 El. 715′
Waterflow: 38
• Turbine Building
Unit 2. 695′
Waterflow: 45
• Turbine Building Unit 2, 715’
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-28 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
71 Containment Unit 2 Containment Annulus Unit 2 (Previously identified as Fire Area 72)
Det. Zone 42
• U2 Rx Bldg 697’ Det. Zone 52
• U2 Rx Bldg 733’ Det. Zone 54
• U2 Rx Bldg 755’ Det. Zone 56 U2 Rx Bldg 711’ Det. Zone 47 U2 Cntmt Annulus
Yes Floor drains to Containment Sump, then pumped to Aux Bldg Aerated Drains System. Liquid is treated, filtered, processed and sampled before release. Liquids drain to sumps in floor and transferred to Aux Bldg Aerated Drain System. Liquid is treated, filtered, processed and sampled before release. Under accident conditions, liquids transferred to Containment Sump.
Containment Internal Cleanup Subsystem recirculates and filters air for Modes 1, 2, 3, 4. Containment Purge and In-Service Purge Systems filter air prior to exhaust through U2 Shield Building Stack for Modes 5 and 6. Shield Building Ventilation System recirculates air in annulus and maintains negative pressure. In exhaust mode, Shield Building Ventilation System filters air before exhausting through Shield Building vent stack.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-29 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
75 Aux Bldg Operating Level Unit 2
Det. Zone 51
• Auxiliary Building Unit 2
El. 735′
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
76 Vent and Fan Room Unit 2
Det. Zone 30
• Unit 1 Aux Bldg
755′ SFP &
Fuel Receipt Area
Det. Zone 53 Auxiliary Building Unit 2 & West Side Fuel
• Handling Area
El. 755′
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-30 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
77 Aux Bldg Low Level Decay Unit 2
None Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
78 Waste Gas Compressor Area
Det. Zone 33
• Fuel Loading & Spent Fuel Area
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-31 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
79 480 V SFGD SWGR Room (Bus 112)
Det. Zone 26
• 480V Swgr 112 Bus Rm & Trn A Em Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
80 480 V SWGR Room (Bus 111)
Det. Zone 43
• 480V Switch Gear 111 and 121
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
81 4.16kV Safeguard SWGR Room (Bus 15)
Det. Zone 11
• Bus Rms 15 &
16 Unit 1 715′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
82 480 V Safeguard SWGR Room (Bus 122)
Det. Zone 50
• 480V Switch Gear 122 Bus Room & Train B EM Room
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
83 Operators Lounge Det. Zone 64
• Operators Study Area / Security Office
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-32 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
84 Counting Room & Labs
None (FA 59, Zone 19) Waterflow: 18
• Lndry, Chem & Inst Labs or Records Room
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Area is positively pressurized with respect to rest of Auxiliary Building. Additional filters and exhaust fans are installed, and air is filtered prior to discharge. Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-33 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
85 Hold-up Tank Area / Demineralizer Area
Det. Zone 8
• U1 695′ Aux
Bldg
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Auxiliary Building Ventilation System maintains negative pressure. Air is monitored prior to release. North side 695ft and entire 715ft elevation portions: The control room can turn off the Auxiliary Building Exhaust and Make-Up Fans from the control room to prevent gaseous effluent from escaping. Auxiliary Building Special Ventilation System Air is filtered prior to release. South side 695ft portions: The control room can turn off the Auxiliary Building Make-Up Fans from the control room and turn off the Filter Room Exhaust fan locally to prevent gaseous effluent from escaping.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Revised fire strategies will incorporate mitigative actions to utilize Filter Room Ventilation to filter potentially contaminated smoke occurring from fires on south side of 695ft elevation based on radiation protection input to fire brigade on radiological conditions. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-34 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
86 Intake Screenhouse None No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Radiation Monitor Station
Det. Zone 70
• Radiation Monitoring Station / De-icing Pump House / Envir. Lab / Intake Screen House
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Deep Well Pump House 1 & 2
Det. Zone 73
• Deep Well Pump House 1 & 2
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
-- Deep Well Pump House 1 & 2
Det. Zone 73
• Deep Well Pump House 1 & 2
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
89 Guard House Det. Zone 89
• Guard House & Emerg Gen Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-35 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
90 Emergency Generator Bldg
Det. Zone 89
• Guard House & Emerg Gen Bldg
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
92 Water Chiller Room Unit 2
Det. Zone 31
• 121 & 122 Cont Rm Chiller Aux
Bldg 755′
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
Chiller rooms are part of the Control Room Habitability Envelope and have internal air handling unit during operations and a closed loop cleanup system during an accident.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-36 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
93 Low Level Rad Waste Area
Det. Zone 104
• Low Level Rad Waste – Building
Waterflow: 101
• Low Level Rad Waste Storage Bldg & Warehouse
Yes Floor drains route to Aerated Drains System. Liquid is treated, filtered, processed and sampled before release.
No ventilation. Potential transfer of contaminated smoke to exterior.
Upon completion of implementation items identified in Attachment S, fire brigade training materials will meet NFPA radioactive release performance criteria
A combination of containerization and administrative controls will limit the amount of exposed contaminated combustible materials. Revised fire strategies (to be completed as identified in Attachment S) will incorporate mitigative actions to filter potentially contaminated smoke based on radiological conditions identified during the conduct of fire fighting activities. Based on use of revised fire strategies and training materials (to be completed as identified in Attachment S), the PINGP approach will meet NFPA 805 radioactive release performance criteria.
94 Service Building/Computer Room
Det. Zone 94
• Service Bldg
695′ - 715′
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment E – Radioactive Release Transition
PINGP Page E-37 – Revision 1
Fire Area
Description Fire Strategies Detection Zones
Screened In?
Engineering Controls Training Review Results
Conclusions Liquid Airborne
97 D5 Diesel Generator Building (Previously identified as Fire Areas 97, 99, 101, 103, 105, 107, 109, 111, 113, 115, 117, 119, 123, 125, & 127)
Det. Zone 97
• D5/D6 Diesel Building
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
98 D6 Diesel Generator Building (Previously identified as Fire Areas 98, 102, 104, 106, 108, 110, 112, 114, 116, 118, 120, 122, 124, 126 & 128)
Det. Zone 97
• D5/D6 Diesel Building
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
100 D5/D6Fuel Oil Receiving Tank 21
Det. Zone 97
• D5/D6 Diesel Building
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
131 New Admin Bldg Det. Zone 99
• New Admin Building
No N/A N/A N/A This area is outside of the permanent RCA and is not used for permanent or temporary storage of radiological materials. A fire in this area will not result in radiological release.
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-1 – Revision 1
F. Fire-Induced Multiple Spurious Operations Resolution
6 Pages Attached
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-2 – Revision 1
MSO Process Summary
The following provides the guidance from FAQ 07-0038, Revision 3, along with the process and results.
Step 1 – Identify potential MSOs of concern
Information sources that may be used as input include:
• Generic lists of Multiple Spurious Operations (MSOs, e.g., from Owners Groups and/or later versions of NEI 00-01, if endorsed by NRC for use in assessing MSOs).
• Self assessment results (e.g., NEI 04-06 assessments performed to address RIS 2004-03).
• PRA insights (e.g., NEI 00-01 Revision 1, Appendix F).
The Prairie Island Nuclear Generating Plant (PINGP) MSO identification process was performed as two complementary tasks. First, a systematic review of plant P&IDs was performed to identify potential spurious operation combinations that can lead to initiating events and that can also impact the function of needed mitigating systems. Second, an Expert Panel was convened to consider potential MSOs from a number of generic and plant specific sources. The sources of information used as inputs for this process are listed below.
• Pressurized Water Reactor (PWR) Generic MSO List, Rev. 1, May 2009, contained in Appendix G of NEI 00-01, Rev. 2, May 2009.
• PINGP Safe Shutdown Analysis, GEN-PI-026.
• PINGP Internal Events PRA Model Revision 3.1.
• System P&IDs and Electrical Drawings.
• PINGP training material for relevant systems.
Step 2 – Conduct an expert panel to assess plant specific vulnerabilities (e.g., per NEI 00-01, Rev. 1 Section F.4.2).
The expert panel should focus on system and component interactions that could impact nuclear safety. This information will be used in later tasks to identify cables and potential locations where vulnerabilities could exist.
The documentation of the results of the expert panel should include how the expert panel was conducted including the members of the expert panel, their experience, education, and areas of expertise. The documentation should include the list of MSOs reviewed as well as the source for each MSO. This documentation should provide a list of the MSOs that were included in the PRA and a separate list of MSOs that were not kept for further analysis (and the reasons for rejecting these MSOs for further analysis).
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-3 – Revision 1
Describe the expert panel process (e.g., when it was held, what training was provided to the panel members, what analyses were reviewed to identify MSOs, how was consensus achieved on which MSOs to keep and any dispute resolution process criteria used in decision process, etc.).
[Note: The physical location of the cables of concern (e.g., fire zone/area routing of the identified MSO cables), if known, may be used at this step in the process to focus the scope of the detailed review in further steps].
Results of Step 2:
An MSO Expert Panel was conducted at the PINGP site in December of 2009. The purpose of the Expert Panel was to review the applicable industry developed Generic Owner’s Group List of MSOs for applicability to PINGP. The Expert Panel commented on whether or not applicable MSOs were accounted for in the plant PRA and Safe Shutdown Analyses. A training session for the panel members was conducted prior to starting the actual assessment. The results of the MSO Expert Panel were documented in V.SMN.13.001, “Fire PRA Multiple Spurious Operations (MSO) Report.” The report also includes:
• The presentation that was used as the training materials.
• The areas of expertise for each of the MSO Expert Panel participants.
• A list of the generic MSOs that were reviewed.
In addition to the Expert Panel meeting, a systematic review of system drawings was performed to identify single and multiple spurious operations of components that can lead to initiating events and that can also impact the function of needed mitigating systems. The purpose of this review was to identify relevant combinations of fire induced single spurious operations (SSOs) and MSOs of equipment which could result in a functional failure leading to an increase in core damage frequency or large early release frequency, and ensure they are evaluated within the context of the PINGP Fire PRA. The systematic review scope included all mechanical systems, including water, oil and air, but specifically excluded electrical systems. Spurious operations of electrical systems (i.e., spurious breaker operation) are already evaluated directly in the PRA and are considered, at least in part, in the PWR Owners Group’s Generic MSO List. The output from this review includes the following:
• A summary of the systematic review of systems for spurious operations combinations potentially affecting initiating events and mitigating systems.
• A listing of the resulting SSO and MSO combinations identified.
V.SMN.13.001 documents the methodology used for this systematic review and the resulting SSO and MSO combinations.
Another MSO Expert Panel was conducted at the PINGP site in October of 2013. The Expert Panel was conducted to review PWR Owners Group’s Generic electrical MSOs as detailed in NEI 00-01, Revision 2, to validate previous disposition efforts, discuss additional related plant-specific MSOs, and provide disposition as required. Training involved Expert Panel members reviewing the previous MSO Expert Panel Overview prior to starting the assessment. The results of the MSO Expert Panel were
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-4 – Revision 1
documented in PINGP Engineering Change EC 23660, “Focused-Scope Electrical Expert Panel for Addressing Multiple Spurious Operations Related to Generic MSOs 47, 48 and 49.” The report includes:
• The presentation that was used as the training materials.
• Fire Induced MSOs related to PWR Owners Group’s Generic electrical MSOs.
• The results of the Focused-Scope Expert Panel, including required disposition when necessary.
Step 3 – Update the Fire PRA model and NSCA to include the MSOs of concern.
This includes the:
• Identification of equipment (NUREG/CR-6850 Task 2).
• Identification of cables that, if damaged by fire, could result in the spurious operation (NUREG/CR-6850 Task 3, Task 9).
• Identification of the routing of cables identified above, including associating that routing with fire areas, fire zones and/or Fire PRA physical analysis units, as applicable.
Include the equipment/cables of concern in the Nuclear Safety Capability Assessment (NSCA). Including the equipment and cable information in the NSCA does not necessarily imply that the interaction is possible since separation/protection may exist throughout the plant fire areas such that the interaction is not possible.
Note: Instances may exist where conditions associated with MSOs do not require update of the Fire PRA and NSCA analysis. For example, Fire PRA analysis in NUREG/CR-6850 Task 2, Component Selection, may determine that the particular interaction may not lead to core damage, or pre-existing equipment and cable routing information may determine that the particular MSO interaction is not physically possible. In other instances, the update of the PRA may not be warranted if the contribution is negligible. The rationale for exclusion of identified MSOs from the Fire PRA and NSCA should be documented and the configuration control mechanisms should be reviewed to provide reasonable confidence that the exclusion basis will remain valid.
Results of Step 3:
The results of the Expert Panel and systematic review were fed into the PINGP Equipment Selection (ES) task (NUREG/CR-6850 Task 2). This task included components susceptible to single and multiple spurious operations identified in the post-fire safe shutdown analysis as well as those from the expert panel and systematic reviews. Cable selection and circuit analysis (NUREG/CR-6850 Tasks 3 and 9) were then performed for those components that did not already have this performed for the current Safe Shutdown Analysis. Components susceptible to spurious operation that were not already included in the PINGP Fire PRA model were added to the model at either the system level or the top logic level representing the probability of proceeding down various event tree paths (NUREG/CR-6850 Task 5). Probabilities for important SSO basic events were generated in the Circuit Failure Mode Likelihood Analysis (NUREG/CR-6850 Task 10). Probabilities for MSO events due to multiple cable hot shorts are also derived from the Circuit Failure Mode Likelihood Analysis results.
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-5 – Revision 1
The results of the Fire PRA model development including the effects of equipment spurious operations are documented in the following PINGP Fire PRA notebooks:
• FPRA-PI-ES, Equipment Selection Notebook.
• FPRA-PI-CS, Cable Selection and Circuit Analysis Notebook.
The last notebook includes the logic changes made to the Fire PRA model to account for MSO scenarios relevant to fire but not already captured by the Internal Events PRA. The Fire Induced Risk Model along with the outputs from the cable selection and circuit failure likelihood analyses were then used to evaluate the impact of fire scenarios in individual Fire Areas and Fire Compartments to support the NSCA.
The MSO combinations of components of concern were evaluated as part of the PINGP NSCA. As necessary, components were added to the NSCA Equipment List and Logics (EC 20612, “Non-Power/NSCA Operations Review for NFPA 805”), and the appropriate circuit analysis and cable routing were performed.
Step 4 – Evaluate for NFPA 805 Compliance
The MSO combinations included in the NSCA should be evaluated with respect to compliance with the deterministic requirements of NFPA 805, as discussed in Section 4.2.3 of NFPA 805. For those situations in which the MSO combination does not meet the deterministic requirements of NFPA 805 and therefore represents a variance from deterministic requirements (VFDR), the issue with the components and associated cables should be mitigated by other means (e.g., performance-based approach per Section 4.2.4 of NFPA 805, plant modification, etc.).
The performance-based approach may include the use of feasible and reliable recovery actions. The use of recovery actions to demonstrate the availability of a success path for the nuclear safety performance criteria requires that the additional risk presented by the use of these recovery actions be evaluated (NFPA 805 Section 4.2.4).
Results of Step 4:
The PINGP PRA quantified the fire-induced risk model containing the MSO failure modes. The quantification addressed the specific electrical cables and the failure mode in each fire area and fire zone that was quantified. Thus, the MSO contribution is included in the fire PRA results, and in the fire PRA results associated with evaluation of VFDRs as documented in applicable fire risk evaluations.
The MSO combinations of components of concern were also evaluated as part of the PINGP NSCA. As part of the review of current fire area safe shutdown strategies, components with a potential for a spurious operation were reviewed. During this review process, the systems were reviewed to determine the overall impact. The methodology used in the review process is detailed in Engineering Change Evaluation EC 20612, “Non-Power/NSCA Operations Review for NFPA 805.”
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-6 – Revision 1
For cases where the MSO combination of components did not meet the requirements for deterministic compliance, the MSO combination of components were identified as VFDRs and added to the scope of the fire risk evaluations.
The process and results associated with the performance of fire risk evaluations are summarized in Section 4.5 of the Transition Report.
Step 5 - Document Results
The results of the process should be documented. The results should provide a detailed description of the MSO identification, analysis, disposition, and evaluation results (e.g., references used to identify MSOs; the composition of the expert panel, the expert panel process, and the results of the expert panel process; disposition and evaluation results for each MSO, etc.). The high level methodology utilized as part of the transition process should be included in the 10 CFR 50.48(c) License Amendment Request/Transition Report.
Results of Step 5:
The PINGP MSO methodology and results are documented in the following:
• V.SMN.13.001, Fire PRA Multiple Spurious Operations (MSO) Report.
• FPRA-PI-ES, Equipment Selection Notebook.
• FPRA-PI-CS, Cable Selection and Circuit Analysis Notebook.
• Engineering Change EC 23660, Focused-Scope Electrical Expert Panel for Addressing Multiple Spurious Operations Related to Generic MSOs 47, 48 and 49.
As part of Step 4 of the process outlined above, MSO combinations were reviewed for their impact on deterministic compliance (i.e., fire area reviews to determine if a fire scenario could result in the potential MSO combinations). During this process, VFDRs were identified where the deterministic requirements of NFPA 805 Section 4.2.3 were not met. These VFDRs were addressed by demonstrating compliance with the performance-based approach of Section 4.2.4 of NFPA 805 (See Section 4.5 and Attachment C).
Note that the spurious operations reviewed as part of the process included components that were part of the original PINGP 10 CFR 50 Appendix R post-fire safe shutdown analysis, as well as components and interactions that were added following a plant-specific review of functional failures and evolved industry issues. No specific distinction is made in the program documentation whether the interaction is related to an SSO or MSO since the risk-informed approach using the Fire PRA provides an integrated plant response model.
Fire-induced MSOs are included in the fire PRA model, and their associated risk is included in the quantification of each fire scenario, the total plant fire risk, and
Northern States Power - Minnesota Attachment F – Fire-Induced MSOs Resolution
PINGP Page F-7 – Revision 1
evaluation of each VFDR. The VFDRs are identified in Attachment C and a summary of the Fire PRA results is provided in Attachment W.
Northern States Power Company Attachment G – Recovery Action Transition
PINGP Page G-1 – Revision 1
G. Recovery Actions Transition
24 Pages Attached
Northern States Power Company Attachment G – Recovery Action Transition
PINGP Page G-2 – Revision 1
In accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions). The methodology consisted of the following steps:
Step 1: Define the primary control station(s) and determine which pre-transition Operator Manual Actions (OMAs) are taken at primary control station(s) (Activities that occur in the Main Control Room are not considered pre-transition OMAs). Activities that take place at primary control station(s) or in the Main Control Room are not recovery actions, by definition.
Step 2: Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth).
Step 3: Evaluate the additional risk presented by the use of recovery actions required to demonstrate the availability of a success path.
Step 4: Evaluate the feasibility of the recovery actions.
Step 5: Evaluate the reliability of the recovery actions.
An overview of these steps and the results of their implementation are provided below.
Step 1 - Clearly define the primary control station(s) and determine which pre-transition OMAs are taken at primary control station(s)
The first task in the process of determining the post-transition population of recovery actions was to apply the NFPA 805 definition of recovery action and the RG 1.205 definition of primary control station to determine those activities that are taken at primary control station(s).
Results of Step 1:
PINGP does not have alternate or dedicated shutdown controls that meet the definition of a primary control station (PCS) as defined by RG 1.205. No credit has been taken for PCS actions in the Fire PRA.
Step 2 – Determine the population of recovery actions that are required to resolve VFDRs (to meet the risk or defense-in-depth criteria)
On a fire area basis all VFDRs were identified in the NEI 04-02, Table B-3 (See Attachment C). Each VFDR not brought into compliance with the deterministic approach was evaluated using the performance-based approach of NFPA 805, Section 4.2.4. The performance-based evaluations resulted in the need for recovery actions to meet the risk acceptance criteria or maintain a sufficient level of defense-in-depth.
Results of Step 2:
The final set of recovery actions are provided in Table G-1 - Recovery Actions.
Step 3: Evaluate the Additional Risk of the Use of Recovery Actions
NFPA 805 Section 4.2.3.1 does not allow recovery actions when using the deterministic approach to meet the nuclear safety performance criteria. However, the use of recovery actions is allowed by NFPA 805 using a risk informed, performance-based approach,
Northern States Power Company Attachment G – Recovery Action Transition
PINGP Page G-3 – Revision 1
provided that the additional risk presented by the recovery actions is evaluated in accordance with NFPA 805 Section 4.2.4.
Results of Step 3:
The set of recovery actions that are necessary to demonstrate the availability of a success path for the nuclear safety performance criteria (See Table G-1) were evaluated for additional risk using the process described in NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205 and compared against the guidelines of RG 1.174 and RG 1.205. The additional risk is provided in Attachment W.
All of the recovery actions were reviewed for adverse impact and dispositioned in fire area-specific Fire Risk Evaluation engineering evaluations. None of the recovery actions were found to have an adverse impact on the Fire PRA.
Step 4: Evaluate the Feasibility of Recovery Actions
Recovery actions were evaluated against the feasibility criteria provided in the NEI 04-02, FAQ 07-0030, Revision 5, and RG 1.205.
Results of Step 4:
Each of the feasibility criteria in FAQ 07-0030 were assessed for the recovery actions listed in Table G-1 with exception to recovery actions which place more than one instrument bus on its associated alternate source. Control room abandonment was found to be the bounding fire scenario and the results of the assessment are included in Calculation GEN-PI-055 Rev. 1, “10CFR50 Appendix R Manual Action Feasibility Study.” This calculation contains the required time constraints in which to perform the recovery actions. Feasibility for recovery actions allowing more than one instrument bus to be placed on its associated alternate source will be assessed by EC 23840, “Panel 117 and Panel 217 Load Calculation for Supplying Two Instrument Buses.”
Implementation items resulting from the feasibility evaluation include:
• Development/revision of procedures.
• Completion of EC 23840, “Panel 117 and Panel 217 Load Calculation for Supplying Two Instrument Buses.”
These items are included in Table S-3.
Step 5: Evaluate the Reliability of Recovery Actions
The evaluation of the reliability of recovery actions depends upon its characterization.
The reliability of recovery actions that were modeled specifically in the Fire PRA were addressed using Fire PRA methods (i.e., HRA).
The reliability of recovery actions not modeled specifically in the Fire PRA is bounded by the treatment of additional risk associated with the applicable VFDR. In calculating the additional risk of the VFDR, the compliant case recovers the fire-induced failure(s) as if the variant condition no longer exists. The resulting delta risk between the variant and compliant condition bounds any additional risk for the recovery action even if that recovery action were modeled.
Northern States Power Company Attachment G – Recovery Action Transition
PINGP Page G-4 – Revision 1
Results of Step 5:
The reliability of recovery actions that are being modeled specifically in the Fire PRA has been addressed using Fire PRA methods as documented in PINGP Calculation FPRA-PI-FHRA, “Fire Human Reliability Analysis.” Bounding reliability results are documented in Attachment W. PINGP procedure F5 Appendix D, “Impact of Fire Outside Control/Relay Room” will be updated to incorporate credited Recovery Actions. This implementation action is included in Table S-3.
Northern States Power Company Attachment G – Recovery Actions Transition
De-energize MV-32076 at MCC 1KA2, BKR 121B-28, located in Fire
Area 058.
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MCC 1A1 Motor Control Center 1A BUS 1
De-energize MV-32077 at MCC 1A1, BKR 111E-9, located in Fire Area 032.
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MCC 1A2 Motor Control Center 1A BUS 2
De-energize MV-32078 at MCC 1A2, BKR 121E-9, located in Fire Area 032.
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MCC 1K2 Motor Control Center 1K BUS 2
Manually trip MTR 121J-1 (11 CHG PMP) at MCC 1K2, BKR 121J-1, in Fire
Area 058.
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MCC 1K2 Motor Control Center 1K BUS 2
Place MCC BKR 1K2-C3, 13 CHARGING PUMP, in the “OFF”
position, in Fire Area 058..
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MTR 111J-1 Local Panel
12 Charging Pump Local Panel
Take local control of MTR 111J-1, 12 charging pump at panel 70810 by
placing CS-7081001 in "LOCAL" and pressing CS-7081002 once to energize the VFD and a second time to start the 12 charging pump at the local panel in 12 charging pump room in Fire Area
058
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MTR 111J-1 VFD
12 Charging Pump Variable Frequency Drive Cabinet
Manually operate MTR 111J-1 (12 CHARGING PUMP) at the VFD panel
by MCC 1K1 in Fire Area 058 by placing the LOC/REM SEL SW in the “LOCAL”
position.
VFDR-013-1-02 VFDR-018-1-02
RA
Northern States Power Company Attachment G – Recovery Actions Transition
Place MCC 1K1-D3, 11 RWST TO 11 RHR PUMP ISOL MV-32084, to the
“OFF” position (Location: G.2/5.2/695′ Aux Bldg near RHR Pits).
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MV-32084 11 RWST to 11 RHR Pump ISOL MV
Manually CLOSE MV-32084, RFLG WTR TO 11 RSDL HT RMVL PMP ISOL
MV (Location: 11 RHR Pit).
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MCC 1K2 Motor Control Center 1K Bus 2
Place MCC 1K2-A3, 11 RWST TO 12 RHR PUMP ISOL MV-32085, to the
“OFF” position (Location: G.8/6.5/695′ Aux Bldg near the Charging Pumps).
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
MV-32085 11 RWST to 12 RHR Pump ISOL MV
Manually CLOSE MV-32085, RFLG WTR TO 12 RSDL HT RMVL PMP ISOL
MV (Location: 12 RHR Pit).
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
PNL 11 DC Distribution Panel 11 De-energize PNL 191 at PNL 11, breaker 11-18, located in Fire Area 033 in order to fail CV-31226 closed. This action will fail all components powered
from PNL 191 to their loss of power position, which will not adversely affect
safe shutdown.
VFDR-013-1-02 VFDR-018-1-02
RA
013, 018
U1 Appendix R Switching Equipment
cabinet
Storage for Electrical Safety Personal Protective
Equipment (PPE) and Equipment
Obtain switching protective equipment and the 4 ft. hot stick from the Bus
13/14 Room Appendix R cabinet. Don the switching protective equipment prior to locally operating circuit breakers at
4kV switchgear.
VFDR-013-1-02 VFDR-018-1-02
RA
Northern States Power Company Attachment G – Recovery Actions Transition
1LI-433C Pressurizer Level Cold Calibration Local Indicator
LOOP 1L-433 (local indicator 1LI-433C) remains available in Fire Area 058 to provide pressurizer level indication for local control of charging pump flow.
VFDR-013-1-05 VFDR-018-1-05
RA
013, 018
034-011 D1 Diesel Generator Stop D1 DSL GEN if running with inadequate cooling water pressure.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear Prior to starting the D1 Diesel, at BUS 15 in Fire Area 081, verify BKR 15-3 is
open.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear Prior to starting the D1 Diesel, at BUS 15 in Fire Area 081, verify BKR 15-7 is
open.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear Verify D1 Diesel running and ensure all loads are stripped from BUS 15.
Manually close BKR 15-2 (BUS 15 SOURCE FROM D1 DSL GEN) at BUS
15 in Fire Area 081 by pulling the manual CLOSURE lever with the hot
stick.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear On the front panel of BKR 15-11, BUS 15 FEED TO 111M XFMR, place the
LOCAL/REMOTE switch in the "LOCAL" position.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear Verify D1 Diesel running and manually close BKR 15-11 at BUS 15 in Fire
Area 081.
VFDR-013-1-06 VFDR-018-1-06
RA
013, 018
BUS 15 BUS 15 4.16KV Switchgear On the front panel of BKR 15-6, BUS 15 FEED TO 112M XFMR, place the
LOCAL/REMOTE switch in the "LOCAL" position.
VFDR-013-1-06 VFDR-018-1-06
RA
Northern States Power Company Attachment G – Recovery Actions Transition
Obtain switching protective equipment and the 4 ft. hot stick from the Appendix
R Switching Equipment cabinet (in hallway outside 25 Bus Room). Don the switching protective equipment prior to locally operating circuit breakers at 4kV
switchgear.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
BUS 26 BUS 26 4.16KV Switchgear At BUS 26 in Fire Area 118, open DC Knife switches (located inside breaker cubicles) for BKR 26-9 (22 CS PMP)
and verify breaker open.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
CV-31211 Charging Line To 21 Regenerative Heat
Exchanger Control Valve
Prior to starting MTR 211J-1, operate valves in Fire Area 085 to isolate and vent the air supply to fail CV-31211 in
the required open position.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MCC 2K1 Motor Control Center 2K BUS 1
De-energize MV-32178 at MCC 2K1, BKR 211J-8, located in Fire Area 073.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MCC 2KA2 Motor Control Center 2KA BUS 2
De-energize MV-32179 at MCC 2KA2, BKR 221B-28, located in Fire
Area 073.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MCC 2A1 Motor Control Center 2A BUS 1
De-energize MV-32180 at MCC 2A1, BKR 211E-9, located in Fire Area 031.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MCC 2A2 Motor Control Center 2A BUS 2
De-energize MV-32181 at MCC 2A2, BKR 221E-9, located in Fire Area 031.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MCC 2K2 Motor Control Center 2K BUS 2
Manually trip MTR 221J-1 (21 CHG PMP) at MCC 2K2, BKR 221J-1, in Fire
Area 073.
VFDR-013-2-02 VFDR-018-2-02
RA
Northern States Power Company Attachment G – Recovery Actions Transition
Manually operate MTR 211J-1 (22 CHARGING PUMP) at the local panel in Fire Area 073 by placing the LOC/REM
SEL switch, CS-7082001 in the “LOCAL” position. Then ENERGIZE the
VFD by momentarily depressing the START pushbutton, CS-7082002. Then START the 22 CHARGING PUMP by momentarily depressing CS-7082002.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
MTR 211J-1 VFD
22 Charging Pump Variable Frequency Drive Cabinet
Switch 22 charging pump local/remote switch to the "LOCAL" position at 22 Charging pump VFD cabinet next to
MCC 2K1 in Fire Area 073.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
PNL 21 DC Distribution Panel 21 De-energize PNL 25 at PNL 21, breaker 21-16, located in Fire Area 035 in order to fail CV-31230 closed. This action will fail all components powered from PNL 25 to their loss of power position, which will not adversely affect safe shutdown.
VFDR-013-2-02 VFDR-018-2-02
RA
013, 018
PNL 22 DC Distribution Panel 22 De-energize PNL 26 at PNL 22, breaker 22-18, located in Fire Area 036 in order to fail "B" Train valves downstream of MSIVs closed. This action will fail all components powered from PNL 26 to their loss of power position, which will
When the AFW pump suction pressure reaches 3 psig (PI-11081), verify MV-32030 (22 TD AFW PMP SUCT CL
SPLY MV) open.
VFDR-013-2-03 VFDR-018-2-03
RA
013, 018
2LI-487A 21 SG water level Monitor 21 SG water level at the Train A Hot Shutdown Panel.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
2NI-51C Unit 2 Source range monitoring
Monitor source range neutron flux at the Train A Hot Shutdown Panel.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
2LI-433B Unit 2 Pressurizer Level Monitor Pressurizer level at the Train A Hot Shutdown Panel.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
2PI-709A Unit 2 RCS Pressure Monitor RCS pressure at the Train A Hot Shutdown Panel.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
2TI-450A 2TI-450B
Unit 2 RCS Hot Leg temperature
Unit 2 RCS Cold Leg temperature
Monitor RCS hot and cold leg temperatures at the Train A Hot
Shutdown Panel.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
2LI-433C Pressurizer Level Cold Calibration Local Indicator
LOOP 2L-433 (local indicator 2LI-433C) remains available in Fire Area 073 to provide pressurizer level indication for local control of charging pump flow.
VFDR-013-2-05 VFDR-018-2-05
RA
013, 018
BUS 25 BUS 25 4.16KV Switchgear Prior to starting the D5 Diesel, at BUS 25 in Fire Area 117, verify BKR 25-13 is
open.
VFDR-013-2-06 VFDR-018-2-06
RA
013, 018
BUS 25 BUS 25 4.16KV Switchgear Prior to starting the D5 Diesel, at BUS 25 in Fire Area 117, verify BKR 25-16 is
open.
VFDR-013-2-06 VFDR-018-2-06
RA
Northern States Power Company Attachment G – Recovery Actions Transition
switching protective equipment prior to locally operating circuit breakers at 4kV
switchgear.
059 BUS 26 BUS 26 4.16KV Switchgear At BUS 26 in Fire Area 118, open DC Knife switches (located inside breaker cubicles) for BKR 26-9 (22 CS PMP)
and verify breaker open.
VFDR-059-2-16
RA
066 PNL 70384 22 Cooling Water Strainer Local Panel
Place CV-31655 in emergency at local panel 70384 in Fire AREA 041A to
provide strainer backwash for 22 CLG WTR STRNR.
VFDR-066-0-01 RA
080 PNL 70384 22 Cooling Water Strainer Local Panel
Place CV-31655 in emergency at local panel 70384 in Fire AREA 041A to
provide strainer backwash for 22 CLG WTR STRNR.
VFDR-080-0-01 RA
081 PNL 70384 22 Cooling Water Strainer Local Panel
Place CV-31655 in emergency at local panel 70384 in Fire AREA 041A to
provide strainer backwash for 22 CLG WTR STRNR.
