R. Michael Glover · R. Michael Glover ENER H. B. Robinson Steam Electric Plant Unit 2 Site Vice...

R. Michael GloverH. B. Robinson SteamENER

Electric Plant Unit 2Site Vice President

Duke Energy Progress3581 West Entrance Road

Hartsville, SC 29550

0:843 857 1704F: 843 857 1319

Mike. Glo vert duke-energy.com,

Serial: RNP-RA/15-0021

APR 01 2015U. S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261 1 RENEWED LICENSE NO. DPR-23

SUPPLEMENTAL RESPONSE TO 120-DAY RESPONSE SUBMITTAL TO REQUEST FORADDITIONAL INFORMATION ASSOCIATED WITH LICENSE AMENDMENT REQUEST TOADOPT NATIONAL FIRE PROTECTION ASSOCIATION (NFPA) STANDARD 805

REFERENCES:

1. Letter from W. R. Gideon (Duke Energy Progress) to U. S. Nuclear Regulatory Commission(USNRC) (Serial: RNP-RA/1 3-0090), License Amendment Request (LAR) to Adopt NFPA805 Performance-Based Standard for Fire Protection for Light Water Reactor GeneratingPlants (2001 Edition), dated September 16, 2013, ADAMS Accession No. ML1 3267A21 1

2. Letter from Martha Barillas (USNRC) to Site Vice President, H. B. Robinson Steam ElectricPlant (Duke Energy Progress), H. B. Robinson Steam Electric Plant, Unit 2 - Request forAdditional Information on License Amendment Request to Adopt National Fire ProtectionAssociation Standard 805, Performance-Based Standard for Fire Protection (TAC No.MF2746), dated October 23, 2014, ADAMS Accession No. ML14289A260

3. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/14-0122), Response (60-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated November 24, 2014

4. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/14-0134), Response (90-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated December 22, 2014

5. Letter from R. Michael Glover (Duke Energy Progress) to U. S. Nuclear RegulatoryCommission (USNRC) (Serial: RNP-RA/1 5-0006), Response (120-Day) to Request forAdditional Information Associated with License Amendment Request to Adopt National FireProtection Association (NFPA) Standard 805, dated January 22, 2015

6. Letter from Martha Barillas (USNRC) to Site Vice President, H. B. Robinson Steam ElectricPlant (Duke Energy Progress), H. B. Robinson Steam Electric Plant, Unit 2 - Request forAdditional Information on 60-Day Response to License Amendment Request to AdoptNational Fire Protection Association Standard 805, Performance-Based Standard for FireProtection (TAC No. MF2746), dated March 26, 2015, ADAMS Accession No.ML15057A403

SDDo

U. S. Nuclear Regulatory CommissionSerial: RNP-RA/15-0021Page 2

Dear Sir/Madam:

By letter dated September 16, 2013 (Reference 1) Duke Energy Progress, Inc. submitted a licenseamendment request to adopt a new risk-informed performance-based fire protection licensing basisfor the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2).

During the week of September 22, 2014, the NRC conducted an audit at HBRSEP2 to supportdevelopment of questions regarding the license amendment request. On October 23, 2014 theNRC provided a set of requests for additional information regarding the license amendment request(Reference 2). That letter divided the requests for additional information into 60-day, 90-day, and120-day required responses. The Duke Energy Progress 60-Day, 90-Day, and 120-Day responseswere conveyed to the NRC Document Control Desk via letters from R. Michael Glover onNovember 24, 2014 (Reference 3), December 22, 2014 (Reference 4), and January 22, 2015,respectively. The NRC and Duke Energy Progress agreed per telecom on March 25, 2015 that theresponse to the Clarification RAIs, PRA RAI 05.a.01, 15.01, 18.01, 21.01, 23.d and 24.01, FM RAI01 .b.01 and 01 .c.01, SSA RAI 07.01 and LAR Attachments A, B and L would be submitted to theNRC by April 3, 2015. Enclosed as agreed are the Duke Energy Progress responses to therequests for additional information conveyed via Reference 6.

Please address any comments or questions regarding this matter to Mr. Richard Hightower,Manager - Nuclear Regulatory Affairs at (843) 857-1329.

There are no new regulatory commitments made in this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on2015.

Sincerely,

R. Michael GloverSite Vice President

RMG/jmw

Enclosure

cc: Mr. V. M. McCree, NRC, Region IIMs. Martha C. Barillas, NRC Project Manager, NRRNRC Resident Inspector, HBRSEP2Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

U. S. Nuclear Regulatory CommissionEnclosure to Serial: RNP-RAI15-0021201 Pages (including this cover page)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING VOLUNTARY FIREPROTECTION RISK INITIATIVE

REQUEST FOR ADDITIONAL INFORMATION

VOLUNTARY FIRE PROTECTION RISK INITIATIVE

DUKE ENERGY PROGRESS

H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2

DOCKET NO. 50-261

Safe Shutdown Analysis (SSA) Request for Additional Information (RAI) 07.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to SSA

RAI 07, item 5, and stated that current transformers (CTs) located in switchgear or components that arenot credited for safe shutdown (SSD) were excluded from the CTs' open secondary circuit analysis. The

purpose of evaluating the potential for causing an open circuit in a CT circuit is to determine the

potential for a secondary fire in a different fire area than the originating fire. The concept for asecondary fire due to an open circuited CT applies to any CT circuit, regardless of its use or credit forSSD. Therefore, the licensee's disposition that, "CTs located in switchgear or components that are not

credited for SSD were excluded," is not applicable.

Provide a response that addresses the potential fires caused by open circuited CTs, regardless

of whether or not the CT has been credited for SSD.

Response:

In the initial response to RAI-SSA-07, item 5, we stated "CTs located in switchgear or components that

are not credited for SSD were excluded".

The evaluation in EC 93120 of Current Transformers (CT) with turns ratios greater than 1200 to 5 usedthe statement for systems 5040 (Generator System) and 5120 (Switchyard and Transformer), that basedon assumption 4 (EC93120) these CTs will not cause damage to any Safe Shutdown relatedcomponents". The basis for this statement is; should an OPEN secondary circuit be created by a fireevent along the cable route, the cable breakdown due to the high voltage created by the CT would alsooccur at the point where the insulation is fire damaged. Should the secondary fire be caused bycatastrophic failure of the CT itself, as postulated in the same reference, the damage would then becontained within the CT enclosure and would not propagate outside the enclosure or damage anyadjacent equipment. (Ref section 6.2.3 of the PIRT Panel's review NUREG CR-7150, Volume 1, which is

attachment 4 of EC 93120).