VFDR-081-0-01 RA
097 PNL 70384 22 Cooling Water Strainer Local Panel
Place CV-31655 in emergency at local panel 70384 in Fire AREA 041A to
provide strainer backwash for 22 CLG WTR STRNR.
VFDR-097-0-01 RA
Northern States Power - Minnesota Attachment H – NEI 04-02 FAQs Summary Table
PINGP Page H-1 – Revision 1
H. NFPA 805 Frequently Asked Question Summary Table
2 Pages Attached
Note: The NFPA 805 FAQ process will continue through the transition of non-pilot NFPA 805 plants. Final closure of the FAQs will occur when RG 1.205 is revised to endorse a new revision of NEI 04-02 that incorporates the outstanding FAQs.
Northern States Power - Minnesota Attachment H – NEI 04-02 FAQs Summary Table
PINGP Page H-2 – Revision 1
This table includes the approved FAQs that have not been incorporated into the current endorsed revision of NEI 04-02 and were utilized in this submittal:
Table H-1 - NEI 04-02 FAQs Utilized in LAR Submittal
No. Rev. Title FAQ Ref. Closure Memo
06-0008 9 NFPA 805 Fire Protection Engineering Analyses
* Note: The FAQ submittal number was 08-0054 but the NRC closure memo for the FAQ was listed as 07-0054. 07-0054 was used to be consistent with the Closure Memo.
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I. Definition of Power Block
1 Page Attached
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The structures in the owner controlled area were evaluated in Prairie Island Nuclear Generating Plant (PINGP) Engineering Evaluation EC 23946, “NFPA 805 LAR Supplement Attachment I – Power Block Definition,” to determine those that are required to meet the nuclear safety performance criteria and/or the radioactive release performance criteria as described in Section 1.5 of NFPA 805.
For the purposes of establishing the structures included in the Fire Protection program in accordance with 10 CFR 50.48(c) and NFPA 805, plant structures listed in the following table are considered to be part of the power block.
* Fire Area designations for Intake Screenhouse (Fire Area 86) and Cooling Tower Transformers (Fire Area
46A) to be added in FHA revision – See Attachment S, Table S-3
Northern States Power - Minnesota Attachment J – Fire Modeling V&V
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J. Fire Modeling V&V
12 Pages Attached
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J. Fire Modeling V&V
Fire modeling tools are used in the PINGP NFPA 805 transition process in support of the Fire PRA. The fire models listed in Table J-1 were used within the Fire PRA to assess the extent of fire generated conditions for the different fire scenarios postulated
in the Fire PRA for quantification of CDF, LERF, ∆CDF and ∆LERF. Table J-1 includes the model identification, the technical references for the model, and the validation work available for it. The selected models are:
• Listed in Regulatory Guide 1.205, Rev. 1 published in December 2009, or
• Described in NUREG/CR-6850 or subsequent FAQ’s
The models identified in Table J-1 are shown to have been appropriately applied within the range of their applicability and V&V as described in the following section.
J.1 Verification and Validation
Section 2.4.1.2.3 in NFPA 805 states that fire models “shall be verified and validated.” The fire modeling analysis in support of the PINGP Fire PRA has been subjected to the Verification and Validation (V&V) process documented in NUREG-1824.
J.2 Model Application Range
The V&V study documented in NUREG-1824 and NUREG-1934 specify a range of applicability for the validation results. This range of applicability is expressed in terms of dimensionless parameters. The range of model input parameters from the validation study are expressed in dimensionless terms so that fire modeling analysts can compare them with plant specific scenarios of different scales. J.2.1 Engineering Calculations (i.e. Hand Calculations)
For the use of engineering calculations (i.e., hand calculations used for screening and/or determining near-field fire generated conditions), Attachment O of the Single Compartment Analysis Notebook, FPRA-PI-SCA, provides an overview of the V&V considerations in the Fire PRA. Specifically, this attachment to FPRA-PI-SCA evaluates the dimensionless parameters for a range of typically used input values in the Fire PRA. The objective is to provide an evaluation that demonstrates the models are appropriate for the practical range of inputs used for these engineering calculations. The engineering calculations were used for the most part in determining:
• The heat release rate required for generating damage to targets located in the fire plume or exposed to flame radiation. This is the fire intensity required to generate fire conditions exceeding the generic damage criteria specified in Appendix H of NUREG/CR-6850 as applicable to the targets for the scenarios defined in the PINGP Fire PRA,
• The temperature of the fire plume at a given location above the fire, and
Northern States Power - Minnesota Attachment J – Fire Modeling V&V
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• The incident heat flux to targets horizontally aligned with the fire source. The V&V process provides reasonable assurance that the models are used within the validation range and over-predict fire generated conditions or produce conservative (i.e. bounding) results. J.2.2 Zone Modeling (CFAST 6.1.1.54)
The zone model CFAST was used for determining hot gas layer temperatures for selected fire zones within the scope of the Fire PRA. These calculations are documented in the Multi-Compartment Analysis Notebook (FPRA-PI-MCA), in the Relay Room Analysis (FPRA-PI-RRA) and in the Turbine Building Analysis (FPRA-PI-TBA). The dimensionless parameters for the CFAST files were evaluated against the available V&V criteria in NUREG-1824. It should be noted that in some calculations, particularly those associated with relatively large fire zones, there are relatively complex configurations not explicitly covered by the V&V criteria in NUREG-1824. However, these calculations are necessary in order to ensure that the contribution from fire scenarios involving damaging hot gas layer conditions in the fire zones are not excluded from the fire risk model. In order to address in part the scope limitations of the V&V study, all hot gas layer calculations were conducted using heat release rate values consisting of the combination of fixed ignition source and intervening combustibles that bound all the potential scenarios in the fire zone. In addition, sensitivity analyses were conducted assuming that natural ventilation, forced ventilation, and room dimensions fall within the validation ranges suggested in NUREG 1824. Although not explicitly compared with the available V&V ranges, the fire modeling approach/results ensure that the CFAST predictions are consistent with the V&V guidance. J.2.3 Field Modeling (FDS Version 5.5.3)
The field model FDS was used for determining main control room abandonment conditions. These calculations are documented in the Main Control Room Analysis Notebook (FPRA-PI-MCR). The dimensionless parameters for the FDS files were evaluated against the available V&V criteria in NUREG-1824.
J.3 Summary of Fire Modeling Verification and Validation
Table J-1 provides a detailed listing of the fire models used in support of the PINGP Fire PRA, applicable technical references and V&V information. The table is organized as follows:
• The “Calculation” column identifies the specific fire model used in support of the PINGP Fire PRA.
• The “Application” column describes the type of calculation done in Fire PRA with the fire model.
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• The “V&V Basis” column lists the references where V&V information for the fire modeling is found and the V&V was conducted for the fire model in the PINGP Fire PRA.
• The “Discussion” column lists references documenting the technical basis for the fire model and provides additional technical information.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Flame Height
(Method of Heskestad)
Calculates the vertical extension of the flame
region of a fire.
• NUREG-1805, Chapter 3, 2004
• NUREG-1824, Volume 3, 2007
• NUREG-1934, 2012
• Society of Fire Protection Engineers (SFPE) Handbook, 4th Edition, Chapter 2-1, Heskestad, 2008
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used consistent with the verification and validation available as documented in FPRA-PI-SCA.
• NUREG/CR-6850, Appendix H - Damage Criteria, 2005
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-SCA or it has been justified that bounding conservative results are produced.
• NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Radiant Heat Flux
(Point Source Method)
Calculates the horizontal separation distance,
based on heat flux, to a target in order to
determine the horizontal extent of the ZOI or severity factor HRR.
• NUREG/CR-6850, Appendix H - Damage Criteria, 2005
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-SCA or it has been justified that bounding conservative results are produced.
• NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
Hot Gas Layer
(Method of MQH)
Calculates the hot gas layer temperature for a
room with natural ventilation.
• NUREG-1805, Chapter 2, 2004
• NUREG-1824, Volume 3, 2007
• NUREG-1934, 2012
• SFPE Handbook, 4th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-SCA or it has been justified that bounding conservative results are produced.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Hot Gas Layer
(Method of Beyler)
Calculates the hot gas layer temperature for a
closed compartment with no ventilation.
• NUREG-1805, Chapter 2, 2004
• NUREG-1824, Volume 3, 2007
• NUREG-1934, 2012
• SFPE Handbook, 4th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-SCA or it has been justified that bounding conservative results are produced.
Hot Gas Layer
(Method of Foote, Pagni, and Alvares
[FPA])
Calculates the hot gas layer temperature for a
room with forced ventilation.
• NUREG-1805, Chapter 2, 2004
• NUREG-1824, Volume 3, 2007
• NUREG-1934, 2012
• SFPE Handbook, 4th Edition, Chapter 3-6, Walton W. and Thomas, P., 2008
• The correlation is used in the NUREG-1805 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-SCA or it has been justified that bounding conservative results are produced.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Ceiling Jet Temperature
(Method of Alpert)
Calculates the horizontal separation distance,
based on temperature at the ceiling of a room, to
a target in order to determine the horizontal
extent of the ZOI.
• FIVE-Rev1, Referenced by EPRI Report 1002981, 2002
• NUREG-1824, Volume 4, 2007
• NUREG-1934, 2012
• SFPE Handbook, 4th Edition, Chapter 2-2, Alpert, R., 2008
• NUREG/CR-6850, Appendix H - Damage Criteria, 2005
• The correlation is used in the FIVE-Rev1 fire model, for which V&V was documented in NUREG-1824.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The verification and validation demonstrates that the ceiling jet correlation is implemented correctly and in all cases provides conservative bounding estimates.
• NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Hot Gas Layer Calculations using
Fire Dynamics Simulator (Version 5)
Evaluates the time at which control room
abandonment is necessary based on
smoke obscuration and average HGL temperature.
• FDS Version 5.5.3
• NIST Special Publication 1018-5, Volume 2: Verification
• NIST Special Publication 1018-5, Volume 3: Validation
• NUREG-1824, Volume 7, 2007
• NUREG-1934, 2012
• V&V of the FDS is documented in NIST Special Publication 1018-5.
• The V&V of FDS specifically for Nuclear Power Plant applications has also been documented in NUREG-1824.
• It was concluded that FDS models the radiant heat and gas temperature in an appropriate manner. Furthermore, the predictions of radiant heat and temperature are deemed to be within the bounds of experimental uncertainty.
• NUREG/CR-6850 generic screening damage criteria is used, which is considered conservative or it has been justified that bounding conservative results are produced.
• The correlation is used within the limits of its range of applicability as documented in FPRA-PI-MCR. For relevant scenarios where the input parameters are outside of the limits, control room abandonment conditions are still predicted.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Hot Gas Layer Calculations using
CFAST
(Version 6)
Calculates the upper and lower layer temperatures
for various compartments, the layer
height, and smoke obscuration.
• NIST Special Publication 1086, 2008
• CFAST Version 6.1.1.54
• NUREG-1824, Volume 5, 2007
• NUREG-1934, 2012
• V&V of the CFAST code is documented in the NIST Special Publication 1086.
• The V&V of CFAST specifically for Nuclear Power Plant applications has also been documented in NUREG-1824.
• It was concluded that CFAST models the hot gas layer height, temperature and smoke concentration in an appropriate manner. Furthermore, the predictions of HGL height and temperature are deemed to be within the bounds of experimental uncertainty.
• CFAST is used within the limits of its range of applicability as documented in FPRA-PI-MCA, FPRA-PI-SCA, and FPRA-PI-RRA or it has been justified that bounding conservative results are produced.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
Corner and Wall HRR
Determines a heat release rate adjustment factor for fires that are proximate to a wall or
• The correlation is recommended by IMC 0609 for fires near walls and corners.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• Consistent with the guidance, the wall and corner effects are input parameters to the model and therefore, applicability limits as defined in NUREG 1824 do not specifically apply. The applicability limits are applicable to the hand calculations (e.g., plume temperature correlation) where the input parameters for wall and corner effects are used.
• The correlation is recommended by NUREG/CR-6850.
• The correlation is documented in an authoritative publication of the SFPE Handbook of Fire Protection Engineering.
• The implementation of the Method of Lee is consistent with the guidance in NUREG/CR-6850 and is conservatively applied so that no fire extinction due to fuel consumption is calculated during the time period where a damaging hot gas layer condition is established.
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Table J-1 - V & V Basis for Fire Models / Model Correlations Used
Calculation Application V & V Basis Discussion
FLASH-CAT Used to calculate the heat release rate from
cable tray stacks • NUREG/CR-7010
• Validation of this model is available in NUREG/CR-7010
• Verification of the use of this model is documented in FPRA-PI-RRA.
• NUREG/CR-7010 contains the V&V for the FLASH-CAT model. The implementation of the FLASH-CAT model is consistent with the guidance in NUREG/CR-7010 and is conservatively applied so that no fire extinction due to fuel consumption is calculated during the time period where a damaging hot gas layer condition is established.
Table J-1 References:
1. NUREG-1805, “Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U. S. Nuclear Regulatory
Commission Fire Protection Inspection Program,” U.S. Nuclear Regulatory Commission, Washington, DC, December 2004.
2. NUREG-1824, “Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications,” U.S. Nuclear Regulatory Commission, Washington, DC, May 2007.
3. NUREG-1934 and EPRI 1023259, Nuclear Power Plant Fire Modeling Analysis Guidelines (NPP FIRE MAG). U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (RES), Washington, DC and Electric Power Research Institute (EPRI), Palo Alto, CA: 2012.
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4. The SFPE Handbook of Fire Protection Engineering, 4th Edition, P. J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.
5. NUREG/CR-6850, “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,” U.S. Nuclear Regulatory Commission, Washington, DC, September 2005.
6. “Fire Modeling Guide for Nuclear Power Plant Applications”, EPRI 1002981, FINAL REPORT, August 2002.
7. NIST Special Publication 1018-5, “Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 2: Verification”, National Institute of Standards and Technology, October 29, 2010.
8. NIST Special Publication 1018-5, “Fire Dynamics Simulator (Version 5) Technical Reference Guide, Volume 3: Validation”, National Institute of Standards and Technology, October 29, 2010.
9. NIST Special Publication 1086, “CFAST – Consolidated Model of Fire Growth and Smoke Transport (Version 6): Software
Development and Model Evaluation Guide,” December 2008.
10. IMC 0609, Appendix F, “Fire Protection Significance Determination Process,” Issue Date February 28, 2005.
11. NUREG/CR-7010, “Cable Heat Release, Ignition, and Spread in Tray Installations during Fire (CHRISTIFIRE) Volume 1: Horizontal Trays,” Final Report, McGrattan, K., Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, July 2012.
12. FPRA-PI-SCA, “Fire Scenario Selection Single Compartment Analysis Notebook,” Revision 1, March 2014
13. FPRA-PI-MCR, “Main Control Room Analysis Notebook,” Revision 1, March 2014.
14. FPRA-PI-MCA, “Multi-Compartment Analysis Notebook,” Revision 1, March 2014.
15. FPRA-PI-RRA, “Relay Room Analysis,” Revision 1, March 2014.
16. FPRA-PI-TBA, “Turbine Building Analysis,” Revision 1, March 2014.
Northern States Power - Minnesota Attachment K – Existing Licensing Action Transition
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K. Existing Licensing Action Transition
24 Pages Attached
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Attachment K Existing Licensing Action Transition
Licensing Action: Appendix R Exemption, Control Room, Lack of automatic fixed
suppression system (III.G.3 criteria), Units 1 and 2, Fire Area 13
Basis Date: February 2, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a December 6, 1982 NSP submittal to the NRC for the lack of a fixed fire suppression system in the control room as required by Section III.G.3 of Appendix R. This exemption was approved by the NRC in a letter dated February 2, 1983, which provided the following justification:
1. Ionization smoke detectors are located throughout the control room.
2. Manual suppression is available within and just outside the control room, with good access for manual suppression.
3. The room is continuously manned and access is controlled.
4. The hot shutdown panels provide an alternate shutdown capability outside of the Control Room.
5. Ventilation system can remove smoke and byproducts.
6. Fire loading is light.
This exemption is no longer required for the Prairie Island Nuclear Generating Plant (PINGP) NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy is in accordance with Section 4.2.4, and uses a performance based approach that does not credit a fixed suppression system in the control room. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 13, Control Room
Initial Exemption Request:
December 6, 1982 Information Related to Compliance with Safe Shutdown Requirements of 10 CFR Part 50 Appendix R and Request for Exemption from Requirements of 10CFR Part 50, Appendix R, Section III.G.3
Exemption Correspondence:
None
Exemption SER: NRC Exemption, February 2, 1983
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Train "A" Hot Shutdown Panel; Instrument Air
Room and Auxiliary Feedwater Pump Room, Lack of 20' separation free of intervening combustibles or lack of a 1-hour fire barrier (III.G.2 criteria), Units 1 and 2, Fire Area 31
Basis Date: May 4, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a June 30, 1982 NSP submittal to the NRC for the lack of 20 feet of horizontal separation free of intervening combustibles or lack of a 1-hour fire barrier as required by Section III.G.2 of Appendix R. Additional information was provided in an NSP letter dated October 22, 1982.
This exemption was approved by the NRC in a letter dated May 4, 1983, which provided the following justification:
1. The area is equipped with an automatic sprinkler system.
2. Ionization smoke detectors are provided.
3. Portable fire extinguishers are provided.
4. A standpipe hose station and fire extinguishers are located immediately outside the area and can provide service to the area.
5. Thermal barriers will be installed on the top and bottom of Division B cable trays and Division B conduits will be wrapped in one-hour fire rated barriers.
6. The existing sprinkler system will be modified to provide coverage below the cable trays and piping in the areas.
7. The area has low in-situ combustible fuel loads with a fire severity of 8 minutes.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. Although the number and configuration of combustibles has changed since the approval of this exemption, the NFPA 805 transition compliance strategy uses a performance based approach in accordance with Section 4.2.4. The NFPA 805 evaluation does not credit 20 feet of horizontal separation with no intervening combustibles. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 31, “A” Train Hot Shutdown Panel & Air Compressor/Auxiliary Feedwater Room
Initial Exemption Request:
June 30, 1982 Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982 Clarification of Information provided in Support of Request for Exemption
Exemption SER: NRC SER, May 4, 1983
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Train "B" Hot Shutdown Panel; Instrument Air Room and Auxiliary Feedwater Pump Room, Lack of 20' separation free of intervening combustibles or lack of a 1-hour fire barrier (III.G.2 criteria), Units 1 and 2, Fire Area 32
Basis Date: May 4, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a June 30, 1982 NSP submittal to the NRC for the lack of 20 feet of horizontal separation free of intervening combustibles or lack of a 1-hr fire barrier as required by Section III.G.2 of Appendix R. Additional information was provided in a letter dated October 22, 1982.
This exemption was approved by the NRC in a letter dated May 4, 1983, which provided the following justification:
1. The area is equipped with an automatic sprinkler system.
2. Ionization smoke detectors are provided.
3. Portable fire extinguishers are provided.
4. A standpipe hose station and fire extinguishers are located immediately outside the area and can provide service to the area.
5. Thermal barriers will be installed on the top and bottom of Division B cable trays and Division B conduits will be wrapped in one-hour fire rated barriers.
6. The existing sprinkler system will be modified to provide coverage below the cable trays and piping in the areas.
7. The area has low in-situ combustible fuel loads with a fire severity of 12 minutes.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. Although the number and configuration of combustibles has changed since the approval of this exemption, the NFPA 805 transition compliance strategy uses a performance based approach in accordance with Section 4.2.4. The NFPA 805 evaluation does not credit 20 feet of horizontal separation with no intervening combustibles. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 32, ”B” Train Hot Shutdown Panel & Air Compressor/Auxiliary Feedwater Room
Initial Exemption Request:
June 30, 1982, Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982 Clarification of Information provided in Support of Request for Exemption
Exemption SER: NRC SER, May 4, 1983
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Operating Level, Lack of
automatic fixed suppression system (III.G.2 criteria), Unit 1, Fire Area 60
Basis Date: May 4, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a June 30, 1982, NSP submittal to NRC for the lack of automatic fixed fire suppression as required by Section III.G.2 of Appendix R. Additional information was provided in an NSP letter dated October 22, 1982.
This exemption was approved by the NRC in a letter dated May 4, 1983, which provided the following justification:
1. The area is equipped with ionization smoke detectors.
2. Portable fire extinguishers are provided.
3. Standpipe hose stations are provided.
4. Motor control centers and associated cabling entering and leaving the motor control centers are horizontally separated by 22 feet.
5. All cables are qualified to IEEE-383.
6. In-situ combustible loadings in the area are light.
7. Hazardous quantities of transient combustibles are not expected. This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. Plant modifications changed the power supplies for steam supply valves and there is no redundant equipment required for safe shutdown located in this fire area. The NFPA 805 transition compliance strategy uses a deterministic approach in accordance with Section 4.2.3 and this exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 60, Auxiliary Building, Operating Level, Unit 1
Initial Exemption Request:
June 30, 1982 Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982 Clarification of Information provided in Support of Request for Exemption from the Requirements of 10 CFR Part 50, Appendix R, Section III.G
Exemption SER: NRC SER, May 4, 1983
Associated EEEEs: None
Evaluation: Based on plant modifications, there is no redundant equipment required for safe shutdown located in this fire area and this exemption is no longer required. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Operating Level, Lack of
automatic fixed suppression system (III.G.2 criteria), Unit 2, Fire Area 75
Basis Date: May 4, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a June 30, 1982 NSP submittal to NRC for the lack of automatic fixed fire suppression as required by Section III.G.2 of Appendix R. Additional information was provided in an NSP letter dated October 22, 1982.
This exemption was approved by the NRC in a letter dated May 4, 1983, which provided the following justification:
1. The area is equipped with ionization smoke detectors.
2. Portable fire extinguishers are provided.
3. Standpipe hose stations are provided.
4. Motor control centers and associated cabling entering and leaving the motor control centers are horizontally separated by 22 feet.
5. All cables are qualified to IEEE-383.
6. In-situ combustible loadings in the area are light.
7. Hazardous quantities of transient combustibles are not expected. This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. Plant modifications changed the power supplies for steam supply valves and there is no redundant equipment required for safe shutdown located in this fire area. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 75, Auxiliary Building, Operating Level, Unit 2
Initial Exemption Request:
June 30, 1982 Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982 Clarification of Information provided in Support of Request for Exemption from the Requirements of 10 CFR Part 50, Appendix R, Section III.G
Exemption SER: NRC SER, May 4, 1983
Associated EEEEs: None
Evaluation: Based on plant modifications, there is no redundant equipment required for safe shutdown located in this fire area and this exemption is no longer required. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Normal Switchgear Room, Lack of automatic fixed suppression system (III.G.2 criteria), Unit 1, Fire Area 37
Basis Date: May 4, 1983
Transitioned? No
Licensing Basis: This exemption was requested in a June 30, 1982 NSP submittal to the NRC for the lack of an automatic fixed suppression system as required by Section III.G.2 of Appendix R. Additional information was provided in an NSP letter dated October 22, 1982.
This exemption was approved by the NRC in a letter dated May 4, 1983, which provided the following justification:
1. The area is equipped with ionization smoke detectors.
2. Portable fire extinguishers are provided.
3. Standpipe hose stations and fire extinguishers are located outside the access door and are available for servicing this fire area.
4. This fire area was described in the SER as containing emergency diesel generator power cables of redundant divisions from which one train is necessary for safe shutdown. Both division cables are routed in conduit through the area. Horizontal separation between redundant emergency diesel generator power supplies is greater than 20 feet.
NOTE: Upon further review NSPM has determined that this cable description is incorrect and should have stated the following: Diesel driven cooling water pump control cables in both divisions are routed in conduit through the area. Existing horizontal separation between redundant diesel driven cooling water pump cables is greater than 20 feet.
5. All cables are qualified to IEEE-383.
6. In-situ combustible loadings in the area are light, consisting of cable insulation that corresponds to a fire severity of approximately 9 minutes.
7. Hazardous quantities of transient combustibles are not expected. Subsequent to approval of this exemption, the safe shutdown components in Fire Area 37 that did not have 20 feet of separation were relocated. This exemption was withdrawn in a letter from NSP to the NRC dated June 9, 1986, and is no longer applicable to PINGP. It will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 37, Unit 1 480V Normal Switchgear Room
Initial Exemption Request:
June 30, 1982 Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982 Clarification of Information provided in Support of Request for Exemption from the Requirements of 10 CFR Part 50, Appendix R, Section III.G June 9, 1986 Fire Protection Safe Shutdown Analysis and Compliance with Section III.G and III.O of 10 CFR 50, Appendix R
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Exemption SER: NRC SER, May 4, 1983
Associated EEEEs: None
Evaluation: This exemption was withdrawn and is no longer applicable to PINGP. It will not be transitioned to the new licensing basis.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Ground Level, Lack of automatic fixed suppression system (III.G.2 criteria), Unit 1, Fire Area 58
Basis Date: January 9, 1984
Transitioned? No
Licensing Basis: This exemption was initially requested in a June 30, 1982 NSP submittal to the NRC for the lack of separation between redundant equipment and the lack of automatic fixed suppression as required by Section III.G.2 of Appendix R. In a letter dated May 4, 1983, the NRC determined that the level of fire protection as described was not adequate and the initial request was denied. NSP re-submitted a request for this exemption in a letter dated March 11, 1983, in which NSP explained the separation provisions for redundant equipment and requested exemption from the Section III.G.2 requirement to install an automatic fire suppression system in this area. This exemption was approved by the NRC in a letter dated January 9, 1984; the SER provided the following justification:
1. The area is equipped with a smoke detection system.
2. Adequate manual fire fighting equipment is available.
3. Division B cables and certain Division A safe shutdown cable trays will be enclosed with one-hour fire barriers.
4. Spatial separation is provided between safe shutdown equipment.
5. Fuel loading is low, consisting of cable insulation which if consumed would correspond to a fire severity of approximately 8 minutes.
6. Hazardous quantities of transient combustibles are not expected.
Modifications: The SER also described the installation of a one-hour fire rated barrier around all Division B safe shutdown cables and Division A safe shutdown cable trays in the vicinity of MCC 1K2 (Division B).
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy is in accordance with Section 4.2.4, and uses a performance based approach that does not credit a fixed fire suppression system. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 58, Auxiliary Building, Ground Level, Unit 1
Initial Exemption Request:
June 30, 1982, Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982, NSP Clarifying Information in Support of Requests for Exemption from 10 CFR Part 50, Appendix R, Section III.G January 12, 1983, NRC Draft SER on Appendix R Exemption Request; this letter was referenced in NSP letter dated March 11, 1983 as the basis for NSP’s re-submittal of the exemption request for Fire Areas No. 58, 59, 73, and 74
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March 11, 1983, NSP Request for Relief from the Requirements of 10 CFR Part 50, Section 50.48(b) for Fire Areas No. 58, 59, 73 and 74 May 4, 1983, NRC letter; this letter denied the exemption for FA 58 as originally requested in the June 30, 1982 NSP letter May 16, 1983, NSP letter Clarifying Information in Support of Exemption Requests for Fire Areas 58, 59, 73, and 74
Exemption SER: NRC SER, Enclosure 2 to the approval letter dated January 9, 1984
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Ground Level, Unit 2, Lack of automatic fixed fire suppression system (III.G.2 criteria), Fire Area 73
Basis Date: January 9, 1984
Transitioned? No
Licensing Basis: This exemption was initially requested in a June 30, 1982 NSP submittal to the NRC for the lack of separation between redundant equipment and the lack of automatic fixed suppression as required by Section III.G.2 of Appendix R. In a letter dated May 4, 1983, the NRC determined that the level of fire protection as described was not adequate and the initial request was denied. NSP re-submitted a request for this exemption in a letter dated March 11, 1983, in which NSP explained the separation provisions for redundant equipment and requested exemption from the Section III.G.2 requirement to install an automatic fire suppression system in this area. This exemption was approved by the NRC in a letter dated January 9, 1984; the SER provided the following justification:
1. The area is equipped with a smoke detection system.
2. Adequate manual fire fighting equipment is available.
3. Division B cables and certain Division A safe shutdown cable trays will be enclosed with one-hour fire barriers.
4. Spatial separation is provided between safe shutdown equipment.
5. Fuel loading is low, consisting of cable insulation which if consumed would correspond to a fire severity of approximately 6 minutes.
6. Hazardous quantities of transient combustibles are not expected.
Modifications: The SER also described installation of a one-hour fire rated barrier around all Division B safe shutdown cables and Division A safe shutdown cable trays in the vicinity of Division B MCC 2K2.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy is in accordance with Section 4.2.4, and uses a performance based approach that does not credit a fixed fire suppression system. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 73, Auxiliary Building, Ground Level, Unit 2
Initial Exemption Request:
June 30, 1982, Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982, NSP Clarifying Information in Support of Requests for Exemption from 10 CFR Part 50, Appendix R, Section III.G January 12, 1983, NRC Draft SER on Appendix R Exemption Request; this letter was referenced in NSP letter dated March 11, 1983 as the basis for NSP’s re-submittal of the exemption request for Fire Areas No. 58, 59, 73,
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and 74 March 11, 1983, NSP Request for Relief from the Requirements of 10 CFR Part 50, Section 50.48(b) for Fire Areas No. 58, 59, 73 and 74 May 4, 1983, NRC letter; this letter denied the exemption for FA 73 as originally requested in the June 30, 1982 NSP letter May 16, 1983, NSP letter Clarifying Information in Support of Exemption Requests for Fire Areas 58, 59, 73, and 74
Exemption SER: NRC SER, Enclosure 2 to the approval letter dated January 9, 1984
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Mezzanine Level, Unit 1, Lack of automatic fixed suppression (III.G.2 criteria), Fire Area 59
Basis Date: January 9, 1984
Transitioned? No
Licensing Basis: This exemption was initially requested in a June 30, 1982 NSP submittal to the NRC for the lack of separation between redundant equipment and the lack of automatic fixed suppression as required by Section III.G.2 of Appendix R. In a letter dated May 4, 1983, the NRC determined that the level of fire protection as described was not adequate and the initial request was denied. NSP re-submitted a request for this exemption in a letter dated March 11, 1983, in which NSP explained the separation provisions for redundant equipment and requested exemption from the Section III.G.2 requirement to install an automatic fire suppression system in this area. This exemption was approved by the NRC in a letter dated January 9, 1984; the SER provided the following justification:
1. The area is equipped with a smoke detection system.
2. Adequate manual fire fighting equipment is available.
3. Safe shutdown cable trays will be enclosed with one-hour fire barriers.
4. Redundant MCCs are separated by 28 feet and spatial separation is provided between redundant cables for safe shutdown equipment.
5. Fuel loading is low, consisting of cable insulation which if consumed would correspond to a fire severity of approximately 15 minutes.
6. Hazardous quantities of transient combustibles are not expected.
Modifications: The SER also described installation of a one-hour fire rated barrier around all Division B safe shutdown cables.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy is in accordance with Section 4.2.4, and uses a performance based approach that does not credit a fixed fire suppression system. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 59, Auxiliary Building, Mezzanine Level, Unit 1
Initial Exemption Request:
June 30, 1982, Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982, NSP Clarifying Information in Support of Requests for Exemption from 10 CFR Part 50, Appendix R, Section III.G January 12, 1983, NRC Draft SER on Appendix R Exemption Request; this letter was referenced in NSP letter dated March 11, 1983 as the basis for NSP’s re-submittal of the exemption request for Fire Areas No. 58, 59, 73, and 74
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March 11, 1983, NSP Request for Relief from the Requirements of 10 CFR Part 50, Section 50.48(b) for Fire Areas No. 58, 59, 73 and 74 May 4, 1983, NRC letter; this letter denied the exemption for FA 59 as originally requested in the June 30, 1982 NSP letter May 16, 1983, NSP letter Clarifying Information in Support of Exemption Requests for Fire Areas 58, 59, 73, and 74
Exemption SER: NRC SER, Enclosure 2 to the approval letter dated January 9, 1984
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Auxiliary Building, Mezzanine Level, Unit 2, Lack of automatic fixed suppression (III.G.2 criteria), Fire Area 74
Basis Date: January 9, 1984
Transitioned? No
Licensing Basis: This exemption was initially requested in a June 30, 1982 NSP submittal to the NRC for the lack of separation between redundant equipment and the lack of automatic fixed suppression as required by Section III.G.2 of Appendix R. In a letter dated May 4, 1983, the NRC determined that the level of fire protection as described was not adequate and the initial request was denied. NSP re-submitted a request for this exemption in a letter dated March 11, 1983, in which NSP explained the separation provisions for redundant equipment and requested exemption from the Section III.G.2 requirement to install an automatic fire suppression system in this area. This exemption was approved by the NRC in a letter dated January 9, 1984; the SER provided the following justification:
1. The area is equipped with a smoke detection system.
2. Adequate manual fire fighting equipment is available.
3. Safe shutdown cable trays will be enclosed with one-hour fire barriers.
4. Redundant MCCs are separated by 28 feet and spatial separation is provided between redundant cables for safe shutdown equipment.
5. Fuel loading is low, consisting of cable insulation which if consumed would correspond to a fire severity of approximately 14 minutes.