Design consideration number 5 in the original RAI-SSA-07 will be revised to be more specific on thereasons which allow this group of CTs to be evaluated. It will read as follows:

5. The majority of the CTs installed in the plant with ratios >1200:5, are designed such that they

supply electrical circuitry (protective / indication) that is enclosed within the switchgear. Forthese cases, no "external to the switchgear" cables are utilized in the design.

Page 2 of 10

CTs with turns ratios >1200:5 located in switchgear or components that are not credited for SSD

but with secondary wiring that does extend into fire areas containing SSD equipment will not

cause damage to any Safe Shutdown related components. The basis for this statement is;should an OPEN secondary circuit be created by a fire event along the cable route, the cable

breakdown due to the high voltage created by the CT would also occur at the point where the

insulation is fire damaged. Should the secondary fire be caused by catastrophic failure of the CT

itself, as postulated in the same reference, the damage would then be contained within the CT

enclosure and would not propagate outside the enclosure or damage any adjacent equipment.(Ref section 6.2.3 of the PIRT Panel's review NUREGCR-7150, Volume 1, which is attachment 4 of

EC 93120).

Fire Modeling (FM) Request for Additional Information (RAI) 01.b.10

In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to FMRAI 01.b and explained how the effect of the increased heat release rate (HRR) due to fire propagationin cable trays was accounted for in the hot gas layer (HGL) and multi-compartment analysis (MCA)calculations. In its response, the licensee stated "Fire spread in each tray is assumed to be offset by theburnout."

Provide technical justification for this assumption as it is not consistent with the flame spread rates forthermoplastic and thermoset cables recommended in Appendix R of NUREG/CR-6850, "EPRI/NRC-RESFire PRA Methodology for Nuclear Power Facilities: Summary and Overview" and Chapter 9 ofNUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire(CHRISTIFIRE)."

In addition, the response does not address the potential effect on the zone of influence (ZOI) from theadditional HRR of the cable trays. Explain how the effect of the increased HRR due to fire propagation inthe cable trays was accounted for in the ZOI calculations; or provide technical justification for ignoringthis effect.

Response:

b. Based on the amount of combustibles (generally cable trays in the ZOI), the fire growth is

estimated using guidance from NUREG/CR-6850, Appendix R. The fire propagation among a

stack of vertical cables trays follows this general timeline:

TIMELINE TIMELINE DESCRIPTION

T, Time to build ignition source fire to scenario HRR = 12 minutesT2 Time to ignite the first target (tray) based on NED-M/MECH-1009

T, Time to ignite second tray = T2 + 4min

"1"4 Time to ignite third tray = T3 + 3min

T5 Time to ignite forth tray = T4 + 2min

T6 Time to ignite fifth tray = T5 + 1min

T, Time to ignite X tray = previous tray + 1min

The following properties are assigned to the horizontal cable fire growth for RNP based on

NUREG/CR-6850 and NUREG/CR-7010:

Page 3 of 10

Cable Tray Width 0.61 m Typical tray width

HRR per unit area 250 kW/m 2 NUREG/CR-7010, section 10.1

for thermoplastic cables

Using the values listed above, the heat release rate for the cable trays is calculated as the

surface area of the tray multiplied by the heat release rate per unit area. The angle of 35'described in Appendix R.4.2 of NUREG/CR-6850 is used for determining the length of the cable

trays in the stack above the ignition source so that the appropriate burning surface for each tray

is determined.

The total fire growth is based on adding the source fire HRR plus each tray HRR per unit time.

Fire spread in each tray is assumed to be offset by the burnout. If the fire grows large enough to

support a HGL, the time to HGL can be estimated. An adjustment was made to the process of

calculating the time to HGL by using cumulative HRR by comparing the energy required to

produce an HGL to the total energy produced by the fire. The MCA was performed in the same

manner as the HGL analysis.

"Fire spread in each tray is assumed to be offset by the burnout." means that as the progression

of the fire extends outwards, the burning region remains somewhat constant. In other words,

the flame spread outward along the cables will be equal to flame extinguishment along the just

consumed cables. This is shown in Figure 9-2 of NUREG/CR-7010, "Cable Heat Release, Ignition,

and Spread in Tray Installations During Fire (CHRISTIFIRE)."

If there are significant intervening combustibles, the total HRR of the fire will grow, expanding

the ZOI. The most probable increase in damage will be above the fire. The vertical ZOI would be

extended to the ceiling if multiple cable trays exist in the vertical ZOI for the source.

The process for determining the heat release rate associated with intervening combustibles

(e.g., those cable trays within the zone of influence of an ignition source) is summarized in

section 5.6.2 of Fire PRA calculation RNP-F/PSA-0094. The heat release rate for the cable trays

that are identified during walk downs is calculated assuming:

1. A cable tray width of 2 feet (0.61 m). Typical cable tray widths range from 6-inches to 3 feet

wide. A 2 foot wide tray was selected to represent for all RNP fire zones. This is consistent

with the average cable tray size as presented in NUREG/CR-6850, section R.4.2.1.

2. Initial cable tray burning length of approximately 3 feet (1 m). This value is representative of

the fire diameter for the ignition source, which is also the length of the first tray burning in

the stack.

3. A heat release rate per unit area of 250 kW/m 2 . This heat release rate value is

recommended in NUREG/CR-7010 for Thermoplastic cables.

Using the values listed above, the heat release rate for the cable trays is calculated as the

surface area of the tray multiplied by the heat release rate per unit area. The angle of 350

described in Appendix R of NUREG/CR-6850 is used for determining the length of the cable trays

in the stack above the ignition source so that the appropriate burning surface for each tray is

determined.Page 4 of 10

FM RAI 01.c.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to FM

RAI 01.c and explained that intervening combustibles within the ignition source ZOI were identified in

the walkdowns, and that the HRR contribution from these combustibles were incorporated in the fire

modeling analyses. It appears that the ZOI calculations were only performed for 69 kW, 211 kW, 317

kW, and 702 kW fires. This would imply that the licensee may not have accounted for the additional

HRR from non-cable intervening combustibles on the ZOI.

Confirm that the effect of the increased HRR from non-cable intervening combustibles was indeed

accounted for in the ZOI calculations; or provide technical justification for ignoring this effect.

In addition, the licensee did not provide any details on how the HRR of non-cable intervening

combustibles was calculated. Describe themethodology that was used to estimate the HRR from non-

cable intervening combustibles in the ZOI, HGL, and MCA calculations.