6. Hazardous quantities of transient combustibles are not expected.
Modifications: The SER also described installation of a one-hour fire rated barrier around all Division B safe shutdown cables.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy is in accordance with Section 4.2.4, and uses a performance based approach that does not credit a fixed fire suppression system. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 74, Auxiliary Building, Mezzanine Level, Unit 2
Initial Exemption Request:
June 30, 1982, Fire Protection Safe Shutdown Analysis and Compliance with Section III.G of 10 CFR Part 50, Appendix R, Including Requests for Relief
Exemption Correspondence:
October 22, 1982, NSP Clarifying Information in Support of Requests for Exemption from 10 CFR Part 50, Appendix R, Section III.G January 12, 1983, NRC Draft SER on Appendix R Exemption Request; this letter was referenced in NSP letter dated March 11, 1983 as the basis for NSP’s re-submittal of the exemption request for Fire Areas No. 58, 59, 73, and 74
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March 11, 1983, NSP Request for Relief from the Requirements of 10 CFR Part 50, Section 50.48(b) for Fire Areas No. 58, 59, 73 and 74 May 4, 1983, NRC letter; this letter denied the exemption for FA 74 as originally requested in the June 30, 1982 NSP letter May 16, 1983, NSP letter Clarifying Information in Support of Exemption Requests for Fire Areas 58, 59, 73, and 74
Exemption SER: NRC SER, January 9, 1984
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, Unit 1 and Unit 2 Containments, Intervening combustibles between redundant shutdown divisions, Fire Areas 1 and 71
Basis Date: July 31, 1984
Transitioned? No
Licensing Basis: This exemption was requested in a January 23, 1984 NSP submittal to the NRC based on the lack of 20 feet horizontal separation between cables and equipment of redundant trains with no intervening combustibles, as required by Section III.G.2(d) of Appendix R. With one exception as described in the submittal, these areas have the required spatial separation between redundant components needed for safe shutdown, but there are intervening combustibles. This exemption was approved by NRC in a letter dated July 31, 1984 with the following justification:
1. All redundant components are separated by 20 feet or more, except for the pressurizer level transmitters for Unit 2. As will be discussed later, this exception actually applies to both Units 1 and 2.
2. Redundant cabling associated with the pressurizer level transmitters is separated by 10 feet and one division will be protected with a one-hour fire barrier. As will be discussed later, the protection of this cabling is different for Units 1 and 2.
3. Combustibles in these fire areas are RCP lubricating oil and cable insulation. The amount of cable insulation, if consumed, would correspond to a fire severity of 7.5 minutes in Unit 1 and 7.7 minutes in Unit 2.
4. Ionization smoke detectors are located on each level with alarms in the Control Room.
5. Standpipe hose stations are located on each level. 6. Access is restricted during power operation due to high radiation
fields. 7. RCP lubricating oil would drain to the sump (approved exemption
from Section III.O). 8. The cable is IEEE-383 qualified. 9. Hazardous quantities of transient combustible materials are not
expected due to access restrictions in containment during power operation.
10. Administrative controls during use of transient combustibles require a dedicated fire watch armed with a fire extinguisher.
During the review of this exemption, it was discovered that the protection of cabling for the pressurizer level transmitters is different for Units 1 and 2, as follows:
• NSP letter dated April 5, 1984 identified the lack of 20’ separation for this cabling and stated that cables of one division would be wrapped in an approved one hour barrier for both Units 1 and 2.
• The NRC SER dated July 31, 1984 approved the exemption request and cited NSP actions to protect cabling of one division in a one-hour barrier, but this SER only identified this action for Unit 2.
• NSP installed a one-hour barrier in Unit 2 in accordance with the SER.
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• During a subsequent review in 1997 NSP recognized that protection of the pressurizer level transmitter cabling was required in Unit 1 and installed a noncombustible radiant energy shield. The previously unprotected condition of this cabling and the installation of a noncombustible radiant energy shield was described in LER 1-97-017, “Separation of Pressurizer level Indication Channels Not in Compliance with 10 CFR 50 Appendix R Section III.G.2,” submitted to the NRC on January 2, 1998.
• PINGP LER 1-97-017 was reviewed by the NRC in Inspection Report 50-28250-306-97023 dated January 30, 1998, and was closed in NRC Inspection Report 50-282/50-306-98016 dated October 9, 1998.
• The existing radiant energy shield in Unit 1 has been determined to be acceptable in accordance with the NFPA 805 performance based approach.
This exemption is no longer required for the PINGP NFPA 805 Fire Protection Program. The NFPA 805 transition compliance strategy for these Fire Areas uses a performance based approach in accordance with Section 4.2.4. Although the installation of fire barriers differs from that described in NSP’s exemption request, the pressurizer level transmitter cable protection features for both Unit 1 and Unit 2 were determined to be acceptable in accordance with the NFPA 805 performance based approach and this exemption is no longer required. This exemption will not be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 1, 71, Containment, Units 1 and 2
Initial Exemption Request:
January 23, 1984, NSP letter, Exemption Requests to the Requirements of Appendix R to 10 CFR 50
Exemption Correspondence:
April 5, 1984, Information in Support of Exemption Requests Submitted January 23, 1984 and Request for Exemption from the Requirements of Section III.O of Appendix R to 10 CFR Part 50 January 2, 1988, NSP letter, LER 1-97-17, Separation of Pressurizer Level Indication Channels Not in Compliance with 10 CFR 50 Appendix R Section III.G.2 January 30, 1998, NRC letter, Notice of Violation and NRC Inspection Report No. 50-282/97023(DRP), 50-306/97023(DRP) for PINGP October 9, 1998, NRC letter, NRC Fire Protection Functional Inspection (FPFI) Reports 50-282/98016(DRS); 50-306/98016(DRS)
Exemption SER: July 31, 1984, NRC Exemption In the Exemption attached to the approval letter the NRC states in part the following: "The licensee requested an exemption from Subsection III.G.2 to the extent that these areas have intervening combustibles between components of redundant trains needed for safe shutdown. In addition, except for the redundant cabling associated with the pressurizer level transmitters for unit 2 (Fire Area 71), all other redundant components are separated by twenty feet or more. The redundant cabling associated with the pressurizer level transmitters is separated by ten feet. For this cabling the licensee has
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committed to protecting one division with a one-hour fire barrier.” [Note: as discussed in the Licensing Basis discussion above, the SER only identifies pressurizer level transmitters in Unit 2, although the NSP submittal identified that this exemption applies to both Units 1 and 2]
Associated EEEEs: None
Evaluation: The NFPA 805 transition compliance strategy for this Fire Area uses a performance based approach and this exemption is no longer required under the new licensing basis. Also, the installation of fire barriers to protect cables for one of the redundant divisions of the pressurizer level transmitters has been evaluated using the performance based approach and, while the installations are different for Unit 1 and Unit 2, both installations have been determined to be acceptable. This exemption will not be transitioned.
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Licensing Action: Appendix R Exemption, RCP Oil Collection, RCP oil collection system is not in strict compliance (III.O criteria), Fire Areas 1, 71
Basis Date: July 31, 1984
Transitioned? Yes
Licensing Basis: This exemption was requested in an April 5, 1984 NSP submittal to the NRC for the lack of a closed vented container inside containment capable of holding the entire inventory of the Reactor Coolant Pump (RCP) lube oil collection system, as required by Section III.O of Appendix R. This exemption was approved by NRC in a letter dated July 31, 1984, with the following justification:
1. The sump is a concrete pit in the basement of containment with a capacity of 990 gallons, which is more than the capacity needed to contain the total inventory of lube oil for the two RCPs for each unit.
2. There is no safe shutdown equipment in the area surrounding the RCPs or Sump A.
3. The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the control room if this level is exceeded.
4. The sump can also manually be pumped down at any time. 5. The sump is normally drained to vented containers in the auxiliary
building which have a total capacity of 2600 gallons. This system is designed to collect contaminated water from pump seal leakage as well as oil leakage.
6. The pipe from the sump to the vented container in the auxiliary building has been designed to seismic category III which meets the requirement of Regulatory Guide 1.29, paragraph C-2.
The bases for this exemption, as approved by the NRC, reflect the current plant configuration and remain valid. This exemption will be transitioned into the NFPA 805 licensing basis.
Applicable Fire Area: 1, 71, Containment, Units 1 and 2
Initial Exemption Request:
April 5, 1984, Information in Support of Exemption Requests Submitted January 23, 1984 and Request for Exemption from the Requirements of Section III.O of Appendix R to 10 CFR Part 50 NSP described the RCP lube oil collection system and requested exemption from the specific requirement of Section III.O to have a closed vented container inside containment. NSP stated its belief that the existing collection system meets the intent of Appendix R, as previously described in a letter dated January 23, 1984, in that all lube oil is collected to a common point which will prevent its contact with hot piping in the area and is isolated from electrical power cable which might cause ignition. NSP provided the following justification for this exemption:
1. The lube oil collection system is seismically designed. 2. High flash point Mobile Synthetic lube oil is being utilized. 3. There is no safety related equipment in the vicinity of the RCP or
Sump A inside containment. 4. Safe shutdown equipment and cabling is separated from the RCPs
and Sump A by the shield wall and 18” thick concrete floors.
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5. During design and installation (1978) of the collection system a decision was made to not install a vented closed collection tank inside containment because the collection system also collects RCP water seal leakage, and it was determined that it would be better to deliver all leakage from both sources directly to the sump, where it would then be pumped to a vented closed tank.
This letter also provided the following description of the collection system: PINGP Units 1 and, 2 have two Reactor Coolant Pumps each. For purposes of this description the units are identical. Each Reactor Coolant Pump contains 265 gallons of lube oil for a total of 530 gallons per unit. The lube oil is Mobile Synthetic lube oil which has a flash point of 480°F and an ignition point of 520°F. A series of drip pans and deflectors are located around the pump such that leakage from all potential pressurized and unpressurized leakage sites in the Reactor Coolant Pump lube oil systems are collected and piped to the adjacent floor drain which empties into Sump A in the basement of the containment. Sump A is a concrete open pit, covered with grating, built into the floor which has a capacity of 990 gallons. There is no Safe Shutdown Equipment in the area surrounding the Reactor Coolant Pumps or Sump A. Sump A is designed to automatically pump down when the level of the tank reaches the 695'-9" elevation. (The bottom of" the sump is at 693' -6".) This is at approximately the 555 gallon point. If level continued to rise due to failure of the automatic pump function, an alarm would sound in the Control room at the 696'-9" level of the sump, approximately 800 gallons. An operator can then initiate manual control of the sump pump for pumping down. The top of the sump pit is at floor level, the 697 '-6" elevation which represents the 990 gallon maximum capacity point of the sump. In addition to the automatic function, operators may at any level manually control the pump to pump down the sump. The sump is normally lined up to pump to the aerated sump tank in the Auxiliary Building which has a capacity of 600 gallons. The aerated sump tank is a vented closed tank. The aerated sump tank then pumps to the aerated drain tanks in the Auxiliary Building. Each aerated drain tank has a capacity of 1000 gallons for a total capacity of 2000 gallons. The aerated sump tank and drain tanks serve both units. The aerated drain tanks are vented closed tanks. The capability also exists to pump from the aerated sump tank to the 25,000 gallon waste hold-up tank which is also a vented closed tank.
Exemption Correspondence:
January 23, 1984, NSP letter, Exemption Requests to the Requirements of Appendix R to 10 CFR 50. This letter described the RCP lube oil collection and stated its understanding that the collection system meets the intent and is in compliance with Section III.O of Appendix R to 10 CFR 50. May 22, 1984 NSP letter, Information in Support of the Request for Exemption from the Requirement of Section III.O of Appendix R to 10 CFR 50 dated April 5, 1984 This letter described that piping from the sump to the vented container in the auxiliary building is designed to either Seismic Category III or Category I.
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Exemption SER: NRC Exemption, July 31, 1984 In the exemption attachment to the approval letter the NRC states: "The licensee requested an exemption from Subsection III.O to the extent that the reactor coolant pump lube oil collection system is piped to the sump inside containment. The contents of the sump can be pumped to a closed vented container located in the auxiliary building. The licensee states that the sump in the basement of the containment is a concrete pit having a capacity of 990 gallons, which is more than the capacity needed to contain the total inventory of lube oil for the two reactor coolant pumps for each unit. There is no safe shutdown equipment in the area. The sump is designed to automatically pump down at a prescribed sump level and an alarm will sound in the control room if this level is exceeded. The operator can initiate manual control of the sump pump at any time, overriding the automatic control of sump level. The sump is normally drained to vented containers in the auxiliary building having a total capacity of 2600 gallons. The basis for the design of this collection system is to collect any contaminated water from the pump seal leakage as well as any oil leakage. In addition, the pipe from the sump to the vented container in the auxiliary building has been designed to seismic category Class III which meets the requirement of Regulatory Guide 1.29, paragraph C-2. If failure of this pipe were to occur during a seismic event, the functions of plant features described in paragraph 1 (a through q) of Regulatory Guide 1.29 will not be affected and the plant can be brought to cold shutdown. This is based on a review conducted by the licensee and confirmed by letter dated May 22, 1984. We agree with the licensee that, although lube oil leakage is collected in the sump before it is pumped to a vented container, the sump design at this plant assures us that oil collected there will not lead to fire during normal or design basis accident conditions. The capacity of the sump and the vented containers is adequate to safely contain any anticipated lube oil leakage and the existing controls provide reasonable assurance that any lube oil collected in the sump can be safely pumped to the vented container in the auxiliary building. Based on our evaluation, the existing lube oil collection system for reactor coolant pumps provides a level of protection equivalent to the requirements specified in Subsection III.O of Appendix R. Therefore, the exemption from the requirements specified in Subsection III.O for the lube oil collection system is granted."
Associated EEEEs: None
Evaluation: The bases for this exemption remain valid. This exemption will be transitioned and will be included in the new licensing basis.
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Licensing Action: Appendix R Exemption, Control Room, Use of repair to remove fuses (III.G.1 criteria), Fire Area 13
Basis Date: February 21, 1995
Transitioned? Yes
Licensing Basis: This exemption was requested in a May 2, 1994 NSP submittal to the NRC to allow manual removal of fuses from the power-operated relief valve (PORV) control circuit in the event of a fire (considered a repair action outside the bounds of Section III.G.1), in lieu of modifying plant hardware which would otherwise be required to achieve compliance with Section III.G.1 of Appendix R. The NRC approved this exemption in a letter dated February 21, 1995 and provided the following justification:
1. Closing the block valves and pulling the PORV control circuit fuses is an effective means of preventing potential loss of RCS inventory in the event of a control room fire that could result in a hot short or short to ground that may cause the PORV to open or be maintained open.
2. The requirement to remove/pull the PORV fuses is included in plant procedures as an immediate action in response to a control room evacuation.
3. The fuse panels are readily accessible and the fuses are clearly identified in the panels.
4. Sufficient space is available to permit access for pulling fuses and emergency lights and fuse pullers are provided in the vicinity of each panel.
5. The operators are trained for a control room evacuation and to remove these fuses.
Action to isolate power to the PORV control circuits in the event of a fire is still a required action and this exemption will be transitioned into the NFPA 805 licensing basis. Subsequent to the approval of this exemption, a modification was installed to allow power to be isolated from the PORV control circuits by opening disconnect switches in lieu of pulling fuses. The basis for the continued applicability of this exemption is clarified in Attachment T.
Applicable Fire Area: 13, Control Room
Initial Exemption Request:
May 2, 1994 NSP letter, Exemption Request, Use of Hot Shutdown Repair to Meet the Requirements of Appendix R NSP requested an exemption from the specific criteria in Section III.G.1 of Appendix R to allow removal of the fuses in the PORV control circuit as a means of ensuring that proper RCS inventory is maintained in the event of a fire that causes an open PORV resulting in the loss of RCS inventory. The following justification is provided:
1. Three separate failures are required to cause this event: - The operator fails to shut the PORV Block valve and/or a
material failure occurs (i.e., a short prevents the valve from closing or causes the valve to open).
- A "hot short" in the control circuitry causes the PORV to open.
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- The PORV circuit failure is sustained long enough for "the open PORV to jeopardize plant safety.
2. A Probabilistic Risk Assessment demonstrated that the event of a PORV hot short due to a control room fire represents a small contribution to overall risk of core damage.
3. Operator actions required to manually remove the fuses in the PORV control circuit are straightforward and are included in plant procedures and training.
Exemption Correspondence:
None
Exemption SER: February 21, 1995, NRC Exemption In the exemption attached to the approval letter the NRC states: “By letter dated May 2, 1994, the licensee requested an exemption to permit it to manually remove fuses from the power-operated relief valve control circuit in the event of a fire, in lieu of modifying plant hardware which would otherwise be required to achieve compliance with Section III.G.1 of Appendix R. The licensee's submittal initially referenced Section III.G.2 of Appendix R as providing the requirements from which the licensee was seeking an exemption, but in a follow-up telephone conversation with the staff the licensee concurred that Section III.G.1 is the appropriate reference. This exemption was requested by the licensee in response to inspection findings identified in inspection reports 50-282/87-004, 50-282/88-013, 50-282/92-011 and 50-282/94-004. These findings addressed a concern with circuit failure modes that could adversely affect the ability to maintain hot shutdown in the event of a control room fire. This condition could occur if the power operated relief valves (PORV) block valves were not shut and a hot short damaged the PORV control circuit causing the PORV to open and remain open. Specifically, this involves the high/low pressure interface spurious signal concerns associated with Unit 1 PORVs CV-31231 and CV-31232 and their associated block valves MOV-32195 and MOV-32196 and with Unit 2 PORVs CV-31233 and CV-31234 and their associated block valves, MOV-32197 and MOV-32198. As a precaution to prevent the potential loss of reactor coolant system (RCS) inventory during a control room fire, the licensee has proposed to close the PORV block valves prior to control room evacuation. The licensee also proposed to remove the PORV control circuit fuses to prevent a hot short or short to ground which may cause the PORV to open or be maintained open. As stated above, removal of fuses for isolation in such circumstances is considered a repair and, therefore, does not meet Appendix R, Section III.G.1, as interpreted by the staff. The licensee's proposed actions of closing the PORV block valves and removing the control circuit fuses was reviewed by the staff and was found to be an effective means of assuring that a control room fire will not result in a sustained loss of RCS inventory. The substance of the licensee's submittal was reviewed by Region III inspectors during the inspection conducted from July 18-22, 1988. The inspection findings were documented in NRC Inspection Report No. 50-282/88-013 and 50-306/88-013. The inspectors walked down the control room evacuation shutdown procedures. Step 3.3.1 of Procedure F5,
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Appendix B, "Control Room Evacuation (Fire)," directs the operators to remove/pull the fuses for the PORVs as an immediate action in response to a control room evacuation. The inspectors found that the fuse panels were readily accessible and the fuses were clearly identified in the panels. The inspectors also found that sufficient space is available to permit access for pulling fuses and that emergency lights and the fuse pullers had been provided in the vicinity of each panel. A training program has been established for all plant operators to enhance the familiarity with and proper response to the control room evacuation. Additionally, as a part of Emergency Operating Procedures (EOP) training, all the operators are trained on the above-mentioned procedures to ensure their familiarity with respect to the removal of fuses during hot shutdown. Therefore, operators are trained and experienced in removing the fuses. On the basis of this evaluation, the Commission concludes that the proposed action to close the PORV block valves prior to control room evacuation and to remove fuses from the PORV control circuit provides reasonable assurance that safe shutdown can be achieved in the event of a control room fire and is acceptable.”
Associated EEEEs: None
Evaluation: The bases for this exemption remain valid. This exemption will be transitioned and will be included in the new licensing basis, subject to clarification in Attachment T.
Northern States Power – Minnesota Attachment L NFPA 805 Chapter 3 Requirements for Approval
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L. NFPA 805 Chapter 3 Requirements for Approval (10 CFR 50.48(c)(2)(vii))
3 Pages Attached
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Approval Request 1
NFPA 805 Section 3.5.16
NFPA 805 Section 3.5.16 states:
“The fire protection water supply system shall be dedicated for fire protection use only.
Exception No. 1: Fire protection water supply systems shall be permitted to be used to provide backup to nuclear safety systems, provided the fire protection water supply systems are designed and maintained to deliver the combined fire and nuclear safety flow demands for the duration specified by the applicable analysis.
Exception No. 2: Fire protection water storage can be provided by plant systems serving other functions, provided the storage has a dedicated capacity capable of providing the maximum fire protection demand for the specified duration as determined in this section.”
Basis for Request:
NRC approval is being requested since the fire water system can be aligned for screenwash system use and emergency uses, and as such it does not meet the requirement or allowed exceptions.
Acceptance Criteria Evaluation:
The Mississippi river provides fire protection water. The system consists of two horizontal centrifugal fire pumps each rated at 2,000 gpm at 125 psig. One pump is motor driven (MDFP) and the other pump is diesel driven (DDFP). The 10” fire header is maintained between 108 and 113 psig by a jockey pump. The motor driven fire pump will automatically start at 95 psig. If the header pressure drops to 90 psig, the diesel-driven fire pump will start. The motor and diesel-driven fire pumps are designed to pump 2,000 gpm at a discharge pressure of 125 psi.
Contrary to the requirements of NFPA 805 Section 3.5.16, the fire protection water supply system at PINGP may periodically be utilized to supply water for non-fire protection purposes (screenwash).
The motor-driven fire pump can be aligned to provide a backup water supply to the Screenhouse screenwash system in the event of a screenwash pump failure.
If the fire pump is required to supplement the screenwash header flow, the pump must be started manually either locally or remotely. Control valve FP-30-10 ties the fire protection water system to the screenwash header. FP-30-10 is normally closed and requires a manual action to open it. Once the pump is operating and no auto start signal exists from the fire protection header, the discharge to the screenwash header valve (CV 31131) opens automatically via Solenoid Valve SV 33049 and maintains the screenwash header pressure at approximately 90 psig.
The non-fire protection use of the PINGP fire protection water system requires prior notification of the Control Room. This process ensures that the fire water system will be restored to full capacity during a fire scenario. Personnel utilizing fire protection water
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for non-fire protection purposes are in contact with the Control Room, therefore ensuring the ability to secure the full fire water system capacity should a fire occur.
There are two potential impacts to the use of the fire protection water system for screenwash purposes – one is a fire within the Screenhouse Fire Area 41, and one for all other fire areas, as delineated below:
For a fire in any fire area except Fire Area 41
The plant P&ID drawing indicates that SV 33049 will only open Control Valve 31131 if the 121 MDFP is not running and the pressure demand on the fire header is not below 90 psi. If there is not a demand on the fire header and the pump is manually started, the control valve will open and allow fire protection water into the Screenwash system to clean the screens. If a demand is placed on the fire header from a suppression system actuation, the pressure drop will cause SV 33049 to close CV 31131 thereby realigning the water flow to the fire protection header.
The control cables for SV 33049/CV 31131 run from the MDFP room FA 41B (elev. 670’ screenhouse) to FA 41, (elev. 695’ screenhouse).
A fire event in any fire area other than FA 41 will cause a pressure drop on the fire header thereby closing CV 31131 via SV 33049 and re-aligning the MDFP water supply from the traveling screens to the fire header.
For a fire in Fire Area 41
The control cables for SV 33049/CV 31131 run from the MDFP room FA 41B (elev. 670’ screenhouse) to FA 41, (elev. 695’ screenhouse). A postulated fire in FA 41 may cause cable damage, thereby creating the potential for a hot short of the circuit resulting in CV 31131 failing in the open position. This would divert some fire protection water from the MDFP from entering the fire header. If this scenario occurs, the DDFP will start, providing the fire header with 2,000 gpm at 125 psi. FA 41 suppression system, PA-9, has a demand of 1,094.1 gpm at 89.1 psi. This demand is within the design capacity of the DDFP. Check valve FP-28-2 will prevent water in the fire protection header from entering the Screenwash diversion piping network.
The use of the MDFP for Screenwash cleaning will not impact the ability of the fire protection header to deliver the system demand for fire suppression activities in any plant fire area.
Nuclear Safety and Radiological Release Performance Criteria:
The ability to use the 121 MDFP for the Screenwash function has no impact on the radiological release performance criteria. The radiological release review was performed based on the release of firefighting water potentially containing radioactive materials and is not dependent on the MDFP alignment to supplement the Screenwash function.
The ability to use the 121 MDFP for the screenwash function does not change the radiological release evaluation and does not add additional radiological materials to the area or challenge system boundaries.
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Safety Margin and Defense-in-Depth:
The ability to use the 121 MDFP for the Screenwash function does not change the safety margin since it has no impact on the ability of the MDFP to supply water to the fire header unless there is a fire in FA 41. If a fire occurs in FA 41, the DDFP will supply the fire header with adequate flow and pressure.
The use of the MDFP for Screenwash cleaning will not impact the ability of the fire protection header to deliver the system demand for fire suppression requirements.
The three echelons of defense-in-depth are to: 1) prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions). The use of the MDFP for Screenwash cleaning does not affect echelons 1, 2 and 3. The ability to use the 121 motor driven fire pump for the Screenwash function does not directly result in compromising automatic fire suppression functions, manual fire suppression functions, or post-fire safe shutdown capability since water will automatically be re-directed to the fire protection header upon actuation of a fire suppression system or supplied from the DDFP in the event the MDFP is unavailable.
Conclusion:
Aligning the 121 MDFP to perform the screenwash function and emergency uses does not meet the requirement or allowed exceptions of NFPA 805 Section 3.5.16. The evaluation determined that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:
a) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
b) Maintains safety margins; and
c) Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
Therefore, NRC approval is being requested to permit the fire water system to be aligned for the screenwash function and other emergency uses, only when the screenwash pump is out of service.
Northern States Power – Minnesota Attachment M – License Condition Changes
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M. License Condition Changes
6 Pages Attached
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NSPM proposes to replace the current PINGP fire protection license conditions 2.C.(4) for Units 1 and 2 with the standard license condition in Regulatory Position 3.1 of Regulatory Guide 1.205, modified as shown in the proposed markups that follow. It is NSPM’s understanding that implicit in the replacement of the current license condition, all prior fire protection program SERs and commitments will be superseded in their entirety by the revised license condition. No other license conditions need to be replaced or revised. NSPM implemented the following process for determining that these are the only license conditions required to be either revised or superseded to implement the new fire protection program which meets the requirements in 10 CFR 50.48(a) and 50.48(c):
• A review was conducted of the PINGP Unit 1 Renewed License Number DPR-42, through Amendment No. 205 and Unit 2 Renewed License Number DPR-60, through Amendment No. 192. The review was performed by reading the Operating License and performing electronic searches. Outstanding LARs that have been submitted to the NRC were also reviewed for potential impact on the license conditions.
Marked-up License Condition pages follow.
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Unit 1 License Condition 2.C(4):
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Unit 2 License Condition 2.C(4):
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Insert A for License Condition 2.C.(4) for both Units 1 and 2:
NSPM shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated ___________ (and supplements dated____________ ) and as approved in the Safety Evaluation Report dated ____________ (and supplements dated _____________). Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval
A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.
(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1×10-7/year (yr) for CDF and less than 1×10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.
Other Changes that May Be Made Without Prior NRC Approval
(1) Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program
Prior NRC review and approval is not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the
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corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the NSPM PINGP NFPA 805 Transition Report – Attachment M, Revision 0, Page M-3 component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.
The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is “adequate for the hazard.” Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:
• “Fire Alarm and Detection Systems” (Section 3.8);
• “Automatic and Manual Water-Based Fire Suppression Systems” (Section 3.9);
• “Gaseous Fire Suppression Systems” (Section 3.10); and,
• “Passive Fire Protection Features” (Section 3.11).
This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.
(2) Fire Protection Program Changes that Have No More than Minimal Risk Impact
Prior NRC review and approval is not required for changes to the licensee’s fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC Safety Evaluation Report dated ________ to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.
Transition License Conditions
(1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensee’s fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.
(2) The licensee shall implement the modifications described in Attachment S, Table S-2, of the September 2012 PINGP NFPA 805 LAR and as supplemented by letters dated [date] to complete the transition to full compliance with 10 CFR 50.48(c) before the end of the second full operating cycle for each unit after approval of the LAR.
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(3) The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above
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N. Technical Specification Changes
5 Pages Attached
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The following PINGP Technical Specifications (TS) will be revised as indicated:
TS 5.0, Administrative Controls, Section 5.4, Procedures, specification 5.4.1 currently states, in part that written procedures shall be established, implemented, and maintained covering the following activities:
5.4.1.d. Fire Protection Program implementation; and
This LAR proposes to delete the words: “Fire Protection Program implementation” and replace this wording with the words: “Not used.” This change is proposed because after completion of the transition to NFPA 805, the requirement for fire protection program implementation procedures will be contained in 10 CFR 50.48(a) and 10 CFR 50.48(c), as specifically outlined in Section 3.2.3, “Procedures,” of NFPA 805. The requirement to maintain a fire protection program in accordance with 10 CFR 50.48(a) and 10 CFR 50.48(c) is included in the new License Condition described in Attachment M.
No changes to the TS Bases are required to support the transition to the new NFPA 805 FPP. Changes to the PINGP Renewed Operating License include a revision to License Condition 2.C.(4), Fire Protection, as described in Attachment M.
NSPM implemented the following process for determining which TS or TS Bases sections were required to be revised or deleted to implement the new FPP which meets the requirements in 10 CFR 50.48(a) and 50.48(c).
• A review was conducted of the PINGP TS by NSPM Regulatory Affairs personnel assigned to the NFPA 805 transition team. The review was performed by reviewing the TS and performing electronic searches. Outstanding TS changes that have been submitted to the NRC were also included.
The TS Markups and retyped “clean” pages follow.
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Technical Specification Markup
1 Page Follows
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Technical Specification Retype
1 Page Follows
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Northern States Power - Minnesota Attachment O – Orders and Exemptions
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O. Orders and Exemptions
2 Pages Attached
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Exemptions
As described in Section 4.2.3, previously approved exemptions from the requirements of 10 CFR 50, Appendix R have been determined to be either compliant with 10 CFR 50.48(c) or are no longer needed. Therefore, NSPM requests that the following exemptions granted against 10 CFR 50, Appendix R, pursuant to 10 CFR 50.12 in NRC letters dated February 2, 1983, May 4, 1983, January 9, 1984, July 31, 1984, and February 21, 1995 be rescinded:
• An exemption from Section III.G.3 for lack of a fixed fire suppression system in the Control Room, Units 1 and 2, Fire Area 13, (February 2, 1983).
• An exemption from Subsection III.G.2 for lack of twenty feet of separation free of intervening combustibles or one hour fire rated barriers between redundant trains needed for safe shutdown in the “A” Train Hot Shutdown Panel, Instrument Air and Auxiliary Feedwater Pump Rooms, Units 1 and 2, Fire Area 31 (May 4, 1983).
• An exemption from Subsection III.G.2 for lack of twenty feet of separation free of intervening combustibles or one hour fire rated barriers between redundant trains needed for safe shutdown in the “B” Train Hot Shutdown Panel, Instrument Air and Auxiliary Feedwater Pump Rooms, Units 1 and 2, Fire Area 32 (May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Operating Level, Unit 1, Fire Area 60 (May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Operating Level, Unit 2, Fire Area 75 (May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Normal Switchgear Room, Unit 1, Fire Area 37 (May 4, 1983).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Ground Floor Level, Unit 1, Fire Area 58 (January 9, 1984).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Ground Floor Level, Unit 2, Fire Area 73 (January 9, 1984).
• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Mezzanine Level, Unit 1, Fire Area 59 (January 9, 1984).
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• An exemption from Section III.G.2 for lack of an automatic fixed fire suppression system in the Auxiliary Building Mezzanine Level, Unit 2, Fire Area 74 (January 9, 1984).
• An exemption from Section III.G.2 for the lack of twenty feet of separation free of intervening combustibles between redundant trains needed for safe shutdown in the Containment, Units 1 and 2, Fire Areas 1 and 71 (July 31, 1984).
• An exemption from Section III.O for a reactor coolant pump lube oil collection system that does not drain to a vented closed container that can hold the entire lube oil system inventory, but instead is piped to a sump inside Containment and then is pumped to a closed vented container located in the Auxiliary Building; Units 1 and 2, Containment Fire Areas 1 and 71 (July 31, 1984).
• An exemption from Section III.G.1 to allow operators to remove fuses from PORV control circuits to preclude inadvertent valve operation in the event of a control room evacuation; this is considered a repair to ensure that one train of safe shutdown equipment remains operable which is contrary to Section III.G.1, Units 1 and 2, Control Room Fire Area 13 (February 21, 1995).
Specific details regarding these exemptions are contained in Attachment K.
Orders
No Orders need to be superseded or revised.
NSPM implemented the following process for making this determination:
• A review was conducted of the PINGP docketed correspondence by NSPM licensing staff. The review was performed by reviewing the correspondence files and performing electronic searches of internal PINGP records and the NRC’s ADAMS document system.
• A specific review was performed of the license amendment that incorporated the mitigation strategies required by Section B.5.b of Commission Order EA-02-026 (TAC Nos. MD4612 and MD4613) to ensure that any changes being made to ensure compliance with 10 CFR 50.48(c) do not invalidate existing commitments applicable to the plant. The review of this order demonstrated that changes to the fire protection program will not affect measures required by B.5.b.
• The Fukushima Orders are being independently evaluated. Any plant changes will continue to be evaluated for impact on the fire protection program in accordance with the PINGP design change process.