Response:

c. To ensure that intervening combustibles (including non-cable intervening combustibles and

cables that are not targets in the Fire PRA) are properly accounted for in the fire modeling

analysis supporting the Fire PRA, walk down project instructions were followed:

" FPIP-0200, Rev. 8, Fire PRA Walk down Instructions. This procedure provides specific

guidance on dealing with intervening combustibles. The guidance consists of identifying the

intervening combustibles within the zone on influence and capturing them in the walk down

forms. Possible intervening combustibles included cables, trays, batteries/chargers, panels,

and equipment, etc.

* FPIP-0208, Rev. 5, Scoping Fire Modeling. This procedure provides guidance to consider

cable trays above the ignition sources up to the enclosure ceiling to prevent missing the

contribution from these cable trays to the heat release rate.

The practical implications of the guidance included in the project instructions listed above are that the

RNP Fire PRA includes the contribution from two types of fire scenarios from the perspective of this RAI.

These fire scenarios are:

" Those fire scenarios affecting the ignition source only. That is, there is no fire propagation

outside the ignition source. No propagation outside the ignition source is due to either no

intervening combustibles within the zone of influence, or credit to passive fire protection

features such as solid bottom trays.

* Those fire scenarios where fire propagates throughout the zone of influence. If during walk

downs, intervening combustibles were identified within the zone of influence, the heat

release rate contribution from these combustibles was included as part of the heat release

rate profile characterizing the fire scenario. The zone of influence extends to the ceiling of

the physical analysis unit.

i. Non-cable intervening combustibles were not found during RNP walkdowns, so there was no

effect on the scenario HRR.

Page 5 of 10

ii. Non-cable intervening combustibles were not found during RNP walkdowns, so there was no

effect on the ZOI, HGL, and MCA calculations.

Probabilistic Risk Assessment (PRA) RAI 05.a.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRARAI 05.a and explained that, as part of the integrated analysis provided in response to PRA RAI 3,scenario development of fire propagation from electrical cabinets greater than 440V will be based ondraft Frequently Asked Questions (FAQ) No. 14-0009. The U.S. Nuclear Regulatory Commission (NRC)staff has provided comments on the treatment of motor control centers (MCCs) in "NRC Comments onMCC Treatment White Paper August 29, 2014" (ML14245A133) on the draft FAQ 14-0009. These NRCstaff comments recognize that the basis for the 0.1 multiplier is a 0.19 estimate of the probability of firebreaching a well-sealed cabinet (based on an evaluation of operating experience) and a 0.45 factor oflikelihood of damage to targets outside of the cabinet (based on a phenomenological fire modelingevaluation of the geometry of fires within MCC cabinets). There is no provision to change the 0.19 valuebut the method allows for re-evaluation of the 0.45 factor using phenomenological modeling of plant-specific cabinet geometry and ignition sources.

1. Provide the technical justification for any revised values used to represent the probability of

damage to targets located outside the MCC cabinet that deviate from accepted values. If thebasis for the revised values relies on phenomenological fire modeling that has not beenreviewed by the NRC, include a detailed discussion of the modeling that was performed and the

results that support the revised values.

2. Describe how these fires will be modeled in the integrated analysis provided in response to PRA

RAI 3.

Response:

With respect to the NRC staff comments (dated August 29, 2014, ML14245A133) on an earlier draftFPRA FAQ 14-0009, draft H of FPRA FAQ 14-0009, clarifies the scope of panel breaching due to energeticarcing faults to be limited to "well-sealed" MCCs greater than 440V, provides an expanded basis forusing a breaching factor of 0.21 and a severity factor of 0.454 to obtain a 0.095 multiplier that a firebreaches a well-sealed MCC and damages nearby (within six inches) external targets, and presents a firemodeling discussion based on NUREG/CR-6850 models to support credit for fire growth and propagationdamaging subsequent targets. While Fire PRA FAQ 14-0009 has not yet been accepted by the NRC, draftH represents the best available guidance produced by considerable collaborative efforts.

1. In the RNP FPRA, well-sealed MCC electrical cabinets greater than 440V were treatedconservatively relative to the guidance in draft H of FPRA FAQ 14-0009. This treatment of arcingfaults, which are not to be confused with HEAFs, is not applicable to non-MCC electricalcabinets. Although draft H of FPRA FAQ 14-0009 permits a case-by-case re-evaluation of the0.454 severity factor when the distance between the closest external target and the MCC ofinterest is more than six inches, the resultant multiplier would be proportionally less than the0.1 which was used by the RNP FPRA.

2. In the response to PRA RAI 3, these fires are included as part of the base model.

Any subsequent change in the accepted version of FPRA FAQ 14-0009 would be addressedthrough the normal PRA maintenance process.

Page 6 of 10

PRA RAI 15.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRA

RAI 15 and explained that inadequate breaker fuse coordination is accounted for in the Fire PRA. Theresponse explains that the "cables causing issues for power supplies" were identified, along with theirrouting, and used "to create an assumed failures list for breaker coordination." Though it appears thatcommon power supply failures were modeled in the PRA, it is not clear whether possible ignition of

affected cables (common enclosure) is considered. The Fire PRA modeling should be consistent withNUREG/CR-6850, "Fire Probabilistic Risk Assessment Methodology for Nuclear Power Facilities,"guidance, Section 3.5.4.2, Step 4.2, which states that, "[i]n evaluating the adequacy of cable thermalprotection, the criteria for acceptance should be based on a secondary fire concern and not simply

exceeding the continuous or overload thermal limit for the cable."

Clarify how the risk associated with secondary fires from cables between uncoordinated breakers isaddressed and describe how these fires will be modeled in the integrated analysis provided in responseto PRA RAI 3. Alternatively, complete breaker coordination work prior to self-approval, updateAttachment S of the LAR as necessary, and discuss how it would be reflected in the Fire PRA for the

response to PRA RAI 3.

Response:

Breaker coordination was reviewed for Robinson. In some cases breaker coordination was achieved by

crediting cable length for the load, in others this credit could not be applied for the NFPA-805 analysis

due to the location of the fire. The upstream power supply was assumed to trip, causing the loss of

power to additional equipment as a result. Cable protection against overload and short circuit is a

consideration in the general design criteria for protective device selection, however is limited in

application. The Fire PRA currently models the upstream power trip, but does not model secondary fires

resulting from a lack of adequate protection.

PRA RAI 18.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to PRA

RAI 18 and explained that treatment of self-ignited and cutting-and-welding fires is consistent with FAQ

No. 13-0005 for one fire compartment, and for other compartments, these fires were screened out

because they were determined to have "insignificant impact." In apparent contrast to this, the response

PRA RAI 08 states "cable fires due to cutting and welding are assigned no target sets because a

continuous fire watch with an extinguisher is required by procedure to be present during hot work

activities and is assumed to extinguish such a fire before it can spread beyond the original tray." The

cited responses to PRA RAI 08 and PRA RAI 18 appear to be inconsistent.