Northern States Power - Minnesota Attachment P – RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)
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P. RI-PB Alternatives to NFPA 805 10 CFR 50.48(c)(4)
No risk-informed or performance-based alternatives to compliance with NFPA 805 (per 10 CFR 50.48(c)(4)) were utilized by Northern States Power - Minnesota.
Northern States Power - Minnesota Attachment Q – No Significant Hazards Evaluation
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Q. No Significant Hazards Evaluations
4 Pages Attached
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A written evaluation of the significant hazards consideration of a proposed license amendment is required by 10 CFR 50.92, “Issuance of amendment.” As described in 10 CFR 50.92, a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or
(2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or
(3) Involve a significant reduction in a margin of safety.
Northern States Power – Minnesota (NSPM) has evaluated the proposed amendment and determined that it involves no significant hazards consideration, using the three standards set forth in 10 CFR 50.92, as discussed below:
1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated? Response: No. Operation of the Prairie Island Nuclear Generating Plant (PINGP) in accordance with the proposed amendment does not increase the probability or consequences of accidents previously evaluated. Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling evaluations, have been performed to demonstrate that the performance-based requirements of National Fire Protection Association Standard 805 (NFPA 805) have been satisfied. The PINGP Updated Safety Analysis Report (USAR) documents the analyses of design basis accidents (DBAs) at PINGP. The proposed amendment does not adversely affect accident initiators nor alter design assumptions, conditions, or configurations of the facility that would increase the probability or consequences of accidents previously evaluated. Further, the changes to be made for fire hazard protection and mitigation do not adversely affect the ability of structures, systems, and components (SSCs) to perform their design functions, nor do they affect the postulated initiators or assumed failure modes for accidents described and evaluated in the USAR. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition will remain capable of performing their design functions. The purpose of this proposed amendment is to permit PINGP to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). Engineering analyses, in accordance
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with NFPA 805, have been performed to demonstrate that the risk-informed, performance-based (RI-PB) requirements per NFPA 805 have been met. NFPA 805, taken as a whole, provides an acceptable alternative to 10 CFR 50.48(b), satisfies 10 CFR 50.48(a) and General Design Criterion (GDC) 3 of Appendix A to 10 CFR 50, and meets the underlying intent of the NRC's existing fire protection regulations and guidance, and provides for defense-in-depth. The goals, performance objectives, and performance criteria specified in Chapter 1 of NFPA 805 ensure that if there are any increases in the net core damage frequency (CDF) or risk associated with this license amendment request (LAR) submittal, the increase will be small and consistent with the Commission's Safety Goal Policy. Based on this, the implementation of this amendment does not significantly increase the probability of any accident previously evaluated. Equipment required to mitigate an accident remains capable of performing the assumed function(s). The proposed amendment will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. Therefore, the consequences of any accident previously evaluated are not significantly increased with the implementation of the proposed amendment.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No. Operation of PINGP in accordance with the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. Any scenario or previously analyzed accident with offsite dose was included in the evaluation of DBAs documented in the USAR. The proposed change does not alter the requirements or function for systems required during accident conditions. Implementation of the new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205 will not result in new or different accidents. The proposed amendment does not introduce new or different accident initiators nor alter design assumptions or conditions of the facility. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and maintain it in a safe shutdown condition remain capable of performing their design functions. The purpose of this amendment is to permit PINGP to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205. The NRC considers that NFPA 805
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provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). The requirements in NFPA 805 address only fire protection and the impacts of fire on the plant that have already been evaluated. Based on this, the implementation of this amendment does not create the possibility of a new or different kind of accident from any kind of accident previously evaluated. The proposed amendment does not introduce any new accident scenarios, transient precursors, failure mechanisms, malfunctions, or limiting single failures that could initiate a new accident. There will be no adverse effect or challenges imposed on a safety related system as a result of this proposed amendment. Therefore, the possibility of a new or different kind of accident from any kind of accident previously evaluated is not created with the implementation of this amendment.
3. Does the proposed amendment involve a significant reduction in a margin of safety? Response: No. Operation of PINGP in accordance with the proposed amendment does not involve a significant reduction in a margin of safety. The proposed amendment does not alter the manner in which safety limits, limiting safety system settings or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by this change. The proposed amendment does not adversely affect existing plant safety margins or the reliability of equipment assumed to mitigate accidents in the USAR. The proposed amendment does not adversely affect the ability of SSCs to perform their design function. SSCs required to safely shut down the reactor and to maintain it in a safe shutdown condition remain capable of performing their design function. The purpose of this amendment is to permit PINGP to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). Engineering analyses, which may include engineering evaluations, probabilistic safety assessments, and fire modeling evaluations, have been performed to demonstrate that the performance-based methods do not result in a significant reduction in a margin of safety.
Northern States Power - Minnesota Attachment Q – No Significant Hazards Evaluation
PINGP Page Q-5 – Revision 0
Based on this, the implementation of this amendment does not significantly reduce a margin of safety. The proposed changes are evaluated to ensure that the risk and safety margins are kept within acceptable limits. Therefore, the transition to NFPA 805 does not involve a significant reduction in a margin of safety.
The proposed amendment to transition to NFPA 805 continues to protect public health and safety and the common defense and security because the overall approach of NFPA 805 is consistent with the key principles for evaluating license basis changes, as described in RG 1.174, is consistent with the defense-in-depth philosophy, and maintains sufficient safety margins.
Based on the responses to questions 1, 2, and 3 above, NSPM has concluded that the proposed amendment presents no significant hazards consideration in accordance with the requirements in 10 CFR 50.92(c).
Northern States Power – Minnesota Attachment R –Environmental Considerations
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R. Environmental Considerations Evaluation
2 Pages Attached
Northern States Power – Minnesota Attachment R –Environmental Considerations
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NSPM has evaluated this LAR against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. NSPM has determined that this LAR meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50. The purpose of this amendment is to permit the Prairie Island Nuclear Generating Plant (PINGP) to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of RG 1.205. The NRC considers that NFPA 805 provides an acceptable methodology and performance criteria for licensees to identify fire protection systems and features that are an acceptable alternative to the 10 CFR 50, Appendix R fire protection features (69 FR 33536, June 16, 2004). The requirements in NFPA 805 address only fire protection and the impacts of fire on the plant have already been evaluated, as part of compliance to 10 CFR 50.48(a) and (b). This proposed amendment to transition the PINGP Fire Protection Program to NFPA 805 would change requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment meets the following specific criteria:
i. The amendment involves no significant hazards consideration. As stated in Section 5.3.1 and Attachment Q, this proposed amendment does not involve a significant hazards consideration.
ii. There is no significant change in the types or significant increase in the amounts
of any effluent that may be released offsite.
Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for radioactive releases to the environment. The radioactive release goal is to provide reasonable assurance that a fire will not result in a radiological release that affects the public, plant personnel, or the environment. Transition to the NFPA 805 requirements does not impact effluents. Therefore, there will be no significant change in the types or significant increase in the amounts of any effluents released offsite.
iii. There is no significant increase in individual or cumulative occupational radiation exposure.
Compliance with NFPA 805 explicitly requires the attainment of performance criteria, objectives, and goals for occupational exposure. There will be no
Northern States Power – Minnesota Attachment R –Environmental Considerations
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significant increase in individual or cumulative occupational radiation exposure resulting from this change.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in conjunction with the proposed amendment.
Northern States Power - Minnesota Attachment S - Modifications and Implementation Items
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S. Modifications and Implementation Items
32 Pages Attached
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Tables S-1, Plant Modifications Completed, and S-2, Plant Modifications Committed, provided below, include a description of the modifications along with the following information:
• A problem statement,
• Risk ranking of the modification,
• An indication if the modification is currently included in the FPRA,
• Compensatory Measure in place if non-compliant with the Current Licensing Basis, and
• A risk-informed characterization of the modification and compensatory measure.
• The following legend should be used when reviewing the Risk Rank in Tables S-1 and S-2:
o High = Modification would have an appreciable impact on reducing overall fire CDF.
o Medium = Modification would have a measurable impact on reducing overall fire CDF.
o Low = Modification would have either an insignificant or no impact on reducing overall fire CDF.
o N/A = Not modeled in the FPRA, therefore a risk ranking is not provided.
NSPM is requesting two full refueling cycles beyond SE issuance to fully implement modifications. This is, in part, due to the outage strategies implemented at PINGP where only one train is removed from service per outage, per unit. Due to the significant modifications required to transition PINGP to NFPA 805, additional time is necessary to fully implement modifications described in Table S-2. As discussed during the April 2012 pre-submittal meeting, NSPM will implement Code Conformance modifications by the completion of the first refueling outage per unit after SE issuance. Therefore: NSPM will complete implementation of the modifications described in Table S-2 as follows:
• By the completion of the first refueling outage per unit after issuance of the NFPA 805 license amendment: Items 8, 9, and 16
• By the completion of the second refueling outage per unit after issuance of the NFPA 805 license amendment: all remaining items A separate evaluation of interim compensatory measures for modifications identified to be risk significant is performed to determine alternative compensatory measures appropriate for risk.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-1 Plant Modifications Completed
Item Rank Unit Problem Statement Modification In
FPRA Comp
Measure Risk Informed
Characterization
1 High 2 The Westinghouse Reactor Coolant Pump (RCP) seal requires cooling from either seal injection, or the component cooling water Thermal Barrier Heat Exchanger (TBHX). If all cooling to the Westinghouse RCP seal is lost, the seal could deform and result in increased leakage from the RCP seal.
Installed new Reactor Coolant Pump (RCP) seals that are not subject to excessive leakage upon loss of all seal cooling.
Yes No The modification reduced risk by installing the Flowserve N-9000 RCP Seal package. The new seal has the ability to preclude larger seal leakage rates during loss of seal cooling scenarios.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
1 High
1
A fire could damage Train B 12 Motor Driven Auxiliary Feedwater Pump (MDAFWP) and the control switches for the 11 Turbine Driven Auxiliary Feedwater Pump (TDAFWP) discharge valves (MV-32238 & MV-32239). Fire damage to CS-51003 could cause spurious closure of MV-32238 which would isolate the 11 TDAFWP flow to the credited 11 Steam Generator. Fire damage to control switch CS-51005 could prevent closing MV-32239 which could divert the 11 TDAFWP flow to the non-credited 12 Steam Generator. The NFPA 805 Nuclear Safety Performance Goal Criteria is not met for Decay Heat Removal. Cables of Concern: 11 AFW to 11 SG MV, MV-32238 (cable 1CA-115) 11 AFW to 12 SG MV, MV-32239 (cable 1CA-116)
Modify equipment in FA 31 to ensure that Train “A” equipment is available for fire safe shutdown. The controls and associated cables for the Unit 1 Train “A” AFWP discharge valves will be moved to Fire Area 32 so they are not damaged by a fire in Fire Area 31.
Yes Yes The modifications proposed by Items 1-4 will reduce risk by modifying FAs 31 and 32 to ensure that each FA has either A-train or B-train related equipment unaffected by a fire. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
2 High
2
A fire in FA 31 could damage the 22 TDAFWP (Train B) and damage the circuits for the Train A 21 MDAFWP (MV-32383 & MV-32384). Fire damage at the Train A Hot Shutdown Panel or MCC 2A1 could affect MV-32383 (21 MDAFWP to 21 SG) or MV-32384 (21 MDAFWP to 22 SG). A fire at MCC 2A1 could affect MV-32026 (21 MDAFWP suction from Cooling Water), MV-32336 (21 MDAFWP suction from CST), MV-32383 (21 MDAFWP to 21 SG) and MV-32384 (21 MDAFWP to 22 SG). The NFPA 805 Nuclear Safety Performance Goal Criteria is not met for Decay Heat Removal. Cables of Concern: 21 MDAFWP to 21 SG, MV-32383 (2A1-5, 2A1-5A, 2CA-115, 2CA-116, 2CA-65) 21 MDAFWP to 22 SG, MV-32384 (2A1-5A, 2A1-6, 2CA-116, 2CA-117, 2CA-66) MDAFWP suction from CST, MV-32026 (2A1-2, 2A1-2A, 2A1-4A, 2CA-30) 21 MDAFWP suction from CST, MV-32336 (2A1-4, 2A1-4A, 2CA-30)
Modify equipment in FA 31 to ensure that Train “A” equipment is available for fire safe shutdown. The controls, MCC power supply, and associated cables for the Unit 2 Train “A” AFWP discharge and suction valves will be moved to Fire Area 32 so they are not damaged by a fire in Fire Area 31. The cables going to Unit 2 Train “A” AFW discharge valves (MV-32383 and MV-32384) will be protected from fire damage so that they will not spuriously close due to a fire in Fire Area 31.
Yes Yes The modifications proposed by Items 1-4 will reduce risk by modifying FAs 31 and 32 to ensure that each FA has either A-train or B-train related equipment unaffected by a fire. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
3 High
1
A fire could damage the 11 TDAFWP (Train A) and the control switches for the 12 MDAFWP discharge valves (MV-32381 & MV-32382). Fire damage at the Train B Hot Shutdown Panel or MCC 1A2 could affect MV-32381 (12 MDAFWP to 11 SG) or MV-32382 (12 MDAFWP to12 SG). A fire at MCC 1A2 could affect MV-32027 (12 MDAFWP suction from Cooling Water), MV-32335 (12 MDAFWP suction from CST), MV-32381 (12 MDAFWP to 11 SG) and MV-32382 (12 MDAFWP to 12 SG). The NFPA 805 Nuclear Safety Performance Goal Criteria is not met for Decay Heat Removal. Cables of Concern: 12 MDAFWP to 11 SG, MV-32381 (1A2-7A, 1CB-52, 1CB-53, 1CB-54) 12 MDAFWP to 12 SG, MV-32382 (1A2-8A, 1CB-52, 1CB-55, 1CB-56) 12 MDAFWP suction from Cooling Water, MV-32027 (1A2-3, 1A2-3A, 1A2-6A) 12 MDAFWP suction from CST, MV-32335 (1A2-6, 1A2-6A)
Modify equipment in FA 32 to ensure that Train “B” equipment is available for fire safe shutdown. The controls, MCC power supply, and associated cables for the Unit 1 “B” AFWP discharge and suction valves will be moved to Fire Area 31 so they are not damaged by a fire in Fire Area 32. The cables going to Unit 1 Train “B” AFW discharge valves (MV-32381 and MV-32382) will be protected from fire damage so that they will not spuriously close due to a fire in Fire Area 32.
Yes Yes The modifications proposed by Items 1-4 will reduce risk by modifying FAs 31 and 32 to ensure that each FA has either A-train or B-train related equipment unaffected by a fire. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
4 High
2
A fire could damage the 21 MDAFWP (Train A) and the control switches for the 22 TDAFWP discharge valves. Fire damage to CS-51605 could cause spurious closure of MV-32247 which would isolate the 22 TDAFWP flow to the credited 22 Steam Generator. Fire damage to control switch CS-51603 could prevent closing MV-32246 which could divert the 22 TDAFWP flow to the non-credited 21 Steam Generator. The NFPA 805 Nuclear Safety Performance Goal Criteria is not met for Decay Heat Removal. Cables of Concern: 22 AFW to 21 SG MV, MV-32246 (cable 2CB-164) 22 AFW to 22 SG MV, MV-32247 (cable 2CB-163)
Modify equipment in FA 32 to ensure that Train “B” equipment is available for fire safe shutdown. The controls and associated cables for the Unit 2 Train “B” AFWP discharge valves will be moved to Fire Area 31 so they are not damaged by a fire in Fire Area 32.
Yes Yes The modifications proposed by Items 1-4 will reduce risk by modifying FAs 31 and 32 to ensure that each FA has either A-train or B-train related equipment unaffected by a fire. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 8 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
5 High 1,2
A fire in FA 18, Relay Room could damage both trains of safe shutdown. Since the risk of Recovery Actions taken in procedure F5 App B, Control Room Evacuation (Fire) are high, installing a Very Early Warning Fire Detection System (VEWFDS) or Incipient Detection is needed to reduce risk in the relay room.
Install Incipient Detection System in the Relay Room that will continuously sample the Relay Room air inside the risk significant cabinets to identify fires based on the detection of the presence of small amounts of products of combustion and, if detected, will sound an alarm in the MCR.
Yes Yes The proposed modification will reduce risk by installing an incipient detection system that will notify operators of fires in their incipent state. This reduces the significance of the fire scenarios that could lead to control room abandonment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
6 Medium 1, 2
Fire damage to cable 1CX-99 in FA 20 could cause a loss of the normal power feed from 13 Inverter to 120VAC Panel 113. Loss of Panel 113 causes CV-31198 (Charging Line to 11 Regenerative Heat Exchanger CV) to fail open causing diversion of flow from RCP seal injection to charging. Loss of Panel 113 causes loss of Control Room indication for instrument Loops 1N51 (Unit 1 Excore Detection Train A), 1T-450A (Unit 1 RCS Loop A Hot Leg Temperature) and 1T-450B (Unit 1 RCS Loop A Cold Leg Temperature). Modification is needed to protect 1CX-99 from fire damage in Fire Area 20 to maintain Process Monitoring indication in the control room.
Reroute the following cables through FA 58 along the "G" line between 8 and 9 and out of FA 20: - 1CX-99 (Instrument Bus III (Blue) Panel 113 Normal Power Feed) - 1CW-99 (Instrument Bus II (White) Panel 111 Normal Power Feed) Install cable - 1DCA-133 (DC Power supply to BUS 15 Load Sequencer) from PNL-11 in FA 33 to BUS-15 Load Sequencer in FA 81 that is not routed through Fire Area 20.
Yes Yes The proposed modification will reduce risk because it will reroute cables associated with the opposite train of equipment to another FA. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 10 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
6 Cont
Fire damage to cable 1CW-99 in FA 20 could cause a loss of the normal power feed from 11 Inverter to Panel 111. Loss of Panel 111 results in the loss of Control Room indication for instrument Loop 1L-487 (11 SG Wide Range Level) displayed on Level Recorder 1LR-470. Modification is needed to protect 1CW-99 (Instrument Bus II (White) Panel 111 Normal Power Feed). Fire damage to cable 1CF-35 in FA 20 could cause a loss of Control Room indication for Loop 1L-433 (Unit 1 Pressurizer Level). Modification to protect cable 1CW-99 from fire damage in FA 20 will ensure Pressurizer Level Indication LOOP 1L-427 remains available in the control room.
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
6 Cont
A fire in FA 20 could damage cable 1C-419 which could affect the ability of BKR 15-3, 1RY source to Bus 15, to clear from the potentially faulted 1RY source to Bus 15. Local manual action is required to open BKR 15-3 so that Bus 15 can be repowered from the D1 Emergency Diesel Generator. A fire in FA 32 or 58 could damage cable 1C-333 affecting the 1RY source to Bus 16, and could damage cable 16408-1, CT11 source to Bus 16, and cables 1DCB-2 and 1DCB-95 which support the D2 source to Bus 16. A modification to route affected conductors of cable 1C-333 out of fire area 32 and 58 is needed to protect the 1RY source to Bus 16 in fire area 32.
Re-route conductors from 1C-419 (Breaker 15-3, Bus 15 Offsite Source from 1R Transformer) to cable 15403-B which is not routed in Fire Area 20. Re-route affected conductors of cable 1C-333 out of FA 32 and FA 58 so the 1RY offsite power supply will be available in FA 32 and FA 58.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
6 Cont
A fire in FA 058/073, 695' elevation of the Aux Building could damage cable 2DCA-105 which provides DC control power to PNL 27 which provides DC control power to Bus 25 to trip 4 KV breakers.
Protect cable - 2DCA-105 (DC Power Cable from 21 Battery 125V DC Panel 27 Train A) from fire induced failure in Fire Area 058/073.
7 N/A N/A
DELETED
N/A
N/A N/A N/A
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
8 Low 1,2
Fire Detection required for the Fire PRA is not code compliant, as required by NFPA 805, for the following Fire Areas: FA-18, 41B, 58/73, 59/74
Install New Fire Detectors per NFPA 72 (Detection) to resolve NFPA 72 code deviations in the following areas: FA 18: Modify the Ionization Fire Detection system to provide two zones of coverage in the Relay Room and P250 Computer Room. Modify the CO2 fire suppression system to actuate if both Ionization zones detect a fire in lieu of heat detectors. FA 41B: Relocate detector from the exhaust stream of a ventilation duct. FA 58/73: Resolve various detector code issues based on S&L Fire Detector Study, Rev 0, PINGP, Project No: 111973-055, 12/20/2008 FA 59/74:Resolve various detector code issues based on S&L Fire Detector Study., Rev 0, PINGP, Project No: 111973-055, 12/20/2008
Yes Yes The proposed modification will reduce risk by allowing the Fire PRA to credit fire detection systems in the listed Fire Areas. Per the 2009 ASME PRA Standard, fire detection systems must be code compliant if they are credited in the Fire PRA. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
9 Low 1,2
Fire Suppression required for the Fire PRA is not code compliant, as required by NFPA 805, for the following Fire Areas: FA 18, 41B, 31, & 32.
Install additional suppression in various rooms to resolve NFPA Suppression code deviations as follows: FA 18: Install an odorizer for the Cardox System. FA 31: Resolve non-compliances with pendant sprinkler heads. FA 32: Resolve non-compliances with pendant sprinkler heads. FA 41B: Install missing Sprinkler #229. FA 41B: Install Heat Activated Detector (HAD) in the enclosure for the 121 Motor Driven Fire Pump. FA 41B: Move or install a sprinkler head above the Diesel Driven Fire Pump because of a large obstruction.
Yes Yes The proposed modification will reduce risk by allowing the Fire PRA to credit fire suppression systems in the listed Fire Areas. Per the 2009 ASME PRA Standard, fire suppression systems must be code compliant if they are credited in the Fire PRA. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
10 High 1,2
A fire could damage DC control cables for 4 KV breakers which could cause the tripping control power fuses to clear which would prevent the breaker from tripping on over-current. The fire could then damage the 4 KV power cable, but since the breaker can't trip, the cable would be subjected to an over-current condition up to the full fault current available to the bus. If the cable is not sized large enough to carry this amount of current, the cable could be damaged and start a fire in other fire areas where it is routed. Affected Breakers: BKR 15-1, BKR 15-3, BKR 15-4, BKR 15-5, BKR 15-7, BKR 15-8, BKR 15-9, BKR 15-12, BKR 16-1, BKR 16-2, BKR 16-3, BKR 16-5, BKR 16-6, BKR 16-7, BKR 16-8, BKR 16-10, BKR 24-1, BKR 25-1, BKR 25-5, BKR 25-7, BKR 25-8, BKR 25-9, BKR 25-10,BKR 25-13, BKR 25-16, BKR 25-17, BKR 26-1, BKR 26-2, BKR 26-5, BKR 26-9, BKR 26-10, BKR 26-11, BKR 26-13, BKR 13-1, BKR 13-3, BKR 14-1, BKR 13-7, BKR 13-8, BKR 14-3, BKR 23-1, BKR 23-4, BKR 23-5, BKR 24-2, BKR 24-5, BKR 24-6
Modify 4160 volt switchgear control circuits so that faults on the control cables will not prevent the over-current trip relay from protecting the cable from over-current conditions that could lead to cable damage and secondary fires or loss of bus coordination.
Yes Yes The FPRA assumes coordination of credited buses. This modification ensures there are no secondary fires. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
11 Medium 1
A fire at CV-31998 could damage cables 1CA-1109, 1CA-1111 and 1CA-1248. This could cause spurious energizing of SV-33299 and maintain CV-31998, 11 TDAFWP steam supply valve closed. Damage to cable 1CA-1111 or 1CA-1248 could cause CV-31153, 11 TDAFWP recirculation lube oil cooler line to close and also damage the 11 TDAFWP.
Protect cable 1CA-1109 from inter-cable hot shorts from cables 1CA-1111 and 1CA-1248 so CV-31998 can open to supply steam to 11 TDAFWP. Re-wire CV-31153 to spare relay contacts in FA 32 instead of the limit switches on CV-31998 in FA 69 so it is not affected by a fire in FA 69.
Yes Yes The modification will reduce risk by ensuring the 11 TDAFW Pump is protected from fires in FA 69. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
12 Medium 1
A fire in Fire Area 29 could damage cables required for operation of Train A Cooling Water Pumps which supply cooling water to D1 Emergency Diesel Generator which powers Train A safeguards Bus 15. A fire in Fire Area 69 could damage cables that supply power to ventilation fans for D2 Emergency Diesel Generator supply to Train B safeguards Bus 16. A fire in Fire Area 69 could also damage cables required for offsite power to Bus 15 and Bus 16. Fire Area 29 and 69 were defined as separate areas in the 1977 Fire Hazards Analysis that was submitted to the NRC and accepted. There is an open pathway between Fire Area 29 and 69; therefore the Fire PRA Plant Partitioning combined Fire Area 29 and 69 (along with 8, 14, 27, and 70) into Fire Compartment 8GRP. Risk for Fire Compartment 8GRP was high when all of these areas were combined.
Install a rated fire barrier between FA 29 and FA 69 to provide separation between cables for Train A Cooling Water Pumps that supply Cooling water to D1 Emergency Diesel Generator and offsite power supply cables in the Turbine Building.
Yes No This modification will reduce risk by increasing the availability of the 12 DDCLP by physically separating Fire Areas 29 and 69.
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
13
Medium 1
A fire in the control room or relay room could cause hot shorts on cables that could spuriously start D1 and close the cooling water supply valve. This condition results in unrecoverable damage to the credited Emergency Diesel Generator during a fire induced control room evacuation.
Wire additional relay contacts off the low speed relay in series with indicating light in the control room so that once D1 > 250 RPM, the potential hot short on the indicating light in the control room is cleared. This work is being performed by plant Electrical Design Engineering under EC 18746.
Yes Yes This modification will reduce risk by increasing the availability of the D1 diesel generator for fires that are postulated in the control room and relay room.
Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
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Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
14 Medium 1,2
A fire in FA 13/18 could damage cables causing multiple spurious operations that could damage D1 Emergency Diesel Generator. If fire induced cable damage caused multiple spurious operations that caused D1 (034-011) to spuriously start with no cooling water (CV-31505, 11 MDCLP MTR 13-8, 12 DDCLP 145-392, 21 MDCLP MTR 23-4, 22 DDCLP 245-392) then the EDG could be damaged. A fire that results in evacuation of the control room requires one operator (STA) to go to D1 to verify it is not running without cooling water. Another operator (U1 RO) goes to the screenhouse to locally start the 12 and 22 Diesel Driven Cooling Water Pumps (DDCLP). Both actions are considered Recovery Actions because they are not performed at the emergency control station so the risk of these actions has to be evaluated based on the guidance in NFPA 805 Frequently Asked Questions (FAQ-07-0030).
Protect control circuits for the Diesel Driven Cooling Water Pump to eliminate the current required recovery action of sending an operator to the D1 Room and Screenhouse.
Yes No
This modification will reduce risk by simplifying restoration of Cooling Water to provide cooling to D1 Emergency Diesel Generator and a backup water supply to the Aux Feedwater Pumps.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 20 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
15 Medium 1,2
Fire-induced damage that could damage cables causing multiple spurious operations resulting in damage to the charging pumps. If fire induced cable damage caused spurious isolation of letdown to the VCT (CV-31226 and CV-31255) and failure to open the RWST supply (MV-32060) and failure to trip the charging pumps, the positive displacement charging pumps (MTR 111J-1) and MTR 211J-1 could be damaged due to lack of Net Positive Suction Head (NPSH). Need to prevent unrecoverable damage to credited charging pump due to fire in FA 13/18 to resolve MSO issue.
Install suction pressure protection for all the charging pumps if inadequate Net Positive Suction Head (NPSH) exists to prevent damage to the charging pumps.
Yes Yes The proposed modification will reduce risk by installing suction pressure protection that will protect the charging pumps against fires that involve spurious valve closure and other failures that impact NPSH for the charging pumps. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 21 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
16 N/A 1
The supply ventilation duct between Fire Areas 32 and 37 is missing a fire damper and is not code compliant for the Fire PRA, as required by NFPA 805. There is a ventilation duct between Fire Areas 32 and 37 that is missing a fire damper.
Install a fire damper in the supply ventilation duct between FA 32 and 37.
No Yes
Not modeled in the FPRA, therefore a risk ranking is not provided. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
17 N/A N/A
DELETED N/A N/A N/A N/A
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 22 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
18 High 1
A fire in Fire Area 1, 13, 18, 59, and 71 could cause a loss of all Reactor Coolant Pump (RCP) seal cooling by damaging RCP seal injection from charging and Component Cooling (CC) water to the Thermal Barrier Heat Exchanger (TBHX). If all seal cooling is lost for more than 13 minutes, the seals could be damaged and leak excessively.
Install new RCP seals that would not be subject to excessive leakage if all seal cooling is lost.
Yes Yes The proposed modification will reduce risk by installing the Flowserve N-9000 RCP Seal package. The new seal will have the ability to preclude larger seal leakage rates during loss of seal cooling scenarios. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
19 N/A N/A DELETED
N/A N/A N/A
N/A
20 High 1,2
The current Fire PRA Model assumes proper coordination exists for all credited power supplies. Per Fire PRA credited power supplies lack selective coordination.
Install the appropriate fuses and/or breakers to establish proper selective coordination for panels credited to be coordinated in the Fire PRA.
Yes No The Fire PRA assumes proper coordination of these power supplies
21 N/A N/A
DELETED N/A N/A N/A N/A
22 N/A N/A DELETED N/A N/A N/A N/A
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 23 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
23 Medium 2 A fire in Bus 27 room (Fire Area 127) could damage DC control power to Bus 25 or Bus 26.
Install fuses to provide coordination so that a fire in the Bus 27 room will not affect DC control power to Bus 25 or Bus 26
Yes Yes The Fire PRA assumes proper coordination of these power supplies
24 High 1, 2 A fire in the Bus 15 (Fire Area 81) or Bus 16 (Fire Area 20) room could damage the cables and bus duct that supply off-site power (CT11 and 1R transformers) to Bus 15 and Bus 16 due to common power supply. The redundant diesel generator remains unaffected by a fire to re-power the unaffected 4 kv safeguards bus (Bus 15 or Bus 16), but risk is higher than desired.
Provide fuse/breaker coordination for the CT11 supply to Bus 15 and Bus 16 so that the CT11 source remains available to Bus 15 if a fire damages Bus 16 or to Bus 16 if a fire damages Bus 15.
Re-wire source and target conductors in cables 1CS-3 and 1CS-4 so fire damage to cables 1CS-3 and 1CS-4 would not spuriously trip the CT11 source to Bus 15 and Bus 16.
Yes No The proposed modification will reduce risk by ensuring one off-site power source to the safeguards 4 kV Bus remains unaffected by a fire in the event of a fire in the opposite train safeguards 4 kV Bus room.
25 Medium 1 A fire in Fire Area 32 could damage cables required to open MV-32077 and MV-32078 to provide recirculation from Sump B
Re-power MV-32078 from an MCC that is not located in Fire Area 32 to re-gain the ability to recirculate water from Sump B.
Yes No This will reduce risk by ensuring a fire in FA 32 does not damage the ability to recirculate water from Sump B.
26 Medium 2 A fire in Fire Area 31 could damage cables required to open MV-32180 and MV-32181 to provide recirculation from Sump B
Re-power MV-32180 from an MCC that is not located in Fire Area 31 to re-gain the ability to recirculate water from Sump B.
Yes No This will reduce risk by ensuring a fire in FA 31 does not damage the ability to recirculate water from Sump B.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 24 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
27 High 1, 2 A fire in the Control Room (Fire Area 13) or the Relay and Cable Spreading Room (Fire Area 18) could cause spurious opening of valves that could lead to a loss of inventory.
Install switches in the Control Room to isolate Excess Letdown, Head Vents, Pressurizer Vents, Pressurizer PORV, and Pressurizer Heaters.
Yes No This will reduce risk by providing a way to isolate RCS Excess Letdown, Head Vents, Pressurizer Vents, Pressurizer PORV, and isolating Pressurizer Heaters from the control room.
28 Medium 2 A fire in the Aux Building 715’ elevation (Fire Area 59) could damage cables required for closing the Main Steam Isolation Valves (MSIV).
Protect cables 2CB-415 and 2CB-703 for operation of MSIV in Fire Area 59.
Yes No This will reduce risk by ensuring the capability to isolate the MSIV to control heat removal from the reactor.
29 Medium 2 A fire in the Aux Building 715’ elevation (Fire Area 59) could damage cables that affect the 2RY source to BUS-25
Protect cables 2C-3919 and 2C-3920 from fire damage in Fire Area 59 so that a fire will not cause a loss of the ability for BUS-25 to be re-powered if the 2RY offsite power source is affected.
Yes No This will reduce risk by ensuring the 2RY source remains available to BUS-25.
30 Medium 2 A fire in Fire Area 31 or 32 could damage cables that provide DC power to vital auxiliaries which impacts risk.
Protect risk significant cables (2DCA-10, 2DCA-105, 2DCA-87) from risk significant fire initiators in Fire Areas 31 and 32.
Yes No This will reduce risk by ensuring DC power remains available for control and instrumentation for fire in FA 31 or 32.
31 Medium 2 A fire in Fire Area 58 could affect cables for MTR-25-10, 21 Motor Driven Aux Feedwater Pump (MDAFWP), CV-31418, 21 MDAFWP lube oil cooler, and cables for MV-32019 and MV-32020 which supply steam to the 22 Turbine Driven Aux Feedwater Pump.