Clarify this apparent inconsistency and identify how these cable fires will be modeled in the integrated

analysis provided in response to PRA RAI 3.

Page 7 of 10

Response:

The statement in the response to PRA RAI 08 (Reference Robinson Letter RNP-RA/14-0122) is consistent

with Fire PRA FAQ 13-0005, as it assumes damage is limited to a single tray and no additional target sets.

For Robinson, a bounding approach was initially used to assess all cable fires due to cutting and welding

by conservatively applying the limiting source scenario CCDP or CLERP (not limited to a single tray) for

the applicable compartment to the cutting and welding scenario, effectively assuming the cutting and

welding scenarios had a target set equal to the most limiting scenario. If the bounding treatment

resulted in significantly high results, then a tray-by-tray assessment was performed per Fire PRA FAQ 13-

0005 to identify the limiting tray CCDP and CLERP for the compartment. The tray-by-tray assessment

was only needed for Fire Compartments 250 (Turbine Building) and 70 (Lower Hallway). This treatment

of the cable fires is included in the quantification results presented in the response to PRA RAI 3.

PRA RAI 21.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML 15079A025), the licensee responded to PRA

RAI 21 and based on the response, fires that are not believed to cause an automatic trip are assigned a

conditional probability of manual trip that reduces the likelihood of the associated fire scenario.

1. Discuss whether your review of the fire-induced initiating events is consistent or

conservative compared to the review steps described in NUREG/CR-6850 guidance in

section 2.5.3, "Step 3: Identify Fire-induced Initiating Events Based on Equipment Affected."

2. Discuss any Fire PRA model changes after completing these evaluations that will be included

in the integrated analysis provided in response to PRA RAI 3.

Response:

1. The review of fire-induced initiating events is consistent with the guidance in Section 2.5.3

of NUREG/CR-6850 in that a reactor trip was assigned as the initiator for fires that are not

believed to cause an automatic trip. Additionally, although the guidance in section 2.5.3

does not require the assignment of an initiating event for fire compartments where none of

the three conditions is judged to occur (i.e., no automatic, manual, or LCO forced trip), a

conditional trip probability is conservatively applied to account for operator discretion to

perform a manual trip even for a fire in an area containing no equipment important to plant

operations.

2. Consequently, no related Fire PRA model change has been included in the integrated

analysis provided in response to PRA RAI 3.

Page 8 of 10

PRA RAI 23.d

Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the

NRC. Section 2.4.4.1 of NFPA-805 further states that the change in public health risk arising from

transition from the current fire protection program to an NFPA-805 based program, and all future plant

changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on

CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to

the plant's licensing basis and describes a general framework to determine the acceptability of risk-

informed changes. The NRC staff review of the information in the LAR has identified the following

information that is required to fully characterize the risk estimates.

Section W.2.1 of the LAR provides some description of how the change-in-risk and the additional risk of

recovery actions associated with VFDRs is determined but not enough detail to make the approach

completely understood. Provide the following:

d) A description of the type of VFDRs identified, and discuss whether and how the VFDRs

identified, but not modeled in the Fire PRA, impact the risk estimates. Include any qualitative

rational for excluding VFDRs from the change-in-risk calculations.

Response:

As per Attachment C of the LAR, VFDRs are characterized as two types. Type 1 VFDRs are forunprotected cables, where damage to these identified cables would cause failure to meet thedeterministic requirements of NFPA 805. Type 2 VFDRs are the dedicated shutdown recovery actions

used in the event the control room is abandoned due to a fire. The actions taken at a remote shutdownlocation that does not meet the definition of a primary control station are considered VFDRs.

Some Type 1 VFDRs identified in Table B-3, "Fire Area Transition," of Attachment C were dispositionedas "Not a VFDR." The VFDRs dispositioned as "Not a VFDR" are not included in the Fire PRA. TheseVFDRs were determined to be "Not a VFDR" within the Nuclear Capability Assessments (NSCA) with thefollowing justifications:

* There is a redundant component available to support the ability to achieve and maintain thenuclear safety performance criteria.

* Failure of the component can be recovered by action(s) taken in the MCR.

* Fire damage to the component will not prevent the component from performing its nuclear

safety function.

Excluding VFDRs dispositioned as "Not a VFDR" from the change-in-risk calculations were not based on

qualitative rationale.

The VFDRs dispositioned as "Not a VFDR" will be removed from LAR Attachment C. An updated LAR

Attachment C will be provided with response to PRA RAI 03.

Page 9 of 10

PRA RAI 24.01

In a letter dated March 16, 2015 (ADAMS Accession No. ML15079A025), the licensee responded to PRARAI 24 and in apparent contrast to the response to PRA RAI 24.e, the response to PRA RAI 01.f indicatesthat there are actions taken in the plant at the remote shutdown locations to recover equipment

affected by fire not associated with main control room (MCR) abandonment that are credited in the FirePRA. Attachment G and the response to PRA RAI 24.e seem to indicate that these actions aredesignated as defense-in-depth actions. Actions taken in the plant at the remote shutdown locations to

recover equipment affected by fire not associated with MCR abandonment may be recovery actions,

since command and control is not established at the remote shutdown panel.

Clarify whether these actions should be considered recovery actions as discussed in Regulatory Guide1.205, or discuss the rationale for their designation as defense-in-depth. If these actions should be

recovery actions, discuss how the additional risk of recovery actions for the applicable scenarios in

which the MCR is not abandoned will be modeled in response to PRA RAI 3.

Response:

The response to PRA RAI 24.e (Reference Robinson Letter RNP-RA/14-0122) has been updated to clarifythat RA-DIDs were categorized as such due to low risk significance. The process for determining thisoutcome is based on assessing the delta risk contribution from modeled actions in the FPRA.

For the Robinson NFPA 805 LAR, recovery actions are identified for operator actions performed in theplant at locations other than in the Main Control Room or at primary control stations, which are

identified in Attachment G. The primary control stations (PCSs) are the Dedicated Shutdown DieselControl Panel, Secondary Control Panel on the Turbine Deck, and the Charging Control Panel in the

Charging Pump Room. Therefore, operator actions identified in Attachment G that are credited in theFPRA and performed at locations other than the MCR or PCS are classified as "recovery actions". Itfollows that actions taken at the remote alternate control stations, for non-abandonment, are "recoveryactions", given that they are activities outside of the MCR to achieve the nuclear safety performance

criteria.