Protect cables 211E-1, 25410-1, 25410-C, 25410-D, 25410-E, and 25410-G from risk significant fire initiators to ensure availability of Aux Feedwater for Unit 2.
Yes No This will reduce risk by ensuring the availability of the 21 MDAFWP for a fire in FA 58.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 25 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
32 Medium 2 A fire in Fire Area 32 could damage cable 1C-2278 which could fail the DC control power to BKR-23-4 which powers the 21 Motor Driven Cooling Water Pump (21 MDCLP). The 21 MDCLP is normally running so damage to this cable would not cause a trip of the 21 MDCLP, but simply loss of the ability to start or stop the 21 MDCLP.
Protect cable 1C-2278 from failing Bus 23 in Fire Area 32.
Yes No This will reduce risk by ensuring the availability of the 21 MDCLP for a fire in FA 32.
33 Medium
1 A fire in Fire Area 32 could damage the cable that provides DC control power to PNL-16 which supports Instrumentation.
Protect cable 1DCB-18 from fire damage in Fire Area 32
Yes No This will reduce risk by ensuring the availability of DC power to PNL-16 to provide vital instrumentation.
34 Medium 1, 2 A fire in FA 13, 18, 32 or 58 could damage cables that control the Emergency Diesel Generator output breaker and bypass the sync check switch or relay which could cause a spurious closure of the breaker which could cause a lock-out of the 4 kV safeguards Bus which powers one train of safeguards equipment. The loss of the Bus is risk significant for some fire scenarios.
Protect cables that could bypass the sync check switch or relay from risk significant fire initiators.
Yes No This will reduce risk by making modifications to reduce the number of fire scenarios that could cause fire damage to a 4kV safeguards bus.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 26 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
35 Medium 1 A fire in Fire Area 32 or 58 could damage cables which support operation of the 1RY offsite power sources to BUS 15 (BKR-15-3) and BUS 16 (BKR-16-2).
Protect cables 1C-332 from fire damage in Fire Area 32 and 58 to ensure BUS 16 can be powered from the 1RY transformer.
Yes Yes The proposed modification will reduce risk because it will reroute cables associated with the opposite train of equipment to another FA. This will limit the number of fire scenarios that could damage both trains of equipment. Compensatory measures in accordance with the Current Fire Protection Licensing Basis are being maintained. Compensatory measures will continue to remain in effect after the NFPA 805 fire protection program becomes effective until this modification is complete.
36 Medium 1, 2 A fire in Fire Area 13 or 18 could damage cables that could over-torque MV-32085 and MV-32188 which are credited in the Fire PRA to be locally closed to isolate a potential drain down of the RWST to Sump B.
Re-wire MV-32085 and MV-32188 so that a fire in fire area 13 or 18 will not bypass the torque and limit switches and subsequently damage the valves.
Yes No The proposed modification will allow the valve to be locally operated to credit this recovery action in the PRA. Locally closing these valves provides an additional way to isolate drain down of the RWST to Sump B.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 27 – Revision 1
Table S-2 Plant Modifications Committed
Item Rank Unit Problem Statement Proposed Modification In
FPRA Comp
Measure Risk Informed
Characterization
37 Low 1, 2 A fire in Fire Area 13 or 18 could damage cables that could cause multiple simultaneous spurious operations that could close multiple 4kV loads onto the 4kV safeguards bus and over-load the Emergency Diesel Generator.
Ensure that there are no credible scenarios where fire induced cable damage could simultaneously close multiple load breakers and over-load the credited Emergency Diesel Generator.
No No This modification does not have an impact on the Fire PRA as the subject MSO requires more than two spurious operations to occur, and therefore is not modeled in the Fire PRA.
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 28 – Revision 1
Table S-3, Implementation Items provided below are those items (procedure changes, process updates, and training to affected plant personnel) that will be completed prior to implementation of the new NFPA 805 fire protection program. This will occur within the later of six (6) months after NRC approval, or six months after a refueling outage if one is in progress at the time of approval. Note that Item 20 is associated with modifications described in Table S-2 and will be completed as part of the modification process in accordance with the timetable provided in Section 5.5.
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
1 1, 2 Implement monitoring program required by NFPA 805 Section 2.6 in accordance with NFPA 805 FAQ 10-0059, including a process that reviews the FPP performance and trends in performance.
4.6.2, Attachment A Section 3.2.3(3)
2 1, 2 Revise plant procedure 5AWI 3.13.3, "Hot Work," to address the following: - Address the requirements for hot tapping. (NFPA 51B–1999, Section 3-5) - Address the requirements for a fire watch where torch-applied roofing hot work operations are in effect. (NFPA 241–1999, Section 5.1.3.2) -Address the requirement that open flames or combustion-generated smoke shall not be permitted for leak or air flow testing. - Consider delaying hot work in the vicinity of risk significant components during High Risk Evolutions
Attachment A Section 3.3.1.3.1 Attachment A Section 3.3.1.3.3
Attachment D
3 1, 2 Revise procedure F5 App J, "Fire Drills," to require that fire brigade drills be conducted in various plant areas.
Attachment A, Section 3.4.3 (C)(3)
4 1,2 Perform a calculation to demonstrate that the fire water supply is capable of delivering the largest design demand with the hydraulically least demanding portion of fire main loop out of service in accordance with NFPA 805 requirements.
Attachment A Section 3.5.1
5 N/A DELETED N/A
6 1, 2 Revise procedure F5, “Firefighting,” Section 7, to include a Section 7.5, Control of Spread of Contamination, to address ventilation, floor drains, opening walkways or stairs between areas, and salvage/overhaul activities.
Attachment E
7 1, 2 Revise Fire Brigade Training Lesson Plan R7637L-010, Section VI, Fire Attack, to address the spread of contamination during firefighting activities.
Attachment E
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 29 – Revision 1
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
8 1, 2 Revise Fire Brigade Training Lesson Plan R7637L-011, Section III, Brigade Member Responsibilities, to identify the responsibilities of each brigade member relative to limiting the spread of cross contamination when fighting fires in radiologically controlled areas.
Attachment E
9 1, 2 Revise Fire Brigade Training Lesson Plan R7637-035, Section IV.F, Size Up Possibilities, to provide sufficient details on the impact of fire fighting activities on the potential spread of contamination, and the methods available for mitigating such cross contamination via ventilation and drainage control.
Attachment E
10
1, 2 Revise procedure F5 App A, “Fire Strategies”, to include information on potential cross-contamination for each fire area.
Attachment E
11 1, 2 Revise procedure F5, “Firefighting”, Section 2.7 to address potential access requirements for the Duty RP Tech or Chemist.
Attachment E
12 1, 2 Revise Radiation Protection Continuing Training to address control of contamination during firefighting activities.
Attachment E
13 1, 2 Revise procedure F5, App A, “Fire Strategies” to address the ability to utilize the Auxiliary Building Special Ventilation, Containment Internal Cleanup Subsystem, Containment Purge, Containment In-Service Purge, and Shield Building Ventilation System for the removal of potentially contaminated smoke in fire areas identified in Attachment E.
Attachment E
14 N/A DELETED N/A
15 1, 2 Provide a container with booms, portable filtered ventilation, and other appropriate equipment for the containment of water in the Low Level Rad Waste Enclosure.
Attachment E
16 1, 2 Provide procedures to utilize a combination of containerization and administrative controls to ensure that exposed contaminated waste in the Low Level Rad Waste Enclosure are kept as low as reasonably achievable.
Attachment E
17 1, 2 Revise F5 App F, “Fire Hazards Analysis” to align with the fire area descriptions listed in Attachment I.
Attachment I
18 N/A DELETED N/A
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 30 – Revision 1
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
19 1, 2 Revise procedures and checklists to operate with 480VAC breakers open for RHR suction valves: MV-32231 (U1), MV-32165 (U1), MV-32233 (U2) and MV-32193 (U2). The series counterparts to these valves already have their 480VAC breakers in the open position during power operations.
Attachment B
20 1, 2 Update the Fire PRA Model, as necessary, after all modifications identified in Table S-2 are complete and as-built. If the revised Fire PRA indicates an increase in risk metrics such that the RG 1.205 acceptance guidelines are not met, the configuration control process described in LAR Section 4.7.2 will be implemented.
4.8.2
21 N/A DELETED N/A
22 1, 2 Create new Fire Protection Design Basis Document to reflect content requirements of NFPA 805.
4.7.1
23 N/A DELETED N/A
24 N/A DELETED N/A
25 1, 2 Provide a Change Evaluation Process procedure in accordance with the requirements of NFPA 805.
4.7.2
26 1, 2 Develop qualification requirements and position-specific training for personnel involved with the Fire PRA.
4.7.3
27 1, 2 Revise procedure 5AWI 3.13.0, “Fire Protection Program,” to add Non Power Operations (NPO) overview, definitions; road map; and risk reduction requirements for all NPO, and High Risk Evolutions (HRE).
4.3.2 and Attachment D
28 1, 2 Revise GEN-PI-059, “10CFR50, App R, Safe Shutdown Database Data Verification” and other configuration control procedures which govern the various PINGP documents and databases that currently exist (or develop new procedures/processes) to reflect the new NFPA 805 licensing bases requirements.
4.7.2
29 1, 2 Revise system level design basis documents to reflect NFPA 805 requirements and superseding of the old fire protection licensing basis.
4.7.2
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 31 – Revision 1
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
30 1,2 Revise/initiate procedures and/or procure additional compressed air bottles to achieve 30 hours to ensure we are “safe and stable” at 24 hours.
4.5
31 N/A DELETED N/A
32 1,2 Revise FP-E-MR-01, “Maintenance Rule Process” to add High Safety Significant SSCs that require monitoring based on the Fire PRA.
4.6.2
33 N/A DELETED N/A
34 1, 2 Revise Design Calculations ENG-EE-177, 194401-2.3-008, 12911.6214-E-01 and ENG-EE-013 to support the Fire PRA credited power supply breaker – fuse coordination. Additionally, revise Design Calculation ENG-EE-177 per AR 01342798-02 to support the Loss of DC Control Power Analysis.
4.5 and Attachment B
35 1, 2 Revise FP-OP-ROM-01, “Refueling Outage Management” procedure for inclusion of NPO requirements.
4.3.2 and Attachment D
36 N/A DELETED N/A
37 N/A Revise 5AWI 3.13.3, Hot Work to avoid hot work in certain areas during high risk evolutions.
N/A
38 1,2 Revise F5 App K, “Fire Protection Systems Functional Requirements” to contain the compensatory actions to be implemented should a fire protection system required to be operable during HRE periods be found to be impaired.
4.3.2 and Attachment D
39 1,2 Revise EM 3.4.1, “Review of Proposed Changes to the Fire Protection Program” to contain guidance to ensure that changes to the fire protection program are reviewed for impact to the NPO requirements and risk reduction actions.
4.3.2 and Attachment D
40 1,2 Revise 5AWI 15.6.1, “Shutdown Safety Assessment” to contain discussion on HRE, risk due to fire, NFPA 805 and the NPO requirements as part of risk management.
4.3.2 and Attachment D
41 1 Revise D2-1, “Draining the Reactor Coolant System – Unit 1”, to contain a check list of the HRE risk reduction actions and requirements to implement the HRE risk reduction actions as a pre-requisite to conduct of the procedure which results in RCS reduced inventory states.
4.3.2 and Attachment D
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 32 – Revision 1
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
42 2 Revise D2-2, Draining the Reactor Coolant System – Unit 2, to contain a check list of the HRE risk reduction actions and requirements to implement the HRE risk reduction actions as a pre-requisite to conduct of the procedure which results in RCS reduced inventory states.
4.3.2 and Attachment D
43 N/A DELETED N/A
44 N/A DELETED N/A
45 1 Revise 1C1.6, “Shutdown Operations – Unit 1” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
46 2 Revise 2C1.6, “Shutdown Operations – Unit 2” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
47 1 Revise 1C4.1, “RCS Inventory Control Pre-refueling” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
48 2 Revise 2C4.1, “RCS Inventory Control Pre-refueling” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
49 1 Revise 1C4.2, “RCS Inventory Control – Post Refueling” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
50 2 Revise 2C4.2, “RCS Inventory Control – Post Refueling” to contain a requirement to maintain a continuous action to monitor T-Boil until the HRE can be exited then an HRE Risk Reduction Actions exit check list is provided.
4.3.2 and Attachment D
51 1, 2 Revise EM 3.4.3, “Safe Shutdown Circuit Analysis” to incorporate applicable details of vendor document EPM-DP-EP-004.
4.3.2 and Attachment B
Northern States Power - Minnesota Attachment S – Modifications and Implementation Items
PINGP Page S- 33 – Revision 1
Table S-3 Implementation Items
Item Unit Description LAR Section / Source
52 1, 2 Develop a calculation titled “Nuclear Safety Capability Assessment (NSCA) Analysis for Compliance with NFPA 805,” to establish a design basis for the NSCA model and supporting analyses.
Attachment B
53 N/A DELETED N/A
54 N/A DELETED N/A
55 N/A DELETED N/A
56 N/A DELETED N/A
57 1, 2 Revise procedure F5 App B “Control Room Evacuation (Fire)” to direct the isolation of containment prior to leaving the control room, and to incorporate credited Recovery Actions.
Attachment G Attachment W
58 1, 2 Revise F5 App D, “Impact of Fire Outside Control/Relay Room” as required to include fire response HFEs in the Fire PRA Model and credited recovery actions from Attachment G.
Attachment G and W
59 N/A DELETED N/A
60 1, 2 Revise ENG-ME-353, “Mechanical MOV Analysis to Support IN 92-18 Response” to incorporate updated vendor information as identified in AR 01422482.
Attachment B
61 1, 2 Verify that site procedures and compensatory measures for control of combustibles agree with assumptions in the Fire PRA.
Generic RAI 32
62 1, 2 Perform an Evaluation for PNL 117 and PNL 217 to confirm the loading capacity to supply more than one Instrument Bus.
Attachment D and G
63 1, 2 Provide a procedure to connect a portable generator to power a temporary fan for the Main Control Room to maintain safe and stable.
4.2.1.2
Northern States Power - Minnesota Attachment T – Clarification of Prior NRC Approvals
PINGP Page T-1 – Revision 0
T. Clarification of Prior NRC Approvals
2 Pages Attached
Northern States Power - Minnesota Attachment T – Clarification of Prior NRC Approvals
PINGP Page T-2 – Revision 0
Introduction
The elements of the pre-transition fire protection program licensing basis for which specific NRC previous approval is uncertain are included in this attachment. Also included is sufficient detail to demonstrate how those elements of the pre-transition fire protection program licensing basis meet the requirements in 10 CFR 50.48(c) (RG 1.205, Revision 1, Regulatory Position 2.2.1).
Prior Approval Clarification Request 1 of 1: Operator Action to Isolate Power to PORV Control Circuits
Pre-transition Fire Protection Program Licensing Basis:
The Prairie Island Nuclear Generating Plant (PINGP) pre-transition licensing basis relative to the preclusion of spurious operation of pressurizer power operated relief valve (PORV) flow paths, for fires involving control room evacuation, included a previously approved exemption from the requirements of Section III.G.1 of Appendix R to 10 CFR 50. Specifically, the exemption allowed operators to close the Unit 1 and 2 PORV block valves prior to evacuating the control room, and then taking the follow-on action to remove control power fuses from the PORV control circuits for both units at their respective branch circuit panels. The exemption was required because the removal of fuses involved the use of a fuse-pulling tool, which was considered to be a “repair action.” This repair action was interpreted as a non-compliance to Section III.G.1 of Appendix R to 10 CFR 50 which requires, in part, that fire protection features shall be provided for structures, systems, and components important to safe shutdown so that one train of systems necessary to achieve and maintain hot shutdown conditions be free of fire damage. This exemption was approved in a letter dated February 21, 1995. In 1999, PINGP performed a plant modification (99DC03), which included modification of the PORV control power supplies such that disconnect switches could be used in lieu of pulling control power fuses. The feasibility of utilizing the disconnect switches (no tool required) has been validated and has proven to be a beneficial change with respect to this activity.
Background/Basis:
NSP Exemption Request Letter, dated May 2, 1994 NSP requested an exemption from the requirements of Section III.G.2 of Appendix R to 10 CFR 50, to allow the manual removal of fuses from the PORV control circuits in the event of a fire, in lieu of modifying plant hardware. The reference to III.G.2 was later revised to III.G.1 during a follow-up phone call between NSP and NRR.
Issuance of Exemption Letter, dated February 21, 1995 The NRC issued an exemption from certain requirements of Appendix R to 10 CFR Part 50 to allow NSP to remove fuses from the PORV control circuits as a means of ensuring the reactor coolant system inventory in the event of a control room fire.
Northern States Power - Minnesota Attachment T – Clarification of Prior NRC Approvals
PINGP Page T-3 – Revision 0
PINGP Plant Modification (99DC03) Summary: This modification relocated EQ circuit power supplies from harsh environments to mild environments. This modification repowered the Unit 1 and 2 PORV control circuits, from new distribution panels PNL 171, PNL 181, PNL 271, and PNL 281 respectively, which were, in turn, powered by upstream feeder distribution panels PNL 11, PNL 12, PNL 21, and PNL 22 respectively. An added benefit of this modification is that it allowed the PORV control circuits to be de-energized via disconnect switches in the feeder distribution panels, thus eliminating the need to pull control power fuses for fire events requiring control room evacuation.
Request
As part of this LAR submittal and transition to NFPA 805, it is requested that the NRC accept the following clarification of a prior NRC approval, with respect to the exemption granted to NSP on February 21, 1995:
This operator action (recovery action) to preclude PORV opening remains a required action for PINGP under NFPA 805. The use of recovery actions is not allowed under the deterministic requirements of NFPA 805 Section 4.2.3.1. Clarification is requested to allow the previous exemption for operator actions to be extended to the NFPA 805 program. In addition, clarification is requested to extend the previous allowance to pull fuses to also allow the operation of disconnect switches. Although fuse removal remains an option, the manual operation to open disconnect switches, demonstrated by PINGP to be feasible and reliable, is simpler than pulling fuses and therefore, for the purposes of this request, is requested to be deemed equivalent in intent and function. Clarification is also requested that the term “control room fire”, as referred to in the exemption letter, applies to fires occurring in Fire Area 013 (Control Room) and Fire Area 018 (Relay Room). Under the pre-transition (Appendix R) program, both Fire Area 013 and Fire Area 018 were analyzed as one analysis area.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-1 – Revision 1
U. Internal Events PRA Quality
18 Pages Attached
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-2 – Revision 1
U.1 Internal Events PRA Model The Prairie Island Nuclear Generating Plant (PINGP) base internal events PRA, Revision 3.1, was the starting point for the Fire PRA. With the exception of Internal Flooding Events, the PINGP Rev 3.1 Level 1 analysis evaluates core damage frequency (CDF) from all internal initiating events consistent with the most current combined PRA Standard, ASME/ANS RA-Sa-2009, and RG 1.200, Revision 2. The NRC clarifications and qualifications in the NRC endorsement of the Standard contained in Revision 2 of RG 1.200, Revision 2 were considered during the peer review. The PINGP Rev 3.1 PRA evaluates large early release frequency (LERF) utilizing the Westinghouse Owner’s Group Simplified Level 2 Analysis Approach, WCAP-16341. A self-assessment of the PINGP PRA was conducted by NSPM personnel, and an independent peer review was performed of the PRA model, data, and documentation in accordance with the 2009 ASME PRA Standard Capability Category (CC) II requirements. Two Peer Reviews were performed against the internal events PRA model, data, and documentation. The first peer review includes the PINGP at-power/internal events PRA and specifically addressed eight of the nine technical elements plus the configuration control element. The first peer review did not include applicable SRs for the internal flooding hazard. A focused peer review was performed to specifically review the internal flooding PRA. For both peer reviews, the peer team met the independence requirement of the 2009 ASME PRA Standard. Since the issuance of the PINGP Rev 3.1, the internal events PRA has undergone one (1) maintenance update and one (1) upgrade. The Rev 3.1 model underwent a maintenance update that resulted in the issuance of the Rev 4.0 Level 1 model. The Rev 4.0 model was issued to include the quantification of the internal flooding hazard along with some PRA model maintenance updates. The Rev 4.0 model was then upgraded to Rev 5.0 to include modeling for Flowserve N-9000 Abeyance Reactor Coolant Pump Seal Loss of Coolant Accident modeling for Unit 2 along with some PRA model maintenance updates. A peer review for the upgraded Reactor Coolant Pump Seal Loss of Coolant Accident modeling is scheduled for May 2014. U.2 Internal Events PRA Peer Review Results An independent peer review team evaluated the PINGP Rev 3.0 PRA, which did not include the Internal Flooding analysis. The peer team concluded that 256 (>96%) of the total 264 numbered supporting requirements (SRs) outlined within the 2009 ASME PRA Standard fully met the Capability Category II or greater, not including the Internal Flooding SRs. Two of the internal events SRs (AS-B4 and QU-B10) were determined to not be applicable to the PINGP PRA and are not discussed any further. Six of the applicable SRs were rated as CC I, or as “Not Met.” LE-C3 and HR-D2 met CC I falling below the CC II threshold while LE-G5, IE-C10, IE-C14, and QU-C2 did not meet the Capability Category requirement criteria. NSPM resolved the 6 SRs that were rated as CC I or as “Not Met,” to meet a CC II per the independent peer review team suggestions. Table U-1 contains the 6 SRs that were rated as CC I or as “Not Met,” the peer review assessment description, and the means by which NSPM resolved these findings.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-3 – Revision 1
The independent peer review team identified 65 Facts and Observations which comprised of 22 findings, 41 suggestions, and 2 best practices. Table U-1 describes the 22 findings including the peer review team’s assessment comments and the PINGP resolutions. The Findings/Observations in Table U-1 are provided from the final PRA peer review report with minor editorial corrections. The internal events peer review report is documented in LTR-RAM-II-11-005, RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the Prairie Island Nuclear Generating Plant Units 1 and 2 Probabilistic Risk Assessment, Final Deliverable, March 2, 2011, and is available upon request. U.3 Internal Flooding Peer Review Results An independent peer reviewer evaluated the RG 1.200 internal flooding analysis that was quantified using the PINGP Rev 3.1 PRA model. The peer review was performed in September 2012 and concluded that 61 (98%) of the total 62 numbered supporting requirements (SRs) outlined within the 2009 ASME PRA Standard for At-Power Internal Flooding met Capability Category II or greater. All SRs were rated as CC II or above with the exception of IFQU-A6, which did not meet the Capability Category required criteria. The NRC clarifications and qualifications in the NRC endorsement of the Standard contained in Revision 2 of RG 1.200, Revision 2 were considered during the peer review. The independent peer review identified 8 Facts and Observations which are comprised of 5 findings, 2 suggestions, and 1 best practice. Table U-2 lists the 5 findings, including the peer review assessment comments. The flooding peer review report is documented in LTR-RAM-II-12-092, RG 1.200 PRA Peer Review Against the ASME PRA Standard Requirements for the Prairie Island Nuclear Generating Plant Internal Flood Probabilistic Risk Assessment and is available upon request.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-4 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
Closed Sections 5.x-6 discuss the adverse environmental condition in terms of increasing containment temperature and pressure. However, there is no discussion of potential plugging of the sump screen by the debris generated in a LOCA (e.g., a large LOCA). Basis for Significance: No discussion of potential plugging of the sump screen by the debris generated in a LOCA (e.g., a large LOCA).
Reviewed WCAP-16882-NP, Rev. 1. This document provides a methodology and example application for determining representative screen plugging frequencies based on the initiating event (Non-LOCAs, LOCAs, and Steam line breaks inside and outside containment) that lead to the need for sump recirculation. Incorporated this into the PRA model. The Residual Heat Removal System Notebook was updated to document the re-modeling of the sump strainer plugging logic by initiating event. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
DA-C9 Data Analysis – Run-time Estimation
Closed PRA-PI-DA, Sections 4.5.5. PINGP used estimates for test run time for some data taken from Operator logs or other plant data as '…25% of the full test duration'. Discussed with PINGP and this is related to lack of good data from around pre-2003 records when improved data recording was done. This should be discussed in the notebook. Basis for Significance: Some run time data is estimated but no clear explanation of why this was necessary or the basis for using 25% was in the notebook.
The Data Analysis notebook was updated to clarify how component run-time was estimated using Operator Logs. The documentation update did not result in a change to the assumption or any Data that was collected. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-5 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
DA-D2 Data Analysis – Expert Judgment
Closed PRA-PI-DA, Sections 3.5 (Data based on Expert Judgment) states 'There were no instances in which expert judgment was used to estimate a failure mode because either plant specific or generic failure data were available. A basis for all probabilities that did not have generic data is documented in Table 7.1 and through its associated references.' Table 7.1 (Special Event Probabilities) contains some instances of use of expert judgment where the basis is not clear. For example 1VVAxxx and 2VVAxxx values in Table 7.1 are identified as "Based on conversation with system engineering this value is conservative, however, updated basis is needed." Also, the basis for Data Analysis Notebook (PRA-PI-DA) Section 1.6, Assumption 1, 'conservatively assumed that a component would not be run for more than 25% of the full test duration if the exact time were not listed' is not clear. Basis for Significance: Table 7.1 (Special Event Probabilities) contains instances of use of expert judgment where basis is not clear. Need to at least document the rationale behind the choice of parameter values in these cases. Also, basis for DA Notebook, Section 1.6, Assumption 1 is not clear.
The Peer Review Finding was issued because the data notebook was unclear how Expert Judgment was utilized in the Data Analysis. There were two specific issues that were identified: The first issue deals with Special Events in the Data Analysis notebook (Table 7.1) because two events were identified as using expert judgment, but sufficient basis for the use of expert judgment was not provided and it contradicted with a statement in another section of the Data Analysis notebook. The second issue that was identified by the Peer Review team was an inadequate basis for assumption 1.6(1). Since this issue was documented as a separate Peer Review Finding (SR DA-C9), it will be discussed further below. The Data Analysis was reviewed for instances where expert judgment was used. Additional justification was provided for the two events identified by the Peer Review Team as well as another event that were identified as not having sufficient supporting information. In the cases where the value changed as a result of providing updated justification, the PRA model was also updated. The updated values had no significant affect on the final result. Also, the Data Analysis Notebook was updated to more clearly discuss the use of expert judgment. The specific issue identified in the Peer Review
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-6 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
DA-D3 Data Analysis – Error Factors
Closed PRA-PI-DA. All basic events for failure modes, CCF and unavailability have mean, distribution and uncertainty values assigned. The significant basic events had Bayes updated uncertainty values. Section 6.1.2 has CCF uncertainty. Unavailability assumes an Error Factor (EF) of 3 but no basis is provided. Basis for Significance: Unavailability assumes EF of 3 but no basis is provided.
The Data Analysis notebook was updated to provide justification for using an EF of 3 using maintenance data provided in NUREG/CR-6928. The model was also reviewed and updated to ensure all maintenance events were using the correct EF. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
DA-D4 Data Analysis – Reasonableness Check
Closed PRA-PI-DA, Sections 4.7 documents the reasonableness check. Did not find documentation of the identified inconsistencies and their disposition. Basis for Significance: Identification and disposition of inconsistencies were not documented.
The Data Notebook provides a narrative that describes the methodology and the results of the consistency check. To clarify the Data Analysis notebook, the ASME Standard Roadmap was updated to point the reader to the appropriate sections of the notebook that discuss the methodology and results of the reasonableness checks. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-7 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
DA-D8 Data Analysis – Modifications
Closed Section 4.5.3 (Identify Equipment Failures) contains the statement 'For instances were a modification or change to an operation practice made past data no longer representative of current performance, the use of such data was evaluated to determine if it was appropriate to use in the data analysis. If the data was not consistent with current plant design and operating practices, the use of the data was limited or not used.' The replacement of steam generators for Unit 1 is an example where PRA model data was updated. Sump screen replacements are also reflected in revised model data. The SR roadmap for this SR needs work. Basis for Significance: The SR roadmap points to Section 5.1 of the notebook (Maintenance Data), which is not relevant to this subject. The SR roadmap comment for this SR was "There were no modifications identified that would have affected the data analysis during the specified time interval for Risk Significant components or maintenance unavailability." Were there no plant modifications for the data analysis time period 1/1/2002 to 12/31/2007 such as sump screens? Were there no significant procedure changes, such as "water management" for the sump screen issue?
The Data Analysis notebook accounts for the modifications that occurred within the analysis timeframe. For the data collection period that was used in the current notebook, there were no additional modifications identified that affected the data analysis, with the exception of the sump strainer modification, which was accounted for in the PRA. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-8 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
DA-E3 Data Analysis - References
Closed No section related to this point. Must be added to only reference Uncert notebook. Basis for Significance: The sources of model uncertainty and related assumptions are not documented in this notebook.
Updated the Data Notebook to reference the Uncertainty Analysis notebook. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
HR-D2 Human Reliability Analysis – Pre-Initiator Human Error Probabilities
Closed Detailed assessments were used to quantify most of the pre-initiator HEPs. However, a screening value (that is relatively small) was used for risk significant HFEs - for example, 1EOPMDAFWRZ in Table 4-1 of PRA-PI-HR. Does not meet CC II because detailed assessments were used to quantify many but not all of the risk significant pre-initiator HEPs. For example a screening value (that is relatively small) was used for risk significant HFEs - 1EOPMDAFWRZ in Table 4-1 of PRA-PI-HR. Basis for Significance: Screening values (although relatively small) were used for risk significant HFEs - for example, 1EOPMDAFWRZ in Table 4-1 of PRA-PI-HR.
All risk-significant pre-initiator operator actions were updated with detailed Human Error Probabilities (HEPs) assessments. HRA Notebook and Database were updated accordingly. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-9 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
HR-G1 Human Reliability Analysis – References
Closed The Human Reliability Analysis Notebook Table 5-3 lists the RAW and FV values for all the post-initiator HFE basic events. Basic Events (BEs) with RAW > 2 or FV > 0.005 are listed as risk significant. Table 5-1 lists the method used to analyze the HFEs. The only risk significant HFEs that use screening values use a value of 1.0. All other risk significant HFEs use a detailed method. Basis for Significance: No reference to which model was used to calculate the importance measures for the post-initiator HFEs.
Provided reference for the current PRA Model revision used to calculate the importance measures for post-initiator Human Failure Events (HFEs) in the Human Reliability Analysis (HRA) Notebook. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
IE-C10
Initiating Events – Mission Time
Closed SR Not Met. PRA-PI-IE-INITS Initiating Events Notebook, Section 4.4.8. Some events in I-LOCL tree have 32 hour mission time (see event 0SPCHZXSCCR). Also 0SE121RFESR events used 32 hours. Why not use 24 hours? Need to confirm fault trees. Basis for Significance: Initiator should use MTTR for standby failures and other system initiator fault trees used 24 hours.
The Support System Initiating Event fault trees were reviewed and the mean time to repair mission time for the standby failures was updated to 24 hours. The Data Analysis Notebook was updated to reflect a mission time of 24 hours for standby components. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-10 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
IE-C14
Initiating Events – Interfacing System Loss of Coolant Accident Mission Time
Closed SR Not Met. RA-PI-IE-INITS Model Initiating Events Notebook, Section 4.3.7, Appendix 5. ISLOCA fault trees are quantified in the single top model CDF and LERF. Section A5.3, 2. Mission time for MOVs and check valves states, 'both valves are assumed to have a mission time for valve rupture of 8760 hours'. This results in cutsets for ISLOCA that have more than one event with a mission time of 8760 hours which is not correct. Only one event should have the 8760 mission time. Basis for Significance: The ISLOCA cutsets have multiple events with longer than the annual frequency.
The Initiating Events notebook was updated to document the justification for using 8760 hours for in-series valve failures for the ISLOCA modeling. Since there is a potential for the valve nearest the Reactor Coolant System (RCS) to leak prior to rupture, thereby exposing the inner valve to the RCS pressure for much of the operating cycle, an 8760 hour mission time was conservatively applied to each valve in-series. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
LE-C3 Large Early Release Frequency – Significant Accident Sequence Progressions
Closed Did not observe a review of significant accident progressions documented in the LERF Notebook (PRAPI- LE, Rev. 0) to determine if repair can be credited. The LERF Notebook Section 4.9.2 (SBO Sequences) includes a statement that "Because there are no PI procedures for recovering power after the batteries are depleted, the assumption is made that power cannot be recovered and that all sequences will progress to vessel breach." This SR is assessed as met at CC I. Basis for Significance: Did not observe a review of significant accident progressions documented in the LERF Notebook (PRA-PI-LE, Rev. 0) to
A section was added to the LERF Notebook to document the review of significant accident progressions that lead to a Large Early Release to determine if repair of equipment could be credited. The review looked at accident progressions such as Interfacing Systems Loss of Coolant Accidnet (ISLOCA), Medium LOCA, Station Blackout (SBO), and Steam Generator Tube Rupture (SGTR). Discussion of the recovery potential of each scenario, and the basis for not considering in the LERF, was added to the LERF notebook. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-11 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
determine if repair can be credited.
LE-F1 Large Early Release Frequency – Dominant Contributors
Closed Relative contribution to LERF from plant damage states and initiating events were provided. LERF contributor listing from Table 2-2.8-9 is not complete. Basis for Significance: LERF contributor listing from Table 2-2.8-9 is not complete.