The change in risk (delta risk) for a recovery action is modeled by the difference between CDF/LERF forthe recovery actions based on the non-compliant (or variant) base case, which applies the nominal

human error probability (HEP), and the compliant case, which has the HEP set to zero (guaranteedalways successful).

The categorization of RA or RA-DID for recoveries modeled in the FPRA is based on the magnitude of the

change in risk of the recovery action.

The results of this analysis are provided in LAR Table W-5.

Page 10 of 10

Duke Energy Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements

A. NEI 04-02 Table B-1 Transition of Fundamental Fire ProtectionProgram & Design Elements

79 Pages Attached

HBRSEP LAR Rev I Page A-I II

HBRSEP LAR Rev 1 Page A-1 I

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Energy

Table B-1 NFPA 805 Ch.3 Transition Details

Chapter 3 Reference: 3.1 General

Chapter 3 Requirement: 3.1* General.

This chapter contains the fundamental elements of the fire protection program and specifies the

minimum design requirements for fire protection systems and features. These fire protection program

elements and minimum design requirements shall not be subject to the performance-based methodspermitted elsewhere in this standard. Previously approved alternatives from the fundamental protectionprogram attributes of this chapter by the AHJ take precedence over the requirements contained herein.

Compliance Statement

N/A

Compliance BasisN/A - General statement; No technical

requirements.


Chapter 3 Reference: 3.2 Fire Protection Plan

Chapter 3 Requirement: N/A

Compliance Statement Compliance Basis

N/A N/A - General statement; No technical

requirements.


Chapter 3 Reference: 3.2.1 Intent

Chapter 3 Requirement: 3.2.1 Intent.

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 impleme t all fire protection program activities.

Complance statement

Complies

.ompliance 3as sNo Additional Clarification

Reference Document Doc Detwl

HBRSEP LAR Rev 1 Page A-2


OMM-002,Fire Protection Manual ALL


Chapter 3 Reference: 3.2.2 Management Policy Direction and Responsibility.

Chapter 3 Requirement: 3.2.2* Management Policy Direction and Responsibility.

A policy document shall be prepared that defines management authority and responsibilities and

establishes the general policy for the site fire protection program.


Complies No Additional Clarification

Reference Document DoDetals

OMM-002,Fire Protection Manual Section 3


Chapter 3 Reference: 3.2.2.1 [Management Policy on Senior Management]

Chapter 3 Requirement: 3.2.2.1*

The policy document shall designate the senior management position with immediate authority and

responsibility for the fire protection program.



Reference Document DocDetails

OMM-002,Fire Protection Manual Section 3.1 & 3.2


Chapter 3 Reference: 3.2.2.2 [Management Policy on Daily Administration]

Chapter 3 Requirement:


3.2.2.2*

The policy document shall designate a position responsible for the daily administration and coordination

of the fire protection program and its implementation.

Compliance Basis


Duke EnergyComplies

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design Elements

No Additional Clarification

Reference Document Doc Details

OMM-002,Fire Protection Manual Section 3.7


Chapter 3 Reference: 3.2.2.3 [Management Policy on Interfaces]


The policy document shall define the fire protection interfaces with other organizations and assignresponsibilities 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.


Complies

Compliance Basis



OMM-002,Fire Protection Manual Section 3


Chapter 3 Reference: 3.2.2.4 [Management Policy on AHJJ


The policy document shall identify the appropriate AHJ for the various areas of the fire protection

program.



Reference Document DocDetamls

OMM-002,Fire Protection Manual Section 4.1.7

l BlTable B-i1 NFPA 805 Ch.3 Transition Details

Chapter 3 Reference: 3.2.3 Procedures



Chapter 3 Requirement: 3.2.3* Procedures.

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.


Main Header: Complies

Section (1):License Amendment Required

Compliance BasisMain Header: No Additional Clarification

Section (1): See Attachment L for

Surveillance Optimization.


OMM-002,Fire Protection Manual ALL

FP-012,Fire Protection Systems Minimum Equipment and ALL

Compensatory Actions

FP-013,Fire Protection Systems Surveillance Requirements ALL

Chapter 3 Requirement: 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.


Section (2): Complies Section (2): No Additional Clarification

Reference Document Doc Dels

FP-012,Fire Protection Systems Minimum Equipment and ALL

Compensatory Actions


Chapter 3 Requirement: (3) * Reviews of fire protection program - related performance and trends.



Reference Document DocDetaih

EGR-NGGC-0008,Engineering Projrams ALL

EGR-NGGC-0010,System & Compcnent Trending Program and Section 1.1 & Enclosure 1

System Notebooks

Chapter 3 Requirement: (4) Reviews of physical plant modifications and procedure changes for impact on the fire protection



program.


Section (4): Complies

Compliance Basis

Section (4): No Additional Clarification

Reference Document

OMM-002,Fire Protection Manual

EGR-NGGC-0003,Design Review Requirements

EGR- NGGC-0005,Engineering Change

PRO-NGGC-0204,Procedure Review and Approval

EGR-NGGC-0102,Safe Shutdown/Fire Protection Review

REG-NGGC-0010,10 CFR 50.59 AND SELECTED REGULATORY

REVIEWS

Doc Dgtails

Section 3

ALL

ALL

ALL

ALL

ALL

Chapter 3 Requirement: (5) Long-term maintenance and configuration of the fire protection program.



Reference Document DocDetaols

EGR-NGGC-0102,Safe Shutdown/Fire Protection Review ALL

EGR-NGGC-0003,Design Review Requirements ALL

EGR-NGGC-0005,Engineering Change ALL

Chapter 3 Requirement: (6) Emergency response procedures for the plant industrial fire brigade.



Reference Document Doc Detall,

FP-001,Fire Emergency ALL

AOP-041,Response to Fire Event ALL


Chapter 3 Reference: 3.3 Prevention

Chapter 3 Requirement: 3.3 Prevention.

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


Duke EnergyAttachment A - NEI 04-02 Table B-1 Transition of Fundamental

Fire Protection Program & Design Elementsfire prevention program shall consist of both of the following:

(1) Prevention of fires and fire spread by controls on operational activities.




Compliance Basis

Main Header: No Additional Clarification


Reference Document DcDeils


FIR-NGGC-0003,Hot Work Permit ALL

Chapter 3 Requirement: (2) Design controls that restrict the use of combustible materials

The design control requirements listed in the remainder of this section shall be provided as described.





FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND ALL

IGNITION SOURCE CONTROLS PROGRAM

EGR-NGGC-0005,Engineering Change Attachment 3, Section A3.0.7 & Attachment 2,

Section A3.3.24


Chapter 3 Reference: 3.3.1 Fire Prevention for Operational Activities.

Chapter 3 Requirement: 3.3.1 Fire Prevention for Operational Activities.

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.


Complies

Compliance Basis


Reference Document

FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND


DocDetaoll

ALL

Section 8.4, ALLOMM-002,Fire Protection Manual


Duke EnergyFIR-NGGC-0003,Hot Work Permit


ALL


Chapter 3 Reference: 3.3.1.1 General Fire Prevention Activities.

Chapter 3 Requirement: 3.3.1.1 General Fire Prevention Activities.

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.


Main Header: Complies Main Header: No Additional Clarification



OMM-002,Fire Protection Manual Attachment 10.1 & Section 3.16

GNI0008N,Initial General Employee Training - Contractors Computer Based Training (CBT)

GETSSG,General Employee Training Self Study Guide ALL

GNBO1N,Initial General Employee Training - Progress Energy Computer Based Training (CBT)

Personnel

Chapter 3 Requirement: (2) * Documented plant inspections including provisions for corrective actions for conditions where

unanalyzed fire hazards are identified.




FP-010,Housekeeping Controls Section 8.2 Attachment 10.2

Chapter 3 Requirement: (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.


Section (3): Complies Section (3): No Additionrl Clarification

Reference Document

EGR-NGGC-0005,Engineering Change

Doc Details

Attachment 3, Section A3.0.7 & Attachment 2,

Section A3.3.24


Duke EnergyOMM-002,Fire Protection Manual

WCP-NGGC-0300,Work Request Initiation, Screening, Prioritization,

and Classification

EGR-NGGC-0102,Safe Shutdown/Fire Protection Review

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsSection 3.8.2

Section 9.2.1.h

ALL

Table B-I NFPA 805 Ch.3 Transition Details

Chapter 3 Reference: 3.3.1.2 Control of Combustible Materials





3.3.1.2* Control of Combustible Materials.

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 Basis




FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.1.8


FP-010,Housekeeping Controls ALL

Chapter 3 Requirement: (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.


Section (2): Complies Section (2): Complies






(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, whichever comes first.

Compliance Basis


Duke EnergySection (3): Complies



Reference Document DoDals



Chapter 3 Requirement: (4) * Combustible storage or staging areas shall be designated, and limits shall be established on the

types and quantities of stored materials.



Compliance Basis



FP-010,Housekeeping Controls Attachments 10.1 -10.3

Chapter 3 Requirement: (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.




FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4


FAQ 06-0020,Identification of "applicable NFPA standards" ALL

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1 B-9

Chapter 3 Requirement: (6) * Controls on use and storage of flammable gases shall be in accordance with applicable NFPA

standards.



Compi2:ance Basis



FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND ALL


FP-006,Handling of Flammable Liquids and Gases ALL

FAQ 06-0020,ldentification of "applicable NFPA standards" ALL


HBRSEP LAR Rev 1 Page A-1 0


Chapter 3 Reference: 3.3.1.3 Control of Ignition Sources

Chapter 3 Requirement: 3.3.1.3 Control of Ignition Sources



requirements.


Chapter 3 Reference: 3.3.1.3.1 [Control of Ignition Sources Code Requirements]

Chapter 3 Requirement: 3.3.1.3.1*

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.


Complies COMPLIES: No Additional Clarification

Complies with Clarification COMPLIES WITH CLARIFICATION:

Compliance with NFPA 241 is by

clarification and is addressed through

compliance with NFPA 51B. NFPA 241,

2009 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot

work, states "Responsibility for hot work

operations and fire prevention precautions

, including permits and fire watches, shall

be in accordance with NFPA 51B,

Standard for Fire Prevention During

Welding, Cutting, and Other Hot Work."

Reference Document DocDetals


NFPA 241 ,Standard for Safeguarding Construction, Alteration, and Section 5.1

Demolition Operations, 2004 Edition

NED-M/BMRK-0001 ,Code Compliance Evaluation for NFPA 51B, ALL

Standard for Fire Prevention during Welding, Cutting, and Other Hot

Work- 1999 Edition




Chapter 3 Reference: 3.3.1.3.2 [Control of Ignition Sources on Smoking Limitations]

Chapter 3 Requirement: 3.3.1.3.2

Smoking and other possible sources of ignition shall be restricted to properly designated and

supervised safe areas of the plant.



Reference Document DocMlls

NO-80-169,Revision to the Administrative Controls for Fire Enclosure No. 4, Page 2

Protection, 2/1/1980

FP-010,Housekeeping Controls Section 5.3



Chapter 3 Reference: 3.3.1.3.3 [Control of Ignition Sources for Leak Testing]

Chapter 3 Requirement: 3.3.1.3.3

Open flames or combustion-generated smoke shall not be permitted for leak or air flow testing



Reference Document DoDetail

FIR-NGGC-0003,Hot Work Permit Section 6.15


Chapter 3 Reference: 3.3.1.3.4 [Control of Ignition Sources on Portable Heaters]

Chapter 3 Requirement: 3.3.1.3.4*

Plant administrfative procedure shall control the use of portable electrical heaters ih 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.


CompliesCompliance Basis


HBRSEP LAR Rev 1 Page A- 12

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elements

Reference Documentc Deails



FIR-NGGC-0003,Hot Work Permit Section 6.13


Chapter 3 Reference: 3.3.2 Structural.

Chapter 3 Requirement: 3.3.2 Structural.

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.




UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1.B-6


Chapter 3 Reference: 3.3.3 Interior Finishes

Chapter 3 Requirement: 3.3.3 Interior Finishes.

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 I interior floor finishes.


Complies

Compliance Basis


Reference Document

NFPA 101,Life Safety Code, 2009 Edition

CPL-)(XXX-W-005,Nuclear Powir Plant Protective Coatings

L2-C-007,Field Coatings

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR)

FPP-RNP-900,Fire Hazards Analysis

Doc Details

(a) Sections 10.2.3.4 & 10.2.7.4

ALL

ALL

HBRSEP LAR Rev 1 Page A- 13

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFirn Prnfr~rtinn Prngrnm P Q. 1"aain Ilementnihikp Fnir0Y


Chapter 3 Reference. 3.3.4 Insulation Materials

Chapter 3 Requirement: 3.3.4 Insulation Materials.

Thermal insulation materials, radiation shielding materials, ventilation duct materials, and

soundproofing materials shall be noncombustible or limited combustible.




UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1 .B-6

CPL-HBR2-M-025,Heating, Ventilation, and Air Conditioning (HVAC) Section 11-2.01

Main Plant Fabrication and Installation

CPL-HBR2-M-028,Specification for RHR Pump Pit to HVE-5 Exhaust Section 11-2.01

Tie-In Fabrication and Installation

L2-M-039,Piping and Equpment Thermal Insulation Section 4.4.1.4

GID/87038-0014,Fire Barrier System ALL


Chapter 3 Reference: 3.3.5 Electrical.




requirements.