The LERF notebook analyzed the list of LERF Contributors from WCAP-16341 because it provided a broader band of evaluated LERF Contributors than the ASME Standard. Per the review of the list provided in the WCAP, all of the LERF contributors listed in the ASME Standard are accounted for and the intent of the ASME Standard is met. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
LE-G5
Large Early Release Frequency – Limitations
Closed SR Not Met. No description and discussion of LERF model limitations. Basis for Significance: No description of limitations of the model and their implications for Risk-Informed applications.
The LERF Notebook was modified to include a technical discussion on the limitations of only considering the large early containment release scenarios needed for LERF quantification. Also included is a discussion of the fact that, since longer term containment releases are not modeled in the PINGP PRA, systems that play a role in controlling and mitigating long-term containment releases (Containment Spray (CS) and Containment Fan Coil Units (CFCUs)) are not modeled. An Appendix was added to the LERF Notebook to document sensitivity studies performed to demonstrate that modeling successful operation or failure of CFCUs and CS will not result in reducing the release magnitude below LERF categorization or cause non-LERF sequences to become LERF due to increased pressure from hydrogen burns from CS or CFCUs de-inerting the post accident containment steam environment.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-12 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
MU-F1 Maintenance and Update – Procedure Implementation
Closed It appears that some of the FP-PE-PRA procedures that are in "Draft" form are already being implemented. Basis for Significance: Some of the FP-PE-PRA procedures are in "Draft" form.
This Peer Review Finding was written since some of the Peer Reviewed procedures were not officially approved at the time of the Peer Review. All the PRA procedures have been approved and are being used by the PRA Program staff. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
QU-A3
Quantification – Interfacing System Loss of Coolant Accident
Closed PRA-PI-QU, Sections 4.1 and 4.2.6. The UNCERT code was used to determine the distribution of the CDF and LERF. The PRA-PI-INITS notebook, section A5.3 discusses how the state of knowledge correlation was evaluated and adjustments made for some ISLOCA events for valves. However, PINGP self identified that these corrected values are not included in the overall CDF/LERF frequency and should have been. However, there should be a reference in QU to the INITS section. Error identified in section 4.2.6 for Unit 1 CDF of 1.46E-5 vs. Figure 4.2-9 value needs to be corrected. Basis for Significance:
Revised ISLOCA correlated valve failure probability calculation in the Initiating Event Notebook to recognize the updated check valve and motor operated valve catastrophic rupture (large internal leakage) failure probabilities in the Data Notebook. The quantification results were updated to reflect the incorporation of this model change. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-13 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
There is no reference to the INITS notebook and PINGP identified that the ISLOCA basic events did not use the adjusted values in the final model. The value of the unit 1 CDF from UNCERT does not agree with that shown in the figure.
QU-C2
Quantification – Human Failure Event Dependency Analysis
Closed SR Not Met. PRA-PI-QU, Section 4.3.4 discusses the process used to adjust multiple HFEs using HRA calculator, EXCEL spreadsheets and utility programs. PRA-PI-HRA, section 3.4.2 and Attachment E address the dependent HFE analysis and resultant values which were used in the final quantification. This should be referenced in the QU notebook. These are included in Appendix F in the HEPCombos.txt file. There is no listing provided of the [item] and no discussion of the details of the adjustments made to the dependent HFEs to justify that the combination HFEs in any cases have extremely low values well below 1E-05. These low values and their impact on CDF/LERF should be justified and evaluated. Basis for Significance: The combination HFEs in many cases have extremely low values well below 1E-05. These low values should be justified and their impact on CDF/LERF evaluated.
This issue was resolved by specifying a minimum value for individual HEPs and to joint dependent HEPs in the dependency analysis. The details of how these adjustments were made were documented in the Quantification Notebook. The information was also included in the HRA Notebook. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-14 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
QU-D4 Quantification – Plant Comparison
Closed Section 4.3.6 gives comparison of PINGP PRA results to Point Beach, Ginna, and Kewaunee plants. The causes for significant differences are only assumed but no evidences are given. Causes for differences are not identified. Basis for Significance: The causes for significant differences are only assumed but no evidences are given. Causes for differences are not identified.
The Quantification notebook discusses the differences between Prairie Island Nuclear Generating Plant (PINGP) and other similar 2-LOOP Westinghouse designs. The Initiating Event contributions for each plant were compared and the differences are discussed by postulating how the PINGP design differs. It was found that some PRA modeling assumptions, in addition to actual plant operation or design differences were the basis for the differences seen in initiating event distributions. Subsequent updates to the internal event or fire PRA were not required as a result of the plant comparison The specific issue identified by this finding was closed after finalization of the Rev 3.1 internal events PRA model, which was used as a starting point for the Fire PRA. However, it was determined that this finding did not have an impact on the Fire PRA.
SY-A4 Systems Analysis – Walkdowns and Interviews
Closed Plant walkdowns and interviews have been done (see System Walkdowns and Interviews notebook). Some issues seem not to have been addressed. Basis for Significance: Some issues were found during the plant walkdowns and interviews but were not addressed in this revision. So the PRA model could be considered as non coherent with the as-build as-operated
Each comment from the system engineering interviews and plant walkdowns was reviewed to determine whether or not the issue had been resolved. Every item that was identified as not being completed was entered into the PRA Change Database (PCD) to determine its impact on the PRA. Depending on the issue’s significance, it was either completed immediately or was deferred to a future PRA model update.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-15 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
plant. However these issues are managed and become parts of the PRA maintenance and update process.
The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
SY-A8 Systems Analysis - Component Boundaries
Open The pressure switch failure for the Component Cooling pump noted in Component Cooling notebook assumption 28 for miscalibration of pressure sensors should not be included in the boundary of the Component Cooling pump. Basis for Significance: The pressure switch failure is separately modeled in CC and miscalibration is not.
The specific model changes have been incorporated into the PRA, but the Finding has not been closed out due to an extent of condition evaluation that was performed to review modeled Instrument and Control (I&C) component failure events relative to the component boundaries of the modeled equipment that they support. Similar to the Component Cooling pressure switch, other inconsistencies related to I&C components were identified. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA. Although the extent of condition reviewed similar inconsistencies, the current model is considered conservative and adequate for the risk-informed NFPA-805 application.
SY-A17 Systems Analysis – References
Closed PRA-PI-SY-CL Section 4.0 which discusses specific operator actions and applicable plant procedures. Cannot find reference to PRA-PI-HR notebook in Section 4.0 of the CL notebook. Do not find a reference in Section 4.0 of the CT notebook either.
The Cooling Water (CL) and External Circulating Water (CT) System Notebooks have been updated to include a reference to HRA Notebook. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-16 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
Basis for Significance: The system model should include HFEs that are expected during the operation of the system. No reference to PRA-PI-HR notebook in CL and CT notebooks.
starting point for the Fire PRA.
SY-B14 Systems Analysis – Containment Cooling
Closed Section 6.13 of the Success Criteria Notebook (CONTAINMENT) states: "Containment cooling is not required for any of the Level 1 accident sequences. In addition, for LERF modeling no containment cooling is required (see the LE notebook, Reference 21). Therefore, the PINGP Level 1 and LERF models do not include containment spray (CS) or [containment] fan cooler units (CFCUs)." Observed one MAAP LLOCA scenario (LLOCA-CONT) that did not credit CS or CFCUs (SC Notebook Table 6.4) but did not find documentation of the results for this case in Appendix B of the SC Notebook or elsewhere. Need documented basis for not crediting containment spray and CFCUs in the PINGP PRA model to ensure no impact on CDF and LERF results. Basis for Significance: Not crediting CS and CFCUs in the PRA may ignore their impact for some scenarios (e.g., NPSH for large LOCA recirculation mode RHR).
The following was completed to address this Peer review Finding from both a Level 1 Plus LERF perspective. (1) A sensitivity calculation was developed to evaluate the affects of Containment Fan Coil Units and Containment spray. This Thermal Hydraulic calculation confirmed that for LLOCA and MSLB scenarios the operation of the Containment Fan Cooler Units (CFCUs) or Containment Sprays (CS) will result in lower peak pressures. This calculation also confirmed that for the LLOCA and MSLB scenarios with failure of CFCUs and CS, the calculated peak containment pressures will remain below the best estimated containment failure pressure. Therefore, it was concluded that there is no need to model CFCUs or CS operation or failure in the Level 1 PRA accident sequences. (2) The Success Criteria Notebook was amended to describe the results of the sensitivity studies performed in the calculation described above. (3) An additional sensitivity calculation was performed to evaluate the effects of the Containment Fan Coil Units and Containment Spray on LERF analysis. This calculation confirmed that for all plant damage states
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-17 – Revision 1
Table U-1 Internal Events PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
previously evaluated operation/failure of the Containment Fan Cooler Units (CFCUs) or Containment Sprays (CS) would not alter the classification of LERF vs. non-LERF sequences. Therefore, it was concluded that there is no need to model CFCUs or CS operation or failure in the Level 1 Plus LERF PRA accident sequences. (4) The LERF Notebook was amended to describe the results of the sensitivity studies performed in additional sensitivity study described above. The specific issue identified in the Peer Review Finding was corrected when finalizing the Revision 3.1 PRA Model, which was used as a starting point for the Fire PRA.
Northern States Power - Minnesota Attachment U – Internal Events PRA Quality
PINGP Page U-18 – Revision 1
Table U-2 Internal Flooding PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
IFPP-B3-01 Identification of assumptions
Closed A parametric uncertainty analysis was performed on the final results. An evaluation of generic sources of model uncertainty from NUREG-1855 was documented. Although some of the plant specific sources of modeling uncertainty are documented in Table 18, not all key assumptions contained in the analysis documents are listed and evaluated for their potential effect on applications.
This issue deals with assumptions contained within the Flooding PRA Analysis. The Internal Flooding documentation has been updated to document applicable assumptions and their potential contribution to model uncertainty. The assumptions associated with this Finding are Internal Flooding related and do not have an effect on the Fire PRA.
IFPP-A2-01 Credit for sealed penetrations
Closed In defining the flood areas, assumption 6 of the accident sequence notebook states: "Sealed penetrations are assumed to be effective at preventing propagation between areas such that the propagation would result in equipment failure in the adjoining area." On page 169, the table states that for zone 419 the sealed penetration fails. This is in conflict with assumption 6.
This Finding deals with conflicting statements within the Flooding PRA Analysis regarding the propagation of a flood through a penetration. The Internal Flooding documentation has been updated to clarify the conflicting wording. This Finding does not affect the Fire PRA.
IFSN-A10-01 Credit for floor drains
Closed The assumption states that no credit was taken for drains. However, the accident analyses and initiating event definitions discriminate based on flooding flow rates and area drain flows. As a result, there is a conflict between the documented analyses and Assumption 2 of the flood area definition notebook and the accident sequence analysis.
This Finding deals with credit for drains in the Flooding PRA Analysis. The Internal Flooding documentation has been updated to discuss how flood drains are credited. This Finding does not have an effect on the Fire PRA.
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Table U-2 Internal Flooding PRA Peer Review – Facts and Observations
SR Topic Status Finding/Observation Disposition
IFSO-A1-01 Explanation of potential flooding source
Closed The heating steam pipe is not considered a flooding source but there is no explanation for this.
This Finding deals with heating steam piping as a flooding source in the Internal Events Flooding Analysis. The Internal Flooding documentation has been updated to clarify why the heating steam pipe is not considered a flooding source. This Peer Finding has no effect on the Fire PRA or any of its postulated initiators.
IFQU-A6-01 Application of internal events HFE’s
Closed There is no evidence that all the applicable HFE's from the internal events model were reviewed to see how they were affected by flood scenarios.
This Finding is related to how internal events HFEs are applicable in the Flooding PRA analysis. To resolve this Finding a review of the HFE’s credited in the Internal Flooding Analysis was performed and documented. No changes to the HRA were required. This Finding does not have an effect on the Fire PRA Analysis. For the Fire PRA, each credited HFE was reviewed and modeled using industry accepted methods (i.e., NUREG 1921).
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V. FIRE PRA QUALITY
39 Pages Attached
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In accordance with RG 1.205, Regulatory Position 4.3: The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard “supporting requirements” important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable. Licensees should justify use of Capability Category I for specific supporting requirements in their NFPA 805 risk assessments, if they contend that it is adequate for the application. Licensees should also evaluate whether portions of the PRA need to meet Capability Category III, as described in the PRA Standard.
The information presented in this attachment addresses the documentation requirements set forth in Regulatory Guide 1.200, Rev. 2. A Peer Review of the PINGP Fire PRA using RG 1.200, Rev. 2, including the NRC clarifications and qualifications, and Section 4 of ASME/ANS Standard RA-Sa-2009 was conducted the week of May 7 through May 11, 2012. Based on this peer review, the PINGP Fire PRA was found to be essentially consistent with the ASME/ANS PRA Standard. Section 4 of the ASME/ANS PRA Standard contains a total of 183 Supporting Requirements (SRs) under 13 technical elements, and configuration control from Section 1.5. Of these 183 SRs, eighteen (18) were determined to be not applicable to the PINGP Fire PRA either due to the fact that the requirements were not applicable to the PINGP approach, or the technical element was not used for the PINGP analysis (i.e., Quantitative Screening, QNS). The remaining SRs were evaluated against the ASME/ANS PRA Standard and were assessed as meeting Capability Category (CC) I, II, III, or “Not Met.” Per Regulatory Guide 1.205, meeting the RG 1.200, Rev. 2, and ASME/ANS-RA-Sa-2009 standard at Capability Category II represents a degree of accuracy such that the PRA may be used for NFPA 805 transition. Capability Category III represents a higher degree of accuracy that exceeds CC II, and so is also acceptable for regulatory submittals. Meeting a requirement of the standard at Capability Category I represents a lower degree of accuracy, and must be shown to be adequate for the application on an application-specific basis. For the PINGP Fire PRA, 92% of the SRs were assessed at Capability Category II or higher, including 5% of the SRs being assessed at Capability Category III. The PINGP Fire PRA had an additional 3% of the applicable SRs assessed at the CC-I level. The PINGP Fire PRA did not meet 5% of the applicable SRs. There were no SRs “Not Reviewed” by the Peer Review Team. There were also no “Unreviewed Analysis Methods” identified by the Team. The Peer Review team also noted a total of 56 Facts and Observations (F&Os). These included fifteen (15) “Suggestions,” forty (40) “Findings” and one (1) “Best Practice.” The Finding F&Os covered a variety of topics, but many dealt with the need to incorporate additional detailed analyses to develop results that are more realistic rather
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than bounding. Several others were on the need to better identify assumptions and discuss their impact on overall results. The Best Practice F&O was issued for the Seismic Fire Interaction Technical Element. Subsequent to the 2012 full-scope Fire PRA Peer Review, several changes were made to the PINGP Fire PRA. All changes were reviewed and, as a result of this review, most were determined to be Maintenance. One change was considered to be an Upgrade since it constitutes the application of a new method to the PINGP Fire PRA (new to the PINGP Fire PRA only, and NOT a new industry method). Prior to the full-scope Fire PRA Peer Review, the PINGP Fire PRA used the Beyler correlation to estimate the hot gas layer (HGL). After further review, PINGP decided that the Beyler correlation was not the most appropriate method to use. As such, PINGP updated the fire modeling to use the McCaffrey, Quintiere, and Harkleroad (MQH) correlation for rooms in which only natural ventilation would occur, and the Foote, Pagni, and Alvares (FPA) correlation for rooms which have mechanical ventilation. Therefore, a Focused-Scope Fire PRA Peer Review of the PINGP Fire PRA was conducted against the requirements of RG 1.200, Rev. 2, including the NRC clarifications and qualifications and Section 4 of ASME/ANS Standard RA-Sa-2009. This review was conducted during the week of November 5, 2013. The scope of this Focused-Scope Peer Review was to review the upgrade from the Beyler correlation to the MQH / FPA correlations for use in determining resultant HGL temperature levels. Based on this Focused-Scope Fire PRA Peer Review, the PINGP Fire PRA was found to remain essentially consistent with RG 1.200, Rev. 2 and the ASME/ANS PRA Standard. All 183 SRs were reviewed by the Focused-Scope Fire PRA Peer Review Team, who made the determination of which SRs should be assessed. Of these 183 SRs, twelve (12) were determined to be applicable to the HGL scope of review. These twelve SRs were evaluated against the ASME/ANS PRA Standard (including any NRC clarifications and qualifications) and were assessed as meeting Capability Category (CC) I, II, III, or “Not Met.” All of the reviewed SRs were assessed at Capability Category II or higher. There were no “Unreviewed Analysis Methods” identified by the Team. The Focused-Scope Fire PRA Peer Review Team also noted a total of 5 F&Os. These included four (4) “Suggestions” and one (1) “Finding.” The F&Os primarily addressed the need to justify the selection of the specific HGL correlations used. For both the Fire PRA Peer Review and the Focused Scope Peer review, the Finding F&Os and their disposition with respect to the NFPA 805 License Amendment Request are provided in Table V-1, organized by Technical Element and Supporting Requirement. The Status column in Table V-1 indicates whether Findings are open or closed. Findings are considered Open until associated documentation updates have been completed and accepted into the NSPM process. In all cases, an explanation is provided in the “Disposition” column. The “Issues” in Table V-1 are presented directly from the Peer Review report, with minor editorial corrections.
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For both the Fire PRA Peer Review and the Focused Scope Peer review,the PINGP Fire PRA was determined by the Peer Review Team to meet Capability Category II in most but not all cases. A limited number of ASME/ANS Standard areas were identified as either not meeting a Supporting Requirement or only meeting Capability Category I requirements. The impact of those areas where a requirement was judged to be “Not Met” or where only the Capability Category I requirement was met is evaluated in Table V-2. These are listed in Table V-2 with the current status and current self-assessed capability category. The full PINGP Fire PRA Peer Review Report and Focused Scope Fire PRA Peer Review Report will be made available to the NRC staff upon request. There have been a number of refinements made to the Fire PRA model since the last full scope peer review. NSPM has reviewed each of these refinements in accordance with our PRA update procedure, and has determined that (with the exception of the HGL change discussed above) all of these refinements fit the definition of PRA maintenance. As discussed above, the one upgrade was subjected to a focused-scope peer review. There have been no other Fire PRA upgrades since the last full-scope peer review. All changes to the Fire PRA Model are evaluated in accordance with established NSPM PRA update procedures. A peer review is initiated if a change to the model is classified as a PRA upgrade. Tables V-1 and V-2 have been significantly changed in Revision 1 to describe changes in Status and Disposition of F&Os. Due to the extensive nature of these changes, sidebar markings to designate changed text are not included.
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
PP-C3-01 Other Affected SR PP-C1
Agreement between the Plant Walkdown Notes and the Plant Partitioning Notebook Appendix
Closed DOCUMENT the general nature and key or unique features of the partitioning elements that define each physical analysis unit defined in plant partitioning in a manner that facilitates Fire PRA applications, upgrades, and peer review. The information in the Appendix A, PINGP Compartment Table, Basis for fire area separation, does not always agree with the information contained in the walkdown sheets. An example is on walkdown sheet 4GRP which states "unsealed cable tray penetration between FC39 and 4GRP. Likely to spread fire due to continuity of combustibles". The PINGP Compartment table states, "Unsealed cable tray penetration FC 39. Unsealed piping penetration to FC 39 and 85. These penetrations aside, the barriers are still sufficient to contain a fire." These two descriptions need to agree and provide reasons for the plant partitioning. Appendix A and the walkdown sheets need to be reviewed. These two documents need to be revised so they agree. The term "active fire door" needs to be revised to a door activated by a fusible link. A fusible link is considered to be a passive device and not an active device. It is acceptable to use the active definition for these fire barriers, but if that does occur, then PP-B5 changes from "N/A" to "CC-I". To go from "CC-I” to "CC-II" PINGP needed to "define and justify" the criteria applied for crediting the active fire barriers in the partitioning report.
This finding discusses a discrepancy between the original walkdown notes and the final fire area separation table that are documented in the Plant Partitioning notebook, FPRA-PI-PP. The finding identifies that the separation between two of the identified Fire Compartments (FC) was not consistent. A review of Appendix A and D was performed to ensure the information from Appendix D was adequately transferred to Appendix A. The notebook updates address the differences between the appendices and clarify the peer reviewer’s comments. This change was made in Revision 0 of FPRA-PI-PP.
ES-C1-01 Other Affected SR CS-A1
Add instruments required for new HFEs to the ES Notebook
Closed HFEs created specifically for Fire Scenarios (identified in FPRA-PI-FHRA) and their credited instrumentation have not been included in the Equipment Selection documentation. Related to this, not all instrumentation for those HFEs are cable selected. Standard requires that all HFEs and associated instrumentation required for FPRA are identified in the Equipment Selection task (and it is noted that Equipment Selection is an iterative process). Also, associated instrumentation requires cable selection if cable routing is not available.
The credited instrumentation for the human failure events in the Fire PRA model are identified in Appendix D of FPRA-PI-ES (Revision 1). Appendix D was developed after the credited instrumentation was incorporated in the CAFTA logic model. Cables were selected and mapped to the corresponding instrumentation in support of the Fire PRA quantification process. The mapping of cables to the instruments is maintained in FRANX and the Genesis
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
Include all HFEs in the Equipment Selection documentation. Cable selection will be required on instrumentation if not already available. It has been noted that there is a revision plan in place to address instrumentation currently not cable selected in the HRA.
database. The human failure events for which instruments were selected are also listed in Appendix D of FPRA-PI-ES.
CS-A10-01 Cable routing through compartments
Closed Open item No. 1 on Page 17 of 17 of FPRA-PI-CS, Revision C, states that "in order to fully comply with Capability Category II, cables that are routed through fire compartments 2A, 41B-1, 46A, 58A, 58B, 58C, 58D, 76A, 78E, 86, 94A, 94B, 94C, 94D, 94E, and 94F need to be identified. This will be accomplished by identifying routing on electrical drawings, and with walkdowns performed as needed." Since this has not yet been completed, and the methodology not specified, this SR is provisionally met. Completeness of Fire PRA. Route cables through the listed fire compartments, and perform walkdowns to confirm accuracy of the routing. Utilize EPM Division Procedure EPM-DP-EP-005, Revision 1-Post-Fire Safe Shutdown Cable Routing and Component Location, and EPM Division Procedure EPM-DP-EP-004, Revision 2-Post Fire Safe Shutdown Cable Identification-February 2011.
This finding is related to fire compartments that are subsets of fire areas in which all cables in the associated fire area were assumed failed even if they were not in the fire compartment (because cables are routed by fire area in the cable database). As part of the detailed fire modeling documented in FPRA-PI-SCA, risk contributing fire compartments identified through the development of the Fire PRA have been subdivided in to Fire Scenarios and the targets have been assigned based on walkdowns and drawing inspections. Only those conduits with unknown routing have been mapped to all the fire scenarios within the fire compartment. This process removed the excess conservatism identified in this finding.
CS-B1-01 Other Affected SR CS-C4
Breaker coordination study
Closed As identified in FPRA-PI-CS, Appendix R overcurrent coordination and protection analysis has been reviewed but the analysis of additional circuits identified during the Fire PRA is currently in progress and is therefore not complete. Appendix R Breaker Coordination study has been reviewed. Additional breaker coordination is still being performed. Finish breaker coordination study.
Breaker fuse coordination studies (PRA Calculation V.SPA.12.016 and PRA Calculation V.SPA.12.018) were conducted and 22 Fire PRA credited power supplies were identified that currently do not coordinate but will be modified to achieve coordination. The Fire PRA assumes that there are not any coordination issues with the power supplies that are credited in the Fire PRA model. Resolution of coordination issues is captured in Attachment S of the LAR.
PRM-A1-01 Screening of ignition Closed This SR states "CONSTRUCT the Fire PRA plant response model so that it is capable of determining
The fire modeling that supports the Fire PRA quantification process has been re-structured
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
sources fire-initiated conditional core damage probabilities (CCDPs) and conditional large early release probabilities (CLERPs) for various fire scenarios." The Fire PRA model is developed so that it is capable of determining fire-initiated CCDPs and LERPs for most ignition sources/scenarios. However, a large number of ignition sources have been "screened" in the Fire Compartment Analysis based on the fact that a fire associated with them only has the potential to impact the equipment that is the ignition source. A basic event showing the ignition source was screened as "ANDed" with the ignition source in the Fire PRA model. This does not appear to be correct modeling for ALL components that were screened since it does not include consideration of whether or not the equipment failure will result in a reactor trip. Verify, and where applicable state, that the "screened" ignition sources do not result in a reactor trip, and that it is accurate to eliminate them as initiators. For those "screened" sources that can result in a reactor trip, they need to be retained as initiators with their target sets including the equipment itself only.
so that the contributions of ignition sources that do not propagate fires are included as impacts. The basic events associated with the ignition source have been included in the Zone to Raceway table in FRANX to ensure that a CCDP and consequently a CDF and LERF scenarios are calculated. This process is documented in the FPRA-PI-SCA notebook (Revision 1). It should be noted that this treatment applies also to scenarios where the fire propagates outside the ignition source. In those situations, the basic events associated with the ignition source are included in the Zone to Raceway table as Fire PRA targets.
PRM-A1-02 Loss of instrument air piping integrity
Closed This SR states "CONSTRUCT the Fire PRA plant response model so that it is capable of determining fire-initiated conditional core damage probabilities (CCDPs) and conditional large early release probabilities (CLERPs) for various fire scenarios." PINGP uses soldered connections in their instrument air system. They acknowledge that fires can result in failure of the soldered connections. However, PINGP contends that failure of a soldered connection will not lead to failure of the instrument air system because they assume that the second air compressor would make up for the lost air. However, this assumption appears to overlook the condition where the failed solder joint will lead to separation of the line. Any or all equipment downstream of the failed solder joint will lose air and go to their loss-of-air position. Furthermore, with an open line, the air flow would be preferentially directed to the line break. Depending
The instrument air system is no longer credited in the Fire PRA, except in a limited number of fire scenarios in the Relay Room (FA 18). Credit for instrument air was determined based on drawing review and plant walkdowns to determine size and configuration of piping within the rooms and location of solder joints. It was determined that the piping of this size runs through only a small portion of the room. Other lengths of piping within the room is of a smaller size that, if severed in a fire, will not result in system failure. Therefore, only a portion of the fire scenarios in this room are assumed to fail instrument air. The Relay Room scenarios in which the Instrument Air system
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
on the break size, this bypass flow may be sufficient to lead to a significant drop in system pressure. This could lead to all air-operated components working off the system going to their loss-of-air position. PINGP needs to re-evaluate the impact of a fire-induced loss of instrument air considering the two scenarios above. For the second scenario, a calculation demonstrating that the two compressors are capable of maintaining adequate system air pressure even given severing of the largest line with soldered connections.
function is credited are identified in the Fire Scenario Selection (FSS) analysis (FPRA-PI-SCA, Revision 1, Attachment G).
PRM-A1-03 Other Affected SRs FQ-D1, FQ-E1
Containment bypass due to multiple spurious operation of isolation valves in small lines
Closed This SR states "CONSTRUCT the Fire PRA plant response model so that it is capable of determining fire-initiated conditional core damage probabilities (CCDPs) and conditional large early release probabilities (CLERPs) for various fire scenarios." When the Containment bypass lines were evaluated for potential inclusion of additional MSOs due to fires, one of the criteria that appears to have been used to screen lines was the size of the line. If the line was not >1 inch (water) or >2 inch (steam), it appears to have been screened out. Although this may be appropriate for internal events, it is not appropriate for fire events since a fire can cause multiple failures that result in multiple smaller lines being impacted simultaneously such that the LERF criteria is met. Re-evaluate the potential Containment Bypass lines and ensure that no lines are screened using the size criteria, or provide a justification for why the 1 inch (water) and 2 inch (steam) screening criteria is appropriate for fire-analyses.
This finding suggests that multiple small flow diversions through lines that would otherwise have been screened out should be considered as a potential path for containment bypass. The probability of multiple small flow diversions being created simultaneously by spurious operations is very small. SR ES-A6 states for Capability Category II: “CONSIDER up to three spurious actuations of equipment alone or in combination with other fire-induced loss of function failures for the special case where fire-induced failures could contribute to an initiating event that in turn leads to core damage and a large early release.” Note 8 for this SR states: “Fire-induced failures leading to interfacing system loss-of-coolant accident (ISLOCA) or containment bypass are examples of cases where fire-induced failures could contribute to an initiating event that in turn leads to core damage and large early release.” To ensure this criterion was met, a search was performed for cases where up to three spurious operations within small lines that
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Finding F&O Topic Status Issue Disposition
have been screened could lead to containment bypass events. No such spurious operation combinations were identified (FPRA-PI-ES, Revision 1, Section 5.3.6).
PRM-A2-01 Other Affected SRs PRM-A3, FQ-A4
Fire scenario analysis inputs
Closed This SR states: CONSTRUCT the Fire PRA plant response model so that it is capable of determining fire-initiated core damage frequencies (CDFs) and fire-initiated large early release frequencies (LERFs) once the fire frequencies (see Section 4.2.7) are also applied to the quantification. Although the scenario specific ignition sources, non-suppression probabilities, and severity factors are included directly in the Fire PRA model on a scenario specific basis, the basic events and probabilities included in the model do not appear to match what is stated in the notebooks. For example, looking at FC 32, the notebook states that the Non-Suppression Probability is based on a wet pipe system, and basic events 0SUPR-----M and 0SWET-----F. These basic events are modeled correctly under gate SUP-WET-G, but the probabilities used are incorrect. The fault tree has probabilities of 1E-2 and 3E-2 respectively, and the document shows probabilities of 2.2E-3 and 1E-2 respectively. These two need to be consistent Additionally, compartment documentation describes how the various damage states are determined. However, the documentation does not clearly state how these damage states are translated into the Plant Response Model fault trees. This is leading to confusion when attempting to validate that the PRM logic is correct. Create a table of the inputs for each scenario to ensure that the inputs into the PRM are complete and correct. It is recommended that this table include the scenario, ignition frequency, NSP [Non-Suppression
The PRM fault tree models have been simplified and de-coupled analytically from the FSS analysis, FPRA-PI-SCA notebook. This was done by removing the fire initiating event logic from the merged PRM fault tree (which previously included not only the ignition frequencies, but also severity factors, frequency modifiers, and non- suppression probabilities). Plant damage states are now determined dynamically through solution of the PRM event tree-fault tree linking structure. This structure determines whether the scenario should be quantified as a transient, very small LOCA, etc. based on the equipment that has been impacted in the specific fire scenario. The fire scenarios are defined in the FSS analysis, FPRA-PI-SCA notebook, including the ignition frequencies, severity factors, non-suppression, etc. These inputs are no longer in the CAFTA model and are imported as part of the scenario frequency into the FRANX quantification. These inputs are summarized and tabulated in the SCA notebook (FPRA-PI-SCA, Revision 1).
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
Probability] (including BE names), and Severity Factor (including BE names)) so that it is easy to ensure the logic reflects the notebook.
PRM-B2-01 Plant Response Model – Internal Events Peer Review Findings
Closed VERIFY the peer review exceptions and deficiencies for the Internal Events PRA are dispositioned, and the disposition does not adversely affect the development of the Fire PRA plant response model. Not all the F&O's were addressed in Appendix F of PINGP-FPRA-PI-PRM notebook. It is not clear that all the suggestions have also been addressed. In its current form, what is in the Appendix F does [not] appear to be complete, but once all F&Os have been added, the table appears sufficient. Completion of the Appendix F and addressing the suggestion F&O for the internal events peer review will met this SR.
A review of internal events peer review Findings and Suggestions was performed to determine if there were any effects on the Fire PRA. The results of the review showed that none of the Findings or Suggestions remaining open impacted the Fire PRA. Attachment U of the LAR provides detailed information regarding the peer review Findings that affect the internal events PRA. All suggestions are tracked in the PINGP PRA Change Database (PCD), which is the official configuration control repository for PRA changes.
PRM-B13-02 Other Affected SR PRM-B12
Naming convention for fire scenario analysis parameters
Closed There is a requirement to develop and use a system model nomenclature to allow model manipulation in the Internal Events PRA (SY-A23), and PRM-B9 states "ALL the SRs under HLR-SY-A and HLR-SY-B in Part 2 are to be addressed in the context of fire scenarios" - therefore the naming scheme established under SY-A23 of Part 2 is also applicable to the Fire PRA. Spurious transfers of valves were added into the model, but did not use the naming scheme established in the Internal Events PRA model. This implies that these events were meant to be used solely as Flag events. If this is true, then the naming scheme for Flag events needs to be used. If the addition of these events was done to include them as spurious failures as well, then data needs to be established for them in accordance with the Data SRs. If these events are to be used as Flag events, change their names to use the established Flag naming scheme. If they are to be used as random failure events, either use the already established "transfers open/closed" type code and data, or develop
When developing the Fire PRA from the existing internal events PRA models, it was often the case that spurious operation basic events (fail to remain open, fail to remain closed, etc.) already existed in the appropriate locations in the model for the Fire PRA. In these cases, cables that could fail in such a way during a fire as to cause the spurious operation were mapped to this basic event. Therefore, during quantification of the fire PRA in FRANX, the event probability for an event of this type either: 1) retains its random failure probability (cables not impacted by the fire), or 2) is set to a higher probability based on Circuit Failure Mode Likelihood Analysis (CFMLA), or 3) is set to 1.0 (TRUE) (if CFMLA is not performed). In other cases (for example, in some MSO logic modeling cases), spurious valve
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
appropriate Type Codes and supporting data for them.
operation events were not already modeled in the internal events model, so additional flag events were included to model failure of equipment due to the spurious operation. The fact that they do not follow the naming convention for other types of basic events does not affect the results of the PRA because when these events are used (i.e., due to fire-induced cable failure), their probabilities are set to either the CFMLA probability value or to 1.0 (TRUE) based on instructions in the FRANX database. The event probabilities are therefore directly assigned at run-time and are not based on use of a standard CAFTA calculation type/type code scheme. The Fire PRA results and the change in risk calculated for the Fire Risk Evaluations are therefore performed correctly, regardless of the event naming.
PRM-C1-01 Fire compartment analysis documentation
Closed This SR is associated with documenting the Fire PRA plant response model in a manner consistent with the document requirements of the Internal Events PRA SRs for IE, AS, SC, SY, and DA. The documentation requirements are developed to ensure that the development of the Fire PRA model is easily understood and reproducible, and that they facilitate PRA applications, upgrades, and peer reviews. Although most of the information appears to be available throughout the various notebooks, it is difficult to find, and verify. Therefore, it is difficult to ensure that the Fire PRA model accurately reflects the information currently contained in the compartment specific notebooks, and in many instances it was determined that the two were inconsistent. A summary section that contains a table of the scenarios, and their applicable ignition frequencies, NSPs, and severity factors would be very useful, and would help ensure that the PRM accurately reflects
The PRM fault tree models have been simplified and de-coupled analytically from the FSS. This was done by removing the fire initiating event logic from the merged PRM fault tree (which previously included not only the ignition frequencies, but also severity factors, frequency modifiers, and non-suppression probabilities). Plant damage states are now determined dynamically through solution of the PRM event tree-fault tree linking structure. This structure determines whether the scenario should be quantified as a transient, very small LOCA, etc. based on the equipment that has been impacted in the specific fire scenario. The fire scenarios are defined in the FSS analysis, including the ignition frequencies, severity factors, non-suppression, etc. These inputs are summarized and tabulated in the
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Table V-1 Fire PRA Peer Review – Facts and Observations
Finding F&O Topic Status Issue Disposition
the latest compartment information. However, this is not the only option. Any documentation restructuring or enhancement that would make it easier to follow the process employed for identifying the required PRM changes, and to find the required changes with their bases would be acceptable.
SCA notebook (FPRA-PI-SCA, Revision 1).
FSS-A5-01 Other Affected SRs FSS-A1, FSS-A2, FSS-D10
Fire scenario development and fire modeling
Closed Section 6.1.1 (Unit 1, Fire Compartment 1-Containment) of FPRA-PI-SCA, Detailed Compartment Analysis Notebook, indicates that "the contribution to the plant Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) from initiators in this compartment is sufficiently low to preclude fire modeling to screen initiators." The report FPRA-PI-FQ, Rev. B, Fire PRA Quantification, indicates that the CDF contribution from fire compartment 1 is 6% and is not sufficiently low. Therefore, additional evaluation is necessary beyond assuming full room burnout. A number of other fire compartments have the same issue. An F&O is being issued to recognize the fact that this SR must be met for the further analysis that will need to be used to ensure that the scenarios are evaluated and/or quantified at a level of detail commensurate with the risk significance of the scenarios. Additionally, IF new analysis methods (methods that are different from the ones currently being used by PINGP and that were reviewed as part of this Peer Review) are required to ensure the risk is adequately characterized, the new analysis methods may require a focused Peer Review to ensure they are applied appropriately. Perform additional ignition source and target set refinement and fire modeling to better characterize the fire risk contribution to CDF/LERF.
Detailed fire modeling has been performed for several additional fire compartments within the Fire PRA, including Unit 1 and Unit 2 Containment. Top contributing fire zones have received detailed fire modeling analysis. Table 4-1 in Section 4.1 of FPRA- PI-SCA (Revision 1) documents the level of analysis performed for each fire compartment. The results of this modeling were incorporated into overall Fire PRA quantification results (CDF/LERF) for the LAR and are provided in the Fire Modeling and Quantification Spreadsheet (Attachment N). No new analysis methods were required to address this finding; as such, a focused scope peer review is not necessary.
FSS-B2-01 Other Affected SR FSS-A6
Main Control Room (MCR) analysis
Closed Per Procedure EPM-DP-RSD-005, the MCR abandonment scenarios were developed to bound the risk in the main control room. This meets the standard at the CC-I level. Additionally, Section 9.2.1 of FPRA-PI-MCR, Main Control Room Analysis, Revision B (page 33 of 79) uses a bounding assumption
As described in Sections 7.1, 7.2, and 7.3 of FPRA-PI-MCR (Revision 1), multiple control room abandonment scenarios have been evaluated (See Figure 6). In particular, it should be noted that the postulated scenarios consider fires occurring in both fixed and
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associated with how the probability of a fire starting near a target is determined, which is acceptable for a bounding analysis, but will need to be revisited in order to ensure the MCR risk is realistically characterized. This F&O is a finding to go from CC-I to CC-II, which requires the FPRA to realistically characterize main control room abandonment fire scenarios instead of using a bounding analysis. The Standard requires an evaluation of two types of Main Control Room scenarios – those that require Full Abandonment, and those that do not lead to MCR Abandonment but do rely on ex-control room actions [i.e., external to the MCR] to safely shutdown the reactor. The scenarios evaluated do not appear to adequately consider or address Main Control Board (MCB) fires that do NOT require full abandonment, but DO require ex-control room operator actions including remote and/or alternate shutdown actions. Note – it is expected that some MCB fires will only impact a single Unit – and the need for ex-control room actions to assist that Unit – so the assumption that every fire results in a Dual Unit trip needs may need to be re-evaluated for some MCB fires. Since there is no single correct way to perform a realistic MCR analysis, it is recommended that PINGP contact other utilities to see the methods they used, and then select the methodology that best fits for PINGP.
transient ignition sources that result in the loss of specific functions due to damage to equipment and raceways found in the MCR but do NOT result in MCR abandonment. It was concluded that the various scenarios reasonably characterize the MCR abandonment risk because the probability of abandonment depends on fire-generated environmental conditions (Section 9.1). The extent of fire damage increases under abandonment conditions to include contributions of cable routing to panels (Section 6.0), Operator actions upon abandoning the control room (Section 7.0) are accounted for, and the HRA development for control room abandonment is based on operators mitigating the event from outside the control room (Section 8.0).
FSS-C1-01 Conduits located between ignition source and nearest cable tray
Closed The screening portion of fires included all fire sizes up to the nearest cable tray, but did not consider conduits between the ignition source and the cable tray that could potentially be a PRA target. This is non-conservative as it could result in screening risk contribution from PRA targets not in cable trays. Not considering conduits between the ignition source and the nearest cable tray is non-conservative as the conduit may be a PRA target. Recommend reducing the fraction of fires screened
As described in Section 4.1.9 of FPRA-PI-SCA (Revision 1), the severity factor is calculated based on the closest target (cable tray or conduit) or intervening combustible. For all ignition sources contained within those compartments where detailed fire modeling was conducted, walkdowns were performed to measure this shortest distance. Documentation of these walkdowns is provided in Attachment L of FPRA-PI-SCA. It should be noted that this approach to
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from an ignition source to those that cannot affect the nearest conduit to an ignition source, which may be a PRA target. It would be acceptable to ignore conduits that are known not to be PRA targets. For the fires beyond the screened portion, a full-room burnup would still need to be considered in the scoping fire modeling in order to capture all PRA targets, even though the conduit / raceway itself is less likely to cause an HGL than if the equipment or an associated cable tray combusted.
calculate the severity factor is conservative because the Fire PRA significance of the closest target is not considered. Instead, the severity factor is assigned assuming that the nearest cable tray, conduit, or intervening combustible is credited in the Fire PRA. Further, compartments for which a full-compartment burnout is postulated, all Fire PRA targets mapped to the compartment of interest are assumed to be damaged by a fire originating in any one of the identified ignition sources.
FSS-C5-01 Assumption for cable damage temperature
Closed Based on the wording in the reports, and a spot check of analyses, cables are assumed to be thermoset for self-ignited cable purposes, and are assumed to be thermoplastic for the purposes of being a target. The cable database has information about which type of cable each individual cable really is, however the cable database does not appear to be used when determining what properties to use in the analysis for individual cables. The cable properties impact the CDF/LERF calculated for scenarios. Utilize the cable database to determine the target cable failure situation. If assumptions are used for screening/full compartment analyses, but actual cable types are used in detailed analyses – this needs to be clearly stated in the report.
Per Assumption 1 in Section 3.0 of FPRA-PI-SCA (Revision 1), the single compartment detailed fire modeling analysis assumes that all Fire PRA target cables are thermoplastic. This assumption conservatively assigns the lowest radiant heat flux damage threshold and damage temperature threshold, 6 kW/m2 and 205 °C, suggested in Appendix H of NUREG/CR-6850. In practice, Attachment J of FPRA-PI-SCA describes the development of the Zone of Influence for each ignition source using the radiant heat flux damage criteria for thermoplastic cabling. Further, the calculation of the hot gas layer temperature provided in Attachment E of FPRA-PI-SCA used the temperature criteria for thermoplastic cable. The use of these damage criteria bounds the fire impacts expected for raceways containing a mixture of cables with varying insulation types. The contribution to the fire intensity due to the ignition of cable trays is a function of the tray heat release rate per unit area. Per the Guidance in NUREG-7010, the recommended heat release rate per unit area for thermoplastic cable trays is 250 kW/m2. In the PINGP Fire PRA, the heat release rate per unit area used is 328 kW/m2, which is the average of bench scale heat release rates for
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thermoset insulation types given in Appendix R to NUREG/CR-6850. (See Section 4.3 of FPRA-PI-SCA. As such, the PINGP conservatively estimates the contribution to the total fire heat release rate from cable trays.
As stated in Attachment F.3 of FPRA-PI-SCA, self-ignited cable fires are screened for all locations in the plant. These fire types were screened on the basis that only unqualified (thermoplastic) cable that are routed in cable trays may cause this type of fire. A review of the Engineering Change Package 20695 concluded that all thermoplastic and unknown cable types were routed through conduit.
FSS-C8-01 Integrity of credited fire wraps
Closed If raceway fire wraps are credited: a) ESTABLISH a technical basis for their fire-resistance rating, and CONFIRM that the fire wrap will not be subject to either mechanical damage or direct flame impingement from a high-hazard ignition source unless the wrap has been subject to qualification or other proof of performance testing under these conditions. There is no discussion of mechanical damage or direct flame impingement. There is also no discussion in the notebooks of the wrap being qualified. The justification found in the self-assessment should be included in one of the fire notebooks. Also a confirmatory walkdown to ensure fire wrap is not damaged and associated documentation is needed.
A list of all credited fire wraps per raceway or cable name is provided in Attachment D.9 of FPRA-PI-SCA (Revision 1). PINGP Modifications 94L483, Thermo-Lag Replacement with Darmatt and 00FP01, Kaowool Replacement with 3M Interam establish the basis for the 1 hour fire rating for these installed barriers. Inspection of the fire wraps such that the 1-hour qualification rating is maintained is governed by Procedure SP 1275. As such, it is appropriate to credit the installed wrap in these areas as documented in Attachment D.9 of FPRA-PI-SCA for the credited rating. It is important to note that these raceway barriers are not credited in HEAF scenarios due to the potential for mechanical damage. Further, fire wrap is not credited in the high hazard scenarios postulated in the Structural Steel Analysis (FPRA-PI-SS, Revision 1).
FSS-D7-01 Other Affected SR
Unreliability of detection in deluge system
Closed The Non-Suppression probability for the deluge system appears to be calculated incorrectly in several places in the FPRA-PI-SS report. In particular, in
Section 4.1.5 of the Structural Steel Fire Interaction Analysis Notebook (FPRA-PI-SS, Revision 1) documents the calculation of the
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PRM-A2 Table 20 on page 25 of 34 in report FPRA-PI-SS, Revision A, the unreliability of the detection system required to activate the deluge valve is not factored into the calculation. Additionally, it appears that the event tree in section P.1.3 and Figure P-1of NUREG/CR-6850 has been solved incorrectly. It also appears that the bullet titled "Pr(failure auto det):" and the bullet titled "Pr(failure auto supp):" on page P-6 of NUREG/CR-6850v2 is being interpreted incorrectly. This impacts the accuracy of the FPRA. Revise the calculation method on pages 25-26 of 34 in report FPRA-PI-SS, Revision A and document its accuracy.
non-suppression probability. Plant specific automatic non-suppression probabilities for the credited pre-action (PA-14 and PA-15) and deluge (DA-3 and DA-4) systems are given in this notebook. These probabilities represent the unavailability of the system based on plant-specific data. The unavailability factor, which represents the portion of time that these detection and suppression systems are unavailable due to testing or maintenance, was developed through a review of the PINGP eSOMs system. The impact of the credited systems is summarized in Section 4.1.6 of FPRA-PI-SS. The basis for these plant specific non-suppression probability values is provided in Attachment M to FPRA-PI-SCA (Revision 1) which contains a narrative description of the research completed to develop the unavailability factors for the credited automatic suppression and detection systems.
FSS-D7-02 Other Affected SR PRM-A2
Suppression system unavailability
Closed PRM model for fire compartment 18 does not reflect the basic events and failure rates identified in FPRA-PI-SCA, Rev. B. In particular, Section 6.19.2 of FPRA-PI-SCA states that the 0SPUR-----M is used in this room, but the PRM model does not include this basic event, but does include a different basic event that is not discussed for this compartment.” A similar issue was also noted for fire compartment 48GRP – Section 6.35 of FPRA-PI-SCA states that the 0SPUR-----M is also used in this room, but again, the PRM does not include this basic event, but does model a different basic event for this fire compartment. It is believed that the PRM is correct, and that the documentation is incorrect in these cases, but a complete review of all fire compartments has not been done to verify that these are isolated instances. This will result in an error in final CDF values.
Attachment M of FPRA-PI-SCA (Revision 1) provides a narrative description of the research completed to develop the unavailability factors for the credited automatic suppression and detection systems. Unavailability factors were developed through a search of the Prairie Island Nuclear Generating Plant eSOMS system for a three year period from November 26, 2006 to November 27, 2009. The total detection and suppression system failure probability is the union of the system unreliability values given in NUREG/CR-6850 and the unavailability values developed through this review. Using this total system failure probability value, the non-suppression probability for the single compartment fire scenarios was
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A complete review of the Single Compartment Analysis write-up needs to be performed to verify that the “Modeling Factors” that are documented for each compartment are correct, and reflect the ones actually used in the PRM. Any discrepancies need to be corrected.
calculated using the detection and suppression event tree shown in Section 4.1.5 of the single compartment detailed fire modeling report. This event tree follows the guidance in Appendix P of NUREG/CR-6850 for calculating the non-suppression probability. Further, the calculation of the non-suppression probability is available in “NSP” tab of the Quantification Spreadsheet (Attachment N). As such, the basic event 0SPUR-----M is no longer used to represent unavailability of the fire suppression system in FC18 or any other fire compartment analyzed in the PINGP Fire PRA.
FSS-D8-01 Evaluation of detection and suppression system effectiveness
Closed The SR states to document the assessment of fire detection and suppression system effectiveness in the context of each fire scenario analyzed. Did not find any evidence that this requirement was documented. Provide copies of the assessment for each fire scenario analyzed. Document that all systems are installed in accordance with NFPA and industry standards.
The PINGP fire PRA credits detection, automatic suppression, and manual suppression to stop the progression of the fire scenario beyond the 98th percentile ZOI only. In other words, detection and suppression systems are not credited to prevent target damage prior to impacting targets located in the ZOI. Section 4.1.5 of the single compartment detailed fire modeling analysis report (FPRA-PI-SCA, Revision 1) documents those detection and suppression systems that are credited in the PINGP Fire PRA for this purpose. Further, credited suppression systems were evaluated during walkdowns while developing the postulated fire scenarios to assess their effectiveness in controlling the fire growth in the corresponding scenarios (See FPRA-PI-SCA, Revision 1, Appendix L, Figure L-1 as an example). Specifically, walkdowns were conducted to ensure that sprinklers were present and capable of performing their intended function and the credited systems were reflected in the detection and suppression event tree
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described in NUREG/CR-6850.
FSS-D9-01 Component sensitivity due to smoke
Closed The self-assessment states that no equipment was felt to be sensitive to smoke damage, therefore no evaluation for smoke damage effects was performed. If electronic equipment is being used, smoke damage is likely. Many electrical parts have data on the effects of smoke, which can be consulted if needed to make a decision. The assumption that no smoke damage will occur needs to be justified. Provide justification and documentation for smoke damage applicability to support Category II and future use of the fire PRA.
Attachment D.7 to the FPRA-PI-SCA (Revision 1) report states that the approach for incorporating smoke damage in the Fire PRA follows the guidance available in NUREG/CR-6850, Appendix T. Long-term equipment damage from exposure to smoke, such as induced corrosion over a time scale ranging from days to months, is not a primary concern in a fire PRA because risk-significant fire scenarios are resolved on a time scale ranging from minutes to hours. Although some components are vulnerable to short-term smoke damage (electrical transmission equipment of >15 kV and higher, instrumentation with fine mechanical motion, and unprotected electronic components and circuit boards, etc.), short-term damage requires severe smoke exposure. In accordance with the discussion in NUREG/CR-6850, Appendix T, ambient smoke exposures that are expected to be encountered in a compartment during a fire would not be severe enough to cause short-term damage. According to the guidance in NUREG/CR-6850, the only situations inside the plant in which smoke conditions are expected to be severe enough to cause short-term smoke damage involve 1) Vulnerable components inside a burning electrical panel or cabinet or in adjacent panels in an interconnected bank or 2) Very high voltage (>15 kV) electrical transmission equipment exposed to a large oil fire. In both of these situations, there may be incidental smoke damage but the associated thermal conditions already cause the vulnerable equipment to fail even in the absence of smoke damage. That is, the Fire PRA assumes that the equipment will completely fail if the fire is postulated within the cabinet enclosure.
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In conclusion, the review of potential smoke damage to FPRA components was performed and no-smoke induced damage to Fire PRA equipment was identified that was not bounded by the dominant failure mechanisms (radiative and temperature).
FSS-D10-01 Other Affected SRs FSS-D11, FSS-H10
Walkdowns needed for added fire scenarios
Closed Since it is recognized that additional detailed fire scenarios need to be developed in order to accurately reflect the fire risk, this finding is being written to document that as more detailed scenarios are evaluated, confirmatory walkdowns of the newly defined scenarios to confirm the combinations of fire sources and target sets will also be required. These walkdowns need to be conducted in accordance with current EPM procedures. Such fire scenario revisions and associated walkdowns will be required to demonstrate an acceptable CDF/LERF value. Conduct and document confirmatory walkdowns of the newly defined scenarios to validate the fire sources and target sets
As documented in Attachment L of FPRA-PI-SCA (Revision 1), walkdowns of the fire modeling parameters, including verification of the combinations of fire sources and targets sets, were conducted. Further discussion of fire modeling walkdowns is contained within Section 4.1.5 and 4.1.9 of FPRA-PI-SCA which discuss the characterization of fire detection and suppression features and the calculation of the appropriate severity factor, respectively.
FSS-F1-01 Hydrogen piping and storage tanks
Closed During walkdowns it was determined that hydrogen lines in the turbine building have the capability of initiating a fire that would directly impinge on the building structural steel. In many plants, the use of fork trucks is not allowed in the area of exposed hydrogen pipelines. However, based on the walkdown performed, fork truck use is allowed in the area of exposed hydrogen pipelines at PINGP. The potential for a fork lift inadvertently impacting an exposed hydrogen line resulting in a direct hydrogen flame on exposed structural steel in the Turbine Building needs to be addressed. Additionally, the Hydrogen Storage Tanks outside the Turbine Building are stored in a physical configuration such that a fire in the storage tank area that impacts the valves and ends of the tanks could result in the tanks becoming missiles and punching through the concrete wall separating them from the Turbine Building. The potential for and impact of this scenario needs to be
Hydrogen fires were evaluated based upon likely break points, such as flanged joints or valves. Specifically, the frequency for Bin 19 Miscellaneous Hydrogen Fires is apportioned to the Auxiliary Building Mezzanine Level, Fire Compartment 59GRP as documented in the ignition frequency notebook, FPRA-PI-IGN Revision 1. These frequencies were included in the scenarios defined for these fire compartments in the detailed fire modeling analysis. In addition the 69NONDS and 70NONDS scenarios documented in FPRA-PI-TB (Revision 1) include the contribution of hydrogen fires in the frequency term. These scenarios conservatively include the failure of each unit's turbine building due to a hydrogen fire. In addition, the Structural Steel notebook, FPRA-PI-SS (Revision 1), documents a turbine building collapse due to
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addressed. Evaluate and document the missing hydrogen scenarios.
Closed If, per SR FSS-F1, one or more scenarios are selected, Complete a quantitative assessment of the risk of the selected fire scenarios consistent with the FQ requirements, including collapse of the exposed structural steel. At present, Table 22 on page 31 of 34 in report FPRA-PI-SS, Revision A, has not been completed and a quantitative assessment has not been done. The quantitative assessment table exists in the report but needs the values for CDF and LERF entered into it. Complete the table with the quantitative values. If the table is completed, this F&O will be satisfied.
FPRA-PI-SS (Revision 1) documents the structural steel analysis for the PINGP fire PRA. The scenarios involving structural steel are developed and quantified to assess total plant risk. FPRA-PI-FQ (Revision 1) documents the total risk contribution from the structural steel analysis.
FSS-G2-01 Outside area fires Closed It was noted during the walkdown that there are large transformers located outside of the turbine building which are close to the turbine building. There appear to be cables /ductwork between the transformers and the turbine building, and these cables/ductwork do not currently appear to be included in any fire compartment. Because there are no physical barriers between the transformers and the outside ductwork/cables, and because the walkdown team determined that an explosion and/or fire in one of the transformers could physically impact the outside of the turbine building, the ductwork/cables need to be considered as part of the transformer fire compartment. Industry events have occurred where large transformers have failed and the resulting fire or debris has impacted adjacent structures. Since the ductwork/cables are on the outside of the Turbine Building, they can be impacted by the Transformer fire, and need to be included in the evaluation. Fire scenarios involving the outside transformers and cables/ductwork external to the Turbine Building need to be evaluated as part of the single compartment
The Yard Fire Area (FC 28GRP) scenario is analyzed in the Single Compartment Analysis Notebook, FPRA-PI-SCA Revision 1, using a very large zone of influence, where all fire scenarios postulated in the fire PRA damage all components in the fire area. In addition, all conduits mapped to this fire area are conservatively failed in all fire scenarios postulated in this fire area.
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Closed Requirement is to quantify the risk contribution of any selected multi-compartment fire scenarios consistent with the FQ [Fire Quantification] requirements. Quantification approach should be consistent with FQ supporting requirements. CCDP’s are assumed to be 1 for combined areas. Multi-compartment fire CCDP’s currently add up to mid-E-4 range, which is conservative. No LERF information is provided, need to provide LERF. Suggest providing more realistic evaluations and including LERF information for unscreened scenarios.
Scenarios that are not screened are included in the plant CDF and LERF quantification consistent with FQ requirements. Realistic estimates of multi-compartment CCDPs and CLERPs were developed following the May 2012 peer review. These new results have been incorporated into the Fire PRA and are reflected in the final quantification results for multi-compartment scenarios. See Section 5.3 and Attachment C of FPRA-PI-MCA (Revision 1).
FSS-H5-01 Other Affected SR PRM-C1
Fire compartment analysis documentation difficult to follow
Closed Section 6 of FPRA-PI-SCA, Detailed Compartment Analysis Notebook, is very cumbersome to use and understand. In addition, many sections have nothing substantive except for a reference to another section - which takes up a lot of pages and makes it difficult to correlate the information that is in fact relevant to the subject. Also, many references are merely to event or computer ID's and codes that have no corresponding name or description. To facilitate use of the FSS notebook in the future, a substantial rewrite is needed that better organizes and condenses common information and references applicable to each compartment, and provides missing information. Improved documentation will be very valuable for future use, upgrade and peer review of the Fire PRA. Provide additional tables and descriptions of computer ID's rather than continual reference to another location that is not specifically identified (i.e. A specific part or page in an appendix, rather than just reference to an appendix). Reduce duplicate statements of general information that increases the number of pages and complexity of the report.
The entirety of the single compartment detailed fire modeling analysis report (FPRA-PI-SCA, Revision 1) has been reorganized such that it follows the outline of the single compartment detailed fire modeling process given in NUREG/CR-6850. As part of this effort, the additional tables describing inputs, credited detection systems, and credited suppression systems were added. References to other sections were edited such that a subsection name and title was given, facilitating the use of this document as a reference. Finally, a number of attachments have been added to the report which compile specific features of the report into stand-alone write-ups. In doing so, the main body of the report became much easier to navigate and less cumbersome to use as a procedure defining the analysis.
FSS-H5-02 Scenario specific parameter uncertainty
Closed Section 6 of FPRA-PI-SCA, Detailed Compartment Analysis Notebook, does not include any kind of detailed documentation related to scenario specific parameter uncertainty evaluations such that it can be
Section 4.2 of FPRA-PI-SCA (Revision 1) provides a summary of the uncertainty sources and treatments associated with fire scenario development and detailed fire
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referenced and utilized in Section 4.1.11 and Table 2 of Appendix A of the Uncertainty and Sensitivity Notebook (FPRA-PI-UNC, Revision B). This missing information is needed for long term use, upgrade and peer review of the Fire PRA. Provide documentation of scenario specific parameter uncertainty evaluations in report FPRA-PI-SCA, Detailed Compartment Analysis Notebook.
modeling for single compartment fire scenarios in the PINGP Fire PRA. Specifically, uncertainties related to the selection of transient zones, fire location, fire growth and propagation, activation and function of the detection and suppression system, the selection of damage criterion, conduit routing, selection of fire models, and the inputs to the chosen fire models are outlined. Sources of uncertainty related to fire scenarios are qualitatively analyzed in Section 6.1.11 of Notebook FPRA-PI-FQ (Revision 1). The parameters analyzed are those that are key to the development and appropriate modeling of fire scenarios. For example, the analysis discusses the uncertainty associated with cable damage criteria, treatment of fire-induced failure of instrument air, and treatment of conduits. It also identifies and discusses those parameters that influence the development of fire scenarios, such as material properties, intervening combustibles, and non-suppression probabilities. In summary, the analysis discusses the uncertainty of parameters at a level of detail sufficient for long-term use, upgrade, and peer review of the Fire PRA.
FSS-H9-01 Fire modeling parameter uncertainty
Closed The uncertainty discussions in FPRA-PI-UNC, Revision B, Uncertainty and Sensitivity Notebook, are very general in nature and not specific to the analyses in the fire scenario selection tasks of the Fire PRA. Therefore, this SR is judged to be met, although the need for more specific information and evaluation has been identified. Lack of specific detail which would allow long term use of the Fire PRA. Provide expanded, upgraded, and enhanced qualitative discussions, and possibly quantitative
Section 4.2 of FPRA-PI-SCA (Revision 1) provides a summary of the uncertainty sources and treatments associated with fire scenario development and detailed fire modeling for single compartment fire scenarios in the PINGP Fire PRA. Specifically, uncertainties related to the selection of transient zones, fire location, fire growth and propagation, activation and function of the detection and suppression system, the selection of damage criterion,
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results related to uncertainty in the FSS portions of the fire PRA.
conduit routing, selection of fire models, and the inputs to the chosen fire models are outlined. Additionally, an expanded analysis of sources of uncertainties to the fire selection task of the Fire PRA is given in Section 6.1.11 of Notebook FPRA-PI-FQ (Revision 1). The analysis includes a detailed discussion of specific parameters that are key to the development of fire scenarios, including, for example, the uncertainty associated with cable damage criteria, treatment of fire-induced failure of instrument air, and treatment of conduits. It also identifies and discusses those parameters that influence the development of fire scenarios, such as material properties, intervening combustibles, and non-suppression probabilities. In summary, the analysis discusses the uncertainty of parameters at a level of detail sufficient for long-term use, upgrade, and peer review of the Fire PRA.
FSS-H10-01 Other Affected SR FSS-D10
Documentation of walkdown information
Closed The electronic copies of walkdown data located at the following path: O:\Xcel Energy - PINGP Fire PRA\Task 1006-FIF\! Old Files from P2007\PINGP Initiator Photos and Data Sheets" needs to be incorporated into the Fire PRA documentation. This is needed in order to use, maintain, and update the Fire PRA in the future and provide completeness of the FPRA Documentation. It is recommended that the walkdown data in O:\Xcel Energy - PINGP Fire PRA\Task 1006-FIF\! Old Files from P2007\PINGP be directly referenced in FPRA-PI-SCA, Detailed Compartment Analysis Notebook or this information should be incorporated into an appendix to report FPRA-PI-SCA, Detailed Compartment Analysis Notebook. The walkdowns performed need to be easily retrievable, and to be able to be verified to be complete (i.e. none are
Attachment L to FPRA-PI-FQ (Revision 1) contains documentation of the walkdowns conducted to validate the inputs to the fire models and target set combination information. Inclusion of this attachment improves the ability of PINGP to use, maintain, and update the Fire PRA in the future.
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inadvertently missing).
IGN-A1-01 Other Affected SRs UNC-A1, UNC-A2
Sensitivity study needed for updated ignition frequencies
Closed PI used the ignition frequencies provided in NUREG/CR-6850 Supplement 1 (FAQ-0048). When using the Supplement 1 frequencies a sensitivity analysis must be performed against the ignition frequencies in NUREG/CR-6850. When using the Supplement 1 frequencies, a sensitivity analysis needs to be performed against using the NUREG/CR-6850 fire ignition frequencies. Perform sensitivity analysis between the NUREG/CR-6850 fire ignition frequencies and the frequencies provided in NUREG/CR-6850 Supplement 1. The sensitivity analysis only needs to be performed for those bins characterized by an alpha less than or equal to 1.
A sensitivity study has been performed in FPRA-PI-FQ (Revision 1) using the generic ignition frequencies provided in NUREG/CR-6850 for those bins with an alpha value of less than or equal to one as given in EPRI 1016735. The exception to this is Bin 9 where NUREG/CR-6850 Supplement 1 states that EPRI 1016735 is incorrect in that Bin 9 should have an alpha value greater than one, hence a sensitivity analysis does not need to be performed for this bin.
IGN-A7-01 Other Affected SR IGN-B3
Ignition source binning Closed Bin 29, Yard Transformers (Others) is not filled. Additionally, Bin 15.1 Electrical Cabinets Non-HEAF may not be filled correctly. Fire Compartment 20 (Switchgear Room) has more HEAF electrical cabinet counts (13) than non-HEAF counts (7). This suggests that electrical cabinets with the potential to have a HEAF event are not assigned a non-HEAF fire event. Thus the failure mode of electrical cabinet fire may not be counted and underestimate risk in compartments with the potential for HEAF. This may inadvertently skew risk away from important areas such as transformer yard and switchgear rooms. Fill Bin 29 similar to how bins 27 and 28 are filled. If Electrical Cabinet can have a HEAF failure mode, ensure that both Bins 15.1 and 15.2 are filled. Also, for Bin 21 (Pumps), believe that the split fractions in 6850 electrical/oil are still valid.
NUREG/CR-6850, Task 6, states that Bin 29 is reserved for items associated with the yard transformers, such as oil-filled power output cables, but not the transformers themselves. After additional review with plant personnel, and review of photographs showing the transformer areas, Bin 29 remains not applicable to PINGP. There are no initiators near the transformers that are not a part of the transformers themselves nor that are not already counted in a different bin. For example, the frequency for the segmented bus ducts in the transformer areas is counted in Bin 16.1. It should be noted that PINGP does not have any oil-filled power cables near the transformers. The ignition frequency notebook, FPRA-PI-IGN Revision 1, has been updated to clarify the lack of Bin 29 fires at PINGP. Electrical Cabinets – HEAF, Bin 15.2, were removed from the Bin 15.1, Electrical Cabinets Non-HEAF, count because of the guidance provided in EPRI 1016735. This
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guidance states that regarding the updated EPRI generic ignition frequency values, “the high energy fire ignition events associated with electrical cabinets have been removed from bin 15, now designated bin 15.1, to a new bin, designated 15.2, electrical cabinets-HEAF” (2-6). Electrical cabinets that are a HEAF concern should be counted in both Bin 15.1 and Bin 15.2 as both types of fires could occur in these electrical cabinets. Initiators counted in Bin 15.2 have been added to Bin 15.1 and documented in the Fire PRA documentation. The split fractions provided in Table 6-1 of NUREG/CR-6850 are still applicable, and have not been superseded by more recent FAQs/method reports. The fraction values are implemented in the detailed fire modeling as documented in the single compartment analysis notebook, FPRA-PI-SCA Revision 1. Note that the oil fire split fractions have been clarified in various FAQs since NUREG/CR-6850, and those are applied as documented in FPRA-PI-SCA Revision 1.