Chapter 3 Reference: 3.3.5.1 [Electrical Wiring Above Suspended Ceiling Limitations]

Chapter 3 Reauiremet: 3.3.5.1 1

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.


Duke EnergyCompliance Statement

License Amendment Required

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsCompliance Basis

NRC approval is being requested in

Attachment L for electrical wiring above

suspended ceilings that may not comply

with the requirements of NFPA 805.

Reference Document


Doc Details

Appendix 9.5.1B-7


Chapter 3 Reference: 3.3.5.2 [Electrical Raceway Construction Limits]

Chapter 3 Requirement: 3.3.5.2

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.


License Amendment Required NRC approval is being requested in

Attachment L for electrical raceway

construction at HBRSEP that may not

comply with the requirements of NFPA

805.

Reference Document

HBR2-0B060 Sht D6,Electrical Installation Practices, Notes and

Details

HBR2-0B060 SH D2,Electrical Installation Practices, Notes and

Details

HBR2-0B060 SHC3,ELECTRICAL INSTALLATION PRACTICES,

NOTES AND DETAILS

DBD/R87038/SD62,Design Basis Document Cable and Raceway

System

DALDetals

ALL

ALL

ALL

ALL


Chapter 3 Reference: 3.3.5.3 [ElectricalCable Flame Propagation Limits]


Electric cable construction shall comply with a flame propagation test as acceptable to the AHJ.


Duke Energy


Complies with Clarification

Complies via Previous NRC Approval

Complies via Engineering Evaluation


Compliance BasisCOMPLIES WITH CLARIFICATION: FAQ

06-0022 evaluates currently recognized

flame propagation tests to the IEEE 383-

1974 Standard, the US NRC minimum test

standard, and acceptance criteria for cable

flame propagation tests. Table 2 in

'Summary of Results' section of FAQ 06-

0022 provides a summary of the testing

methods that are more severe than IEEE

383-1974. Non-IEEE-383-1974 qualified

cables used at RNP are IEEE-383-1974

equivalent since they meet the cable

standards identified in Table 2 of FAQ 06-

0022 except for some original PVC

jacketed cabling. Depending on when the

PVC jacketed cables were installed they

might not have met the requirements of

IEEE-383-1974 or equivalent. These

original cables not meeting the

requirements of IEEE-383-1974 or

equivalent were coated with fire retardant

material which meets or exceeds the

original cable coating requirements to

prevent propagation of fire.

COMPLIES VIA PREVIOUS NRC

APPROVAL: In the SER dated 2/28/78,

the NRC stated the following:

"4.8 Electrical Cables

In the plant areas outside the containment,

cable jacket and insulation material is

polyvinyl chloride. Inside the containment,

cable insulation is silicone rubber. The

flame test standard for cables IEEE 383

was not in effect at the time electrical

cables were purchased and installed at H.

B. Robinson. Cables in critical areas,

inside and outside containment will be

coated with a flame retardant coating.

Detailed discussion of these areas can befound in Section 5.0 of this report."

Section 5.0 listed the following areas

where cables would L`e coated:

Safety Injection Pump Room (Fire Area

No. 3)

Component Cooling Water Pump Room

(Fire Area No. 5)

Aux Feedwater Pump room (Fire Area No.



7)Cable Vaults (Fire Areas No. 9 and No. 34)

Aux Building Hallway, Lower Level (Fire

Area 10A, 10B, 10C)

Aux Building Hallway, Upper Level (Fire

Areas 14A thru 14G, except 14D)

Unit 2 Cable Spreading Room, Computer

Room (Fire Area No. 18)

Electrical Equipment Area (Fire Area No.

19)

Rod Control Room (Fire Area No. 20)

In the evaluation dated 2/21/80, the NRC

stated:

"3.1.4 Fire Retardant Cable Coating

Fire retardant coating will be applied to

cables located in 13 different fire areas of

the plant (4.8)."

By letter dated December 5, 1978, the

licensee stated that the flame-retardant

coating would be applied in accordance

with manufacturer's recommendations and

that the manufacturer would be consulted

to determine alternate application methods

for situations not covered by the

manufacturer's standard

recommendations.

We accept the licensee's proposal."

Per the 2/21/80 evaluation, the status of

Fire Retardant Coating was "Complete".

Proposed modifications were evaluated

and implemented per the SERs, where

applicable, to fulfill the intent of this

requirements.

There have been no plant modifications or

other changes that would invalidate the

basis for approval.

COMPLIES VIA ENGINEERING

EVALUATION: Engineering evaluations

EE-84-0043, EE-90-0037, EE-92-0090,

and NED-I3/BOP-1001 are applicable to

electric cable construction at HBRSEP.



Duke EneNLU-78-7A,License Amendment 31

NLU-80-106,RFI and Requirements to Resolve Issues Concerning

Fire Protection


FAQ 06-0022,Acceptable Electrical Cable Construction Tests

DBD/R87038/SD62,Design Basis Document Cable and Raceway

System

EE-84-0043,Qualification of Rockbestos Fire Zone R Cable to IEEE-

383 Vertical Flame Test

EE-90-0037,Evaluation Of Use Of Non-IEEE 383, Vertical Flame

Test Cable Proposed By Modification M-1001

EE-92-0090,Evaluation Of Abandoned Cables (Belden # 8424) Inside

Containment (General Area)

NED-B/BOP-1001,Comparative Analysis of Fire Propagation

Characteristics of UL-910 and IEEE-383 Qualified Cables

Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsSection 4.8 & 5.0

Section 3.1.4

Section 9.5.1.4.4.4.2

ALL

Sections 3.5.1.3.3 and 3.5.1.3.5

ALL

ALL

ALL

ALL


Chapter 3 Reference: 3.3.6 Roofs.

Chapter 3 Requirement: 3.3.6 Roofs.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 ofFire Tests of Roof Coverings.


Complies

Compliance Basis


Reference Document


Doc Details

Appendix 9.5.1B-6


Chapter 3 Reference: 3.3.7 Bulk Flammable Gas Storage.



3.3.7 Bulk Flammable Gas Storage.

Bulk compressed or cryogenic flammable gas storage shall not be permitted inside structures housing

systems, equipment, or components important to nuclear safety.