IGN-A7-02 Ignition frequency summation
Closed The summation of the fire ignition frequency bins does not equal the summation of all ignition sources. The ignition frequency is not distributed correctly (with respect to transient ignition source bins). As discussed with FPRA personnel, ensure that the generic plant locations (battery rooms, diesel rooms) are lumped into the correct transient bin (control/aux reactor, or plant wide transient bins). Document the sum of the ignition frequencies in the calculation and verify ignition frequencies are conserved for each revision.
After review of the fire compartment generic locations it was determined that, based on plant configuration and the guidance available in Table 6-2 of NUREG/CR-6850, the generic plant locations should be corrected as suggested. The PINGP fire ignition frequencies have been updated to properly assign generic locations to fire compartments (FPRA-PI-IGN, Revision 1). The ignition frequencies were verified to ensure that fire ignition frequency was conserved.
IGN-A9-01 Assignment of maintenance weighting
Closed No maintenance factors of 50 have been assigned in the ignition frequency task. A fire compartment, like
The Finding was assessed against IGN-A9 for not assigning a maintenance factor of 50. In
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factors the maintenance shop, which has contestant [sic] maintenance activities should have this number assigned. No fire compartments have been assigned a maintenance factor of 50. By not assigning a maintenance factor of 50 to high maintenance compartments, this can divert risk to these compartments. Review plant information (review work orders/interview operators) to identify fire compartments which may have higher than normal maintenance activities. If, because of the large size of the PAUs, no single PAU is determined to warrant a maintenance factor of 50, even if rooms within the PAU would typically have a maintenance factor of 50 applied to them, the basis and methodology employed for determining the appropriate maintenance factor for the whole PAU should be documented.
addition, the Finding stipulates that a review of plant information (review work orders/interview operator) should be completed to identify fire compartments with higher than normal maintenance activities. The compartment of interest for this Finding was the 8GRP. Fire Compartment 8GRP includes nearly the whole turbine building and was assigned a maintenance weighting factor of 10. The Maintenance Shop is part of the 8GRP and it would normally be assigned a maintenance weighting factor of 50. But 8GRP is a very large fire compartment, and the maintenance shop is only a small portion of the compartment, so assigning a Maintenance Weighting Factor of 50 to the 8GRP would disproportionally weight maintenance for the Fire Compartment. As such, the methodology described in NUREG/CR-6850 as implemented in the Fire Ignition Frequency analysis was followed correctly. The documentation in the ignition frequency notebook, FPRA-PI-IGN (Revision 1), has been updated to reflect this. In addition, weighting factors have been redistributed such that “3” is the median value for MF, OF, and SF. As such, the PINGP FPRA is aligned with the intention of the weighting factor analysis.
IGN-B5-01 Other Affected SRs IGN-A10, CF-A1, UNC-A2
Identification assumptions and sources of uncertainty
Closed Although there is a section in the Ignition Frequency notebook entitled Assumptions, it does not appear to be complete. This is evidenced by the following examples: In section 5.2.7-2, the notebook states: “To estimate the ignition frequency for transient sources or activities, the concept of “influencing factors” is used
Assumption 7 addresses the judgment-based influencing factor analysis. Detailed reviews regarding screened equipment have since been completed, so this comment is no longer applicable. FPRA-PI-IGN Revision 1 has been updated in accordance.
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to assess the likelihood of fire ignition. These factors represent a subjective judgment of the relative weighting of transient sources of ignition.” A subjective judgment is an assumption by definition and should be identified as such. In Section 5.2.6, the notebook states: “Equipment screened during the walk down process is not included in the total plant count for its bin, and does not contribute to the ignition frequency of the compartment where it is located; however, later reviews during the Detailed Compartment Analysis task may result in screened targets being added back as ignition sources.” Since the detailed reviews do not appear to be complete, this is a potential source of uncertainty. Additionally, Section 6.0 on Uncertainties and Sensitivities refers to the Uncertainty and Sensitivity Task report, which also does not appear to be complete. This is evidenced by the following examples: Section 9.0 of the Ignition Frequency notebook contains Open Items associated with the Ignition Source Frequency calculation. The first one states that several fire compartments do not currently contain cable loading estimates, and that these will need to be estimated. Estimations, by definition, contain uncertainty and should be identified as such. The second one states: “The fire ignition sources for Unit 2 Containment were based off the Unit 1 walk down. “ This is equivalent to assuming that Unit 1 and Unit 2 Containments are identical with respect to ignition sources, and should be identified as such. Section 4.1.6 of the Uncertainty and Sensitivity notebook states “Point values for the associated severity factors and floor area ratios (where applicable) were calculated, because these do not
Open items have been closed for FPRA-PI-IGN Revision 1, including cable loading estimates and Unit 1 vs. Unit 2 ignition source counts. Area calculations and severity factor calculation is completed and documented in FPRA-PI-SCA, Revision 1. Also note that the uncertainty analysis (now included in the quantification notebook, FPRA-PI-FQ Revision 1) has been updated to improve the resolution of the PINGP uncertainty analysis.
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have uncertainty associated with them; the point values were entered into the fire PRA with no variability.” Although the floor area ratios may be calculated, it is not evident where these calculations are performed, so their accuracy could not be confirmed. Additionally, there are no severity factors associated with ignition frequencies, so this statement makes no sense. Section 4.1.6 of the Uncertainty and Sensitivity notebook also states “With respect to accuracy, issues arising from mapping of plant specific locations to generic locations, equipment counting, determination of location-weighting and ignition source weighting factors, and plant specific fire experience were properly identified, categorized and analyzed.” Although the majority of ignition sources are easily partitioned, it is very difficult to ensure that ALL ignition sources have been identified and appropriately partitioned in the plant since NOT ALL locations have had detailed walk downs associated with them to confirm the presence or absence of ignition sources. Additionally, the use of weighting factors introduces uncertainties into the evaluation, and should be identified as such. Ensure all assumptions, and implied assumptions, are contained in the assumptions section. Additionally, ensure that sources of uncertainty associated with the ignition source frequency development are appropriately identified and included, either in this notebook, or in the Uncertainty and Sensitivity notebook. It is recommended that a Table be added in PRA-PI-IGN with the mean frequencies, 5% bound and 95% bound.
FQ-A2-01 Dual unit trip assumption
Closed This SR states: For each fire scenario selected per the FSS requirements that will be quantified as a contributor to fire-induced plant CDF and/or LERF, IDENTIFY the specific initiating event or events (e.g.,
The approach for selection of fire-induced initiating events was changed during the transition of the fire PRA model from PRISM to FRANX. Two new unit-specific transient
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general transient, LOOP) that will be used to quantify CDF and LERF. The quantification process currently assumes BOTH units experience a Reactor Trip with loss of PCS [Power Conversion System] for EVERY scenario. Although it is reasonable to assume a Single Unit reactor trip with loss of PCS is initiated by each fire initiator, assuming a dual unit trip is overly restrictive, and is masking real risk insights. Although it is not possible to determine the potential impacts of a fire in all PAUs because not all cables were traced, there are a number of obvious PAUs that can be more realistically evaluated. These include PAUs like Containment and ECCS pump rooms, where opposite Unit cables would not realistically expected to be present. For these types of PAUs, at a minimum, the assumption of a dual unit trip needs to be re-evaluated.
initiators were added to the model that represent transient events that can be caused by the fire. The addition of these new fire-induced transient initiators to the fault tree logic was done so that all sequences that represent equivalent sequence progressions are combined in the fault tree. In the top logic for the sequences, all potential initiating events which can cause the initiating event condition, either by themselves or in combination with equipment failures, are input into the logical representation of the sequence (for example under the small LOCA logic sequence, the initiating event for a small pipe break LOCA is included as well as the fire-induced transient event initiator combined with equipment failure (e.g. stuck open PORV), which causes a small LOCA). Since the FRANX or CAFTA quantification applies fire impacts to equipment, the fire-induced component failures combined with the use of the fire-induced transient event initiator addresses all fire induced initiators. Since the addition of these initiators is unit-specific, this eliminates most risk-significant dual-unit impacts that are not legitimate contributors. However, a review of the final results for this issue did identify several non-risk significant cutsets. With regards to the finding, because the dual-unit cutsets are not risk-significant, they are not masking real risk insights.
FQ-B1-02 Other Affected SRs FQ-A1, FQ-A3, FQ-A4, FSS-D2
Computer code verification and validation
Closed This SR has look backs to the Internal Events QU [Quantification] SRs. QU-B1 provides requirements for ensuring that computer codes used in the analysis have demonstrated to generate appropriate results when compared to those from accepted algorithms and with ensuring that method-specific limitations and features of software used in the quantification process that could impact the results are identified.
Since the peer review, the Fire PRA model has been converted from the PRISM software to the use of FRANX for quantification. Since the PRISM program was abandoned, it was not verified and validated (V&V'd). In terms of the fire modeling, a comprehensive V&V has been conducted to
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This F&O does not address standard quantifying software (CAFTA, FTREX, etc.), but rather it focuses on non-standard software which is used in the quantification process. Per discussions with EPM and Xcel Energy personnel the following non-standard software is used in the quantification process: PRISM - This EPM developed software is used for associating basic events to fire areas based on the fire location, equipment location, cable routing, etc. This software is also used to create the flag files used in the quantification process for "failing" specific target sets based upon the associations performed by the software. EPM has also developed a spreadsheet based on NUREG-1805 for performing FDT evaluations, but no validation of this software could be located. This software is subject to the requirements of QU-B1. Since this software directly impacts the development of the fire impacts, they directly impact the results and insights of the Fire PRA. As such, any errors in the software can have a significant impact on results, and the software needs to be V&V’d. Any limitations of the software also need to be identified and discussed. Ensure the software is V&V’d in accordance with standard site Software Quality Assurance Procedures, and identify any software limitations.
all the fire modeling supporting the Fire PRA as documented in the corresponding notebooks (FPRA-PI-MCR, FPRA-PI-MCA, and FPRA-PI-SCA, all Revision 1; FPRA-PI-RRA and FPRA-PI-TBA, both Revision 0). The V&V was conducted following the guidance available in NUREG 1824. A summary of the V&V has been documented in the Attachment J of the NFPA 805 LAR.
FQ-E1-01 Completion of additional detailed scenarios for high risk compartments
Closed IDENTIFY significant contributors in accordance with HLR-QU-D and HLR-LE-F and their SRs in Section 2 with the following clarifications: a) SR QU-D5a and QU-D5b of Section 2 are to be met including identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors; b) SR QU-D5b of Section 2 is to be met recognizing that "component" in Pa Section 2 is generally equivalent to
Since the May 2012 peer review, the model has been updated to use FRANX as the primary quantification software. This change, along with other improvements to the model, allow for the development of importance measures without the removal of any scenarios. This item is considered closed because the updated FQ analysis (FPRA-PI-FQ Revision 1) no longer removes scenarios from the quantification results and all
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"equipment" in this standard; c) SR QU-D3 for comparison to similar plants is not applicable; d) SR LE-F3 including the "Note" for that SR of Section 2 is to be met; (1) following HLR-QU-D of Section 2 with clarifications above concerning SRs QU-D5a and (2) but the uncertainty requirement and reference to Table 2.2.7.6(e) in Section 2 does not apply here. See 4-2.13 and DEVELOP a defined basis to support the claim of non-applicability of any of the requirements under the sections in Part 2. There are several identified fire scenarios that have been removed from the quantification results. These scenarios were removed to prevent the skewing of the importance calculations. This removal was noted in the notebook so that when detailed analysis of the removed scenarios was performed the removed scenarios would be included in the total CDF and LERF. Perform the detailed analysis for the scenarios that were removed and include the new results in the CDF and LERF.
scenarios are subject to a detailed analysis.
FQ-F1-01 Completion of revised quantification analyses
Closed DOCUMENT the CDF and LERF analyses in accordance with the HLR-QU-F and HLR-LE-G and their SRs in Section 2 with the following clarifications: a) SRs QU-F2 and QU-F3 of Section 2 are to be met including identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors; b) SR QU-F4 of Section 2 is to be met consistent with Section 4.2.13; c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Section 2 are to be met consistent with Section 4.2.13, and DEVELOP a defined basis to support the claim of non-applicability of any of the requirements under these sections in Section 2. There are several tables and sections in the notebook which are incomplete. There are discussions about uncertainty and truncation, but they are based on the incomplete tables and sections in the notebook.
This Finding discusses documentation discrepancies related to the Fire Quantification (FQ) notebook at the time of the peer review. This document has since been updated to support the NFPA-805 LAR submittal and Supplement. Specific to the issues discussed in the peer review Finding, the parametric uncertainty analysis is covered in Section 6.0 of the Fire Quantification notebook (PI-FPRA-FQ, Rev. 1), while the truncation analysis is discussed in Section 5.7 of the Fire Quantification notebook. Also, the qualitative uncertainties are discussed in Section 6.0 of the FQ notebook. The Fire Quantification notebook has been updated to provide complete information (tables, etc.) consistent with all the required
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Ensure that all tables and section of the notebook are complete. Update the uncertainty discussion and truncation study in the notebook to reflect the final information [sic].
quality elements in the Fire PRA standard (2009 Ed.) and other information necessary to support the NFPA-805 LAR submittal.
UNC-A1-01 Discussion of uncertainty introduced by assumptions
Closed This SR requires in part: "PERFORM the uncertainty analysis in accordance with HLR-QU-E and its SRs" which include QU-E1, QU-E2, QU-E3, and QU-E4. This element has not been satisfied. There are a number of assumptions and sources of uncertainty which were not identified or which were identified but whose impact was not provided. Examples of areas of uncertainty which should be addressed further include: HRR values used, non-suppression probabilities used, choices of target sets, location of instrument air headers relative to fire scenarios and fire impacts due to soldered joint failures. Other examples include the issue discussed in F&O IGN-A1-01, regarding a comparison of new vs. old ignition frequencies; the issue discussed in F&O FSS-A5-02, regarding parameter uncertainty evaluations required for category two for SR FSS-H5. It was assumed that containment heat removal is not required for success in any scenario; this assumption was not explored. Additionally Table 2 in the uncertainty notebook is confusing. The information associated with how kerite was treated, and the basis for the treatment is unclear, and the potential impacts of the assumptions which were made about kerite on the model were not identified. Some aspects of required search for, and characterization of, sources of uncertainty / assumptions have not been performed completely. Address the specifically cited examples, and search all main Fire PRA reports for assumptions. If additional items are found then list them and address
An expanded analysis of sources of uncertainties of the Fire PRA is given in Section 6.0 of Notebook FPRA-PI-FQ Revision 1 (sources of uncertainties specific to the internal events PRA are addressed separately as part the internal events PRA). The analysis uses an ordered approach where each task of the Fire PRA is reviewed to identify and characterize the sources of uncertainties. As part of that approach, the analysis discusses uncertainties associated with key elements of the Fire PRA, such as those cited in the finding, including, for example, heat release rates, non-suppression probabilities used, choice of target sets, treatment of fire-induced failure of instrument air, and treatment of conduits. In summary, the approach provides a structure to adequately identify and characterize the sources of uncertainty of the Fire PRA.
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them.
UNC-A2-01 Quantitative uncertainty analysis incorporating all input parameters
Closed This SR says: INCLUDE the treatment of uncertainties, including their documentation, as called out in SRs PRM-A4, FQ-F1, IGN-A10, IGN-B5, FSS-E3, FSS-E4, FSS-H5, FSS-H9, and CF-A2 and that required by performing Part 2 referenced requirements throughout this Standard. Some of these back-referenced SR's appear not to be met: FSS-E3 requires “Provide a mean value of, and a statistical representation of, the uncertainty intervals for the parameters used for modeling the significant fire scenarios.” Statistical representations of the uncertainty intervals for the parameters used for modeling significant fire scenarios were identified for HRA, data, and ignition frequencies. However uncertainty information was not provided for other parameters. FSS-E4 requires “PROVIDE a characterization of the uncertainties associated with cases where cable routing has [Note (1)] been assumed based on SRs CS-A10 and/or CS-A11. NOTE: (1) Uncertainties associated with cases where cable routing was assumed may be associated with the exact location of the cables with respect to the ignition sources, and fire-resistance characteristics and fire protection (e.g., fire-resistant covers) of the cables.” However; No characterization of uncertainties associated with cable routing was identified. (assumptions re. conduit for example). FSS-H5 requires “Document fire modeling output results for each analyzed fire scenario, including the results of parameter uncertainty evaluations (as Performed) in a manner that facilitates Fire PRA applications, upgrades, and peer review.” However, no parameter uncertainty evaluations were identified. Uncertainty information has not been developed for a number of important parameters. This makes it
The identification and characterization of sources of uncertainties is documented in Section 6.0 of Notebook FPRA-PI-FQ (Revision 1). An ordered approach was used where each task of the Fire PRA was reviewed and a qualitative characterization of the identified sources of uncertainty was given. This review included the characterization of the uncertainty associated with fire scenario development, which involved, for example, an evaluation of the treatment of fire impacts on conduits, instrument air piping, and more generally target damage. In addition, a quantitative analysis of uncertainty was performed for the parameters lending themselves to a statistical representation of uncertainty. The results of that quantitative evaluation are given in Section 6.0 of Notebook FPRA-PI-FQ. In summary, the analysis discusses the uncertainty of parameters at a level of detail sufficient for long-term use, upgrade, and peer review of the Fire PRA
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impossible to provide a complete understanding of the uncertainty of the results. Provide more detailed characterizations of uncertainty for the identified elements.
FSS-D6-01 (from Focused Scope Peer Review conducted on November 5, 2013)
Justification for the underlying assumptions used in the hot gas layer calculations
Closed Discussion: While determination of the hot gas layer temperatures using the updated MQH and FPA methods is deemed appropriate, further justification for the underlying assumptions used in the hot gas layer calculations need to be made. Basis for Significance: There are potential non-conservatisms in the presented work, and additional work should be undertaken to resolve if the hot gas layer input assumptions are biasing the calculations. As such, a level of finding is appropriate. Possible Resolution: Include sensitivity cases that demonstrate the use of a 1 m^2 opening for the MQH method is appropriate for rooms with expected sealed penetrations and that do not have wall/floor/ceiling openings but that may have small opening such as under door(s) as compared to hot gas layer calculations using MQH methods with much smaller openings and/or comparison with Beyler or FPA methods. Should the results indicate that for rooms with expected openings of less than 1 m^2 that the MQH method is non-conservative, consider alternate approaches for hot gas layer temperature determination or retain potentially screened PAUs for more detailed hot gas layer calculations.
Table 4-1 in FPRA-PI-SCA (Revision 1) shows those compartments that were analyzed using detailed fire modeling. For each of these compartments, the progress to and timing of the onset of hot gas layer conditions were analyzed. In general, the MQH correlation as described in Attachment E of FPRA-PI-SCA was used for this analysis. There are several compartments in which the hot gas layer scenarios were screened by crediting the results of the multi-compartment fire modeling analysis (FPRA-PI-MCA, Revision 1) where CFAST was used for analyzing these types of scenarios. Specifically, the development of a hot gas layer was screened for the following compartments: FC1, FC28GRP, FC58GRP, FC59GRP, FC71, and FC101GRP based on CFAST results. For the scenarios where the MQH correlation was used to calculate hot gas layer temperature, the application was verified and validated as shown in FPRA-PI-SCA. The results of the validation demonstrate that the input parameters are within the limits of applicability and that the results are conservative because the heat losses to the walls have been ignored. For Fire Area 18 (i.e., the relay room), the hot gas layer temperatures have been calculated using the zone model CFAST as documented in FPRA-PI-RRA, Revision 0.
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Table V-2 Fire PRA–Summary of Category I and Not Met SRs
SR Capability Category Topic Status
CS-A10 CC I As identified in Finding CS-A10-01, a limited number of Fire Areas did not have specific cable routing information available. Open item No. 1 on Page 17 of 17 of FPRA-PI-CS, Revision C, states that "in order to fully comply with Capability Category II, cables that are routed through fire compartments 2A, 41B-1, 46A, 58A, 58B, 58C, 58D, 76A, 78E, 86, 94A, 94B, 94C, 94D, 94E, and 94F need to be identified. This will be accomplished by identifying routing on electrical drawings, and with walkdowns performed as needed." Since this has not yet been completed, and the methodology not specified, this SR is provisionally met. Route cables through the listed fire compartments, and perform walkdowns to confirm accuracy of the routing. Utilize EPM Division Procedure EPM-DP-EP-005, Revision 1-Post-Fire Safe Shutdown Cable Routing and Component Location, and EPM Division Procedure EPM-DP-EP-004, Revision 2-Post Fire Safe Shutdown Cable Identification-February 2011.
Current CC: Met at CC I As part of the detailed fire modeling documented in FPRA-PI-SCA, risk contributing fire compartments identified through the development of the Fire PRA have been subdivided into Fire Scenarios and the targets have been assigned based on walkdowns and drawing inspections. Only those conduits with unknown routing have been mapped to all the fire scenarios within the fire compartment. This process removed the excess conservatism identified in this finding. This SR is considered to be Met at Capability Category I due to cable routing not being by fire compartment or physical analysis unit in all cases.
CS-B1 CC I As identified in FPRA-PI-CS, Appendix R overcurrent coordination and protection analysis has been reviewed but the analysis of additional circuits identified during the Fire PRA is currently in progress and is therefore not complete. Appendix R Breaker Coordination study has been reviewed. Additional breaker coordination is still being performed.
Current CC: Met at CC II Breaker fuse coordination studies (PRA Calculation V.SPA.12.016 and PRA Calculation V.SPA.12.018) were conducted and 22 Fire PRA credited power supplies were identified that currently do not coordinate but will be modified to achieve coordination. The Fire PRA assumes that there are not any coordination issues with the power supplies that are credited in the Fire PRA model. Resolution of coordination issues is captured in Attachment S of the NFPA-805 LAR.
CS-C4 Not Met As identified in FPRA-PI-CS, Appendix R overcurrent coordination and protection analysis has been reviewed but the analysis of additional circuits identified during the Fire PRA is currently in progress and is therefore not complete. Appendix R Breaker Coordination study has been reviewed. Additional breaker coordination is still being performed.
Current CC: Met at CC II Breaker fuse coordination studies (PRA Calculation V.SPA.12.016 and PRA Calculation V.SPA.12.018) were conducted and 22 Fire PRA credited power supplies were identified that currently do not coordinate but will be modified to achieve
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SR Capability Category Topic Status
coordination. The Fire PRA assumes that there are not any coordination issues with the power supplies that are credited in the Fire PRA model. Resolution of coordination issues is captured in Attachment S of the LAR.
FSS-B2 CC I Per Procedure EPM-DP-RSD-005, the MCR abandonment scenarios were developed to bound the risk in the main control room. This meets the standard at the CC-I level. Additionally, Section 9.2.1 of FPRA-PI-MCR, Main Control Room Analysis, Revision B (page 33 of 79) uses a bounding assumption associated with how the probability of a fire starting near a target is determined, which is acceptable for a bounding analysis, but will need to be revisited in order to ensure the MCR risk is realistically characterized.
Current CC: Met at CC III As described in Sections 7.1, 7.2, and 7.3 of FPRA-PI-MCR, multiple control room abandonment scenarios have been evaluated (See Figure 6). In particular, it should be noted that the postulated scenarios consider fires occurring in both fixed and transient ignition sources that result in the loss of specific functions due to damage to equipment and raceways found in the MCR but do NOT result in MCR abandonment. It was concluded that the various scenarios reasonably characterize the MCR abandonment risk because the probability of abandonment depends on fire-generated environmental conditions (Section 9.1). The extent of fire damage increases under abandonment conditions to include contributions of cable routing to panels (Section 6.0), Operator actions upon abandoning the control room (Section 7.0) are accounted for, and the HRA development for control room abandonment is based on operators mitigating the event from the remote shutdown panels (Section 8.0). Therefore, this requirement is met at Category III.
FSS-C8 Not Met There is no discussion of mechanical damage or direct flame impingement. There is also no discussion in the notebooks of the wrap being qualified but there is justification in the self-assessment. The justification from the self-assessment should be included in one of the fire notebooks. Also, a confirmatory walkdown to ensure fire wrap is not damaged, and supporting documentation would be needed.
Current CC: Met at CC II A list of all credited electrical raceway fire barrier systems (ERFBS) per raceway or cable name is provided in Attachment D.9 of FPRA-PI-SCA (Revision 1). PINGP Modifications 94L483, Thermo-Lag Replacement with Darmatt and 00FP01, Kaowool Replacement with 3M Interam establish the basis for the 1 hour fire rating for these installed barriers.
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Table V-2 Fire PRA–Summary of Category I and Not Met SRs
SR Capability Category Topic Status
Inspection of the fire wraps such that the 1-hour qualification rating is maintained is governed by Procedure SP 1275. As such, it is appropriate to credit the installed wrap in these areas as documented in Attachment D.9 of FPRA-PI-SCA for the credited rating.
FSS-D9 CC I The self-assessment states that no equipment was felt to be sensitive to smoke damage therefore no evaluation for smoke damage effects was performed. An assumption that no smoke damage occurs needs to be justified.
Current CC: Met at CC II As discussed in the disposition of Finding FSS-D9-01 in Table V-1, the current analysis accounts for the effects of smoke damage. Attachment D.7 to the FPRA-PI-SCA (Revision 1) report states that the approach for incorporating smoke damage in the Fire PRA follows the guidance available in NUREG/CR-6850, Appendix T.
FSS-F3 CC I At the time of the review, Table 22 on page 31 of 34 in report FPRA-PI-SS, Revision A, has not been completed and a quantitative assessment has not been done. The quantitative assessment table exists in the report but needs the values for CDF and LERF entered into it. This SR is considered met at a CC I level because Table 19 in notebook PINGP-FPRA-PI-SS provides fire scenario frequencies that will be used in the quantification task.
Current CC: Met at CC II FPRA-PI-SS (Revision 1) documents the structural steel analysis for the PINGP fire PRA. The scenarios involving structural steel are developed and quantified to assess total plant risk. FPRA-PI-FQ documents the total risk contribution from the structural steel analysis.
FSS-G6 Not Met Requirement is to quantify the risk contribution of any selected multi-compartment fire scenarios consistent with the FQ requirements. Quantification seems generally similar to FQ. It is noted that CCDPs are assumed to be 1 for combined areas. Multi-compartment fire CCDPs currently add up to mid-E-4 range, this is conservative. Need realistic evaluations for unscreened scenarios. No LERF information is provided, need to provide LERF.
Current CC: Met at CC II/III Section 5.3 and Attachment C of FPRA-PI-MCA (Revision 1). Scenarios that are not screened are included in the plant CDF and LERF quantification consistent with FQ requirements. Realistic estimates of multi-compartment CCDPs and CLERPs were developed following the May 2012 peer review. These new results have been incorporated into the Fire PRA and are reflected in the final quantification results for multi-compartment scenarios.
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SR Capability Category Topic Status
FSS-H5 CC I There were two (2) Findings that were issued against Supporting Requirement FSS-H2. Finding FSS-H5-01 identified that the Detailed Compartment Analysis notebook was cumbersome and difficult to interpret. Finding FSS-H5-02 identified that the Detailed Compartment Analysis notebook did not include detailed documentation related to scenario specific parameter uncertainty evaluations.
Current CC: Met at CC II As suggested, the Detailed Compartment Analysis Notebook, FPRA-PI-SCA, has been substantially rewritten in Revision 1 to address these concerns. The FPRA-PI-SCA notebook, section 4.2, identifies the specific uncertainties found in fire modeling in the PINGP fire PRA. In addition, the quantification notebook now includes the uncertainty analysis and has also been substantially rewritten in Revision 1 to FPRA-PI-FQ to better define and quantify uncertainty for the PINGP Fire PRA.
IGN-B5 Not Met Although there is a section in the Ignition Frequency notebook entitled Assumptions, it does not appear to be complete.
Current CC: Met at CC I/II/III Assumptions and sources of uncertainty have been updated in Revision 1 of the documentation for the ignition frequency and quantification notebooks, FPRA-PI-IGN and FPRA-PI-FQ. See also Table V-1.
FQ-A2 Not Met The quantification process currently assumes BOTH units experience a Reactor Trip with loss of PCS for EVERY scenario. Although it is reasonable to assume a reactor trip with PCS is initiated by each fire initiator, assuming a dual unit trip is overly restrictive, and is masking real risk insights. A review of "opposite unit" equipment failures needs to be performed to determine if a dual unit trip is realistic, and if it is not - then do not assume one.
Current CC: Met at CC I/II/III Since comprehensive cable tracing for all systems that could result in a plant trip has not been done, it cannot be assured that a fire in a compartment on one unit will not also affect the other unit. The prudent approach is to make a moderately conservative assumption to bound the range of possibilities.
FQ-E1 Not Met There are several identified fire scenarios that have been removed from the quantification results. These scenarios were removed to prevent the skewing of the importance calculations. This removal was noted in the notebook so that when detailed analysis of the removed scenarios was performed the removed scenarios would be included in the total CDF and LERF.
Current CC: Met at CC I/II/III The finding was based on the available quantification results at the time of the peer review, which have been superseded by results of an analysis that is more complete and includes more rigorous modeling techniques. All scenarios have been revised to add further detail and to credit additional fire modeling and
Northern States Power - Minnesota Attachment V – Fire PRA Quality
PINGP Page V-39 – Revision 1
Table V-2 Fire PRA–Summary of Category I and Not Met SRs
SR Capability Category Topic Status
mitigation as appropriate. All scenarios are now included in the final results.
FQ-F1 Not Met This SR is not met due to the fact that the Quantification notebook contains many areas that are not complete (Tables 5-9, 5-10, 5-11, etc.). There is a good discussion of the top cutsets.
Current CC: Met at CC I/II/III This Finding discusses documentation discrepancies related to the Fire Quantification (FQ) notebook at the time of the peer review. This document has since been updated to support the NFPA-805 LAR submittal and Supplement. Specific to the issues discussed in the peer review Finding, the parametric uncertainty analysis is covered in Section 6.0 of the Fire Quantification notebook (PI-FPRA-FQ, Revision 1), while the truncation analysis is discussed in Section 5.7 of the Fire Quantification notebook. Also, the qualitative uncertainties are discussed in Section 6.0 of the FQ notebook.
UNC-A1 Not Met Task discussions and Table 2 in UNC notebook provide identification of some sources of model uncertainty. However the team identified additional sources which should be explored. Task discussions and Table 2 in the UNC notebook provide identification of some assumptions. However, the team determined that assumptions were stated or implied in the various notebooks which were not addressed. The uncertainty intervals associated with parameter uncertainties were estimated and an estimate of the uncertainty interval of the CDF results was prepared. Some potential sources of uncertainty were neglected. Referenced SR QU-E4 states: “For each source of model uncertainty and related assumption identified in QU-E1 and QU-E2, respectively, IDENTIFY how the PRA model is affected (e.g., introduction of a new basic event, changes to basic event probabilities, change in success criterion, introduction of a new initiating event) [Note (1)].” This referenced element has not been satisfied. There are a number of assumptions and sources of uncertainty whose impact was not provided. LERF sources of model uncertainty were identified and characterized similar to CDF. Discussion related to CDF above applies equally to LERF.
Current CC: Met at CC I/II/III An expanded analysis of sources of uncertainties of the Fire PRA is given in Section 6.0 of Notebook FPRA-PI-FQ Revision 1 (sources of uncertainties specific to the internal events PRA are addressed separately as part the internal events PRA). The analysis uses an ordered approach where each task of the Fire PRA is reviewed to identify and characterize the sources of uncertainties. As part of that approach, the analysis discusses uncertainties associated with key elements of the Fire PRA, such as those cited in the finding, including, for example, heat release rates, non-suppression probabilities used, choice of target sets, treatment of fire-induced failure of instrument air, and treatment of conduits. In summary, the approach provides a structure to adequately identify and characterize the sources of uncertainty of the Fire PRA.
Northern States Power - Minnesota Attachment V – Fire PRA Quality
PINGP Page V-40 – Revision 1
Table V-2 Fire PRA–Summary of Category I and Not Met SRs
SR Capability Category Topic Status
UNC-A2 Not Met This SR says: INCLUDE the treatment of uncertainties, including their documentation, as called out in SRs PRM-A4, FQ-F1, IGN-A10, IGN-B5, FSS-E3, FSS-E4, FSS-H5, FSS-H9, and CF-A2 and that required by performing Part 2 referenced requirements throughout this Standard. Several of these back referenced requirements were not met. Statistical representations of the uncertainty intervals for the parameters used for modeling significant fire scenarios were identified for HRA, data, and ignition frequencies. However uncertainty information was not provided for other parameters, such as generic modeling factors. No characterization of uncertainties associated with cable routing was identified.
Current CC: Met at CC I/II/III The identification and characterization of sources of uncertainties is documented in Section 6.0 of Notebook FPRA-PI-FQ (Revision 1). An ordered approach was used where each task of the Fire PRA was reviewed and a qualitative characterization of the identified sources of uncertainty was given. This review included the characterization of the uncertainty associated with fire scenario development, which involved, for example, an evaluation of the treatment of fire impacts on conduits, instrument air piping, and more generally target damage. In addition, a quantitative analysis of uncertainty was performed for the parameters lending themselves to a statistical representation of uncertainty. The results of that quantitative evaluation are given in Section 6.0 of Notebook FPRA-PI-FQ. In summary, the analysis discusses the uncertainty of parameters at a level of detail sufficient for long-term use, upgrade, and peer review of the Fire PRA.