Compliance Basis





FP-010,Housekeeping Controls Section 8.1.7

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8

EGR- NGGC-0005,Engineering Change ALL


Chapter 3 Reference: 3.3.7.1 [Bulk Flammable Gas Location Requirements]


Storage of flammable gas shall be located outdoors, or in separate detached buildings, so that a fire orexplosion 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 forhydrogen storage.


Complies via Engineering Evaluation HBRSEP complies with NFPA 50A asevaluated in RNP-M/BMRK-1015.


UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Appendix 9.5.1B-8

RNP-M/BMRK-1015,Code Compliance Evaluation for NFPA 50A, ALLStandard for Gaseous Hydrogen Systems at Consumer Sites - 1999Edition


Chapter 3 Reference: 3.3.7.2 [Bulk Flammable Gas Container Restrictions]

Chapter 3 Requirement: 3.3.7.2Outdoor high-pressure flammable gas storage containers shall be located so that the long axis is notpointed at buildings.


Complies

Compliance BasisNo Additional Clarification

Reference Document


Doc Details

Appendix 9.5.1B-8


Duke EnergySAF-SUBS-00023,Compressed Gases

MCP-NGGC-0402,Material Management (Storage, Issue, and

Maintenance)


Section 5.J.8

Section 9.1.6


Chapter 3 Reference: 3.3.7.3 [Bulk Flammable Gas Cylinder Limitations]


Flammable gas storage cylinders not required for normal operation shall be isolated from the system.




SAF-NGGC-2172,1ndustrial Safety Section 9.14

SAF-SUBS-00023,Compressed Gases Section 5.g.5


Chapter 3 Reference: 3.3.8 Bulk Storage of Flammable and Combustible Liquids.

Chapter 3 Requirement: 3.3.8 Bulk Storage of Flammable and Combustible Liquids.

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 Basis

HBRSEP complies with NFPA 30 asevaluated in RNP-M/BMRK-1002.


RNP-M/BMRK-1002,Code Compliance Evaluation NFPA 30 - Unit 2 ALLDiesel Fuel Oil Storage Tanks

FP-010,Housekeeping Controls Section 8.1.7, Attachment 10.1

UFSAR,HBR 2 Updated Fifal Safety Analysis Report (FSAR) Appendix 9.5.1B-9

FP-006,Handling of Flammable Liquids and Gases ALL


HBRSEP LAR Rev I Page A-20


Fire Protection Program & Design Elements

Chapter 3 Reference: 3.3.9 Transformers.

Chapter 3 Requirement: 3.3.9* Transformers.

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.


Complies

Compliance Basis



OST-642,Main Transformer Deluge System Flow Test (Refueling ALL

Interval)


Chapter 3 Reference: 3.3.10 Hot Pipes and Surfaces.

Chapter 3 Requirement: 3.3.10* Hot Pipes and Surfaces.

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.


Complies

Compliance Basis



FP-010,Housekeeping Controls Section 8.1

FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4.1.1



Chapter 3 Reference: 3.3.11 Electrical Equipment

Chapter 3 Requirement: 313.111 Electrical Equipment

Adequate clearance, free of combustible material, shall be maintained around energized electrical

equipment.


Complies

Compliance Basis





FIR-NGGC-0009,NFPA 805 TRANSIENT COMBUSTIBLES AND Section 9.4.12IGNITION SOURCE CONTROLS PROGRAM


Chapter 3 Reference: 3.3.12 Reactor Coolant Pumps.

Chapter 3 Requirement: 3.3.12* Reactor Coolant Pumps.For facilities with non-inerted containments, reactor coolant pumps with an external lubrication systemshall be provided with an oil collection system. The oil collection system shall be designed and installedsuch that leakage from the oil system is safely contained for off normal conditions such as accidentconditions or earthquakes. All of the following shall apply.(1) The oil collection system for each reactor coolant pump shall be capable of collecting lubricating oilfrom all potential pressurized and nonpressurized leakage sites in each reactor coolant pump oil

system.


Complies via Previous NRC Approval No oil collection system is provided for thereactor coolant pumps at HBRSEP.

Section (1): Complies via Previous NRCApproval By letter NLS-85-176 (3/7/1985), in

response to HBRSEP request for

exemption for requiring reactor coolantpump oil collection systems, the NRCstated the following:

"The containment contains three reactorcoolant pumps (A, B and C). These arelocated in bays (A, B and C). These baysalso contain safety related cabling for thereactor coolant loop instrumentation. BaysA and B share a common ceiling; Bay C isisolated from Bays A & B to some extent.The bays are covered by removable

concrete blocks. These blocks will causethe plume from an unmitigated fire to bediverted through the steam generator area.This area contains safety related steam

flow instrumentation sensing lines.

Oil spilled in Bay A, will be confined to BayA; however, oil spilled in Bays B and C canflow to adjacent areas. The foundation forthe reactor coolant pumps is at the237.000' level. The foundation for the



steam generators is at the 238.33' level.

The reactor coolant pump is located

between the pressurized portion of the oil

system and the steam generator supports,

and serves to shield the steam generator

supports in the event of an oil system

rupture.

The major combustible in each bay is the200 gallons of oil in each reactor coolant

pump.

The existing fire detection system in each

reactor coolant pump bay is a two-zone

detection system. One zone consists of asingle infrared flame detector; the other

zone consists of a 325°F fixed-

temperature heat detector. Activation of

one zone of detection sends an alarm to

the control room; activation of the second

zone of detection alarms in the controlroom and also opens the preaction water

deluge valve to the bay. Both detectors

are wall mounted.

The existing fire suppression system for

each bay, is a preaction sprinkler system.

Each bay has its own deluge valve, supply

header, and a ring

header that encircles the reactor coolant

pumps at elevation 239 feet 4 inches.

Each of the five risers off the ring header

have three 220°F closed head side wall

sprinklers at approximately 240 feet, 245

feet and 252 feet. elevations. These

systems are designed to meet theminimum residual pressure and flow

requirements of NFPA-Std-15.

The suppression system ring header

piping in Bay A is designed to withstand an

SSE, while Bays B and C are designed

such that a seismic event would not impact

safety related equipment due to

suppression system rupture The risersare restrained to withstand tt e nozzle

reaction forces. These forces are greater

than those anticipated from a seismic

event.

The existing containment spray system


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elementswould be used as an emergency back-up

to the bay suppression system if

necessary to cool the operating level and

containment annulus outside of the RCP

bays.

By letter dated June 7, 1983, the licensee

proposed to:

(1) Provide additional ceiling mounted heat

detectors to meet the spacing and location

requirements of NFPA-STD-72E,

"Standard on Automatic Fire Detectors.

(2) Replace existing closed head

sprinklers with special open water spray

nozzles and manual actuation from the

control room.

(3) Construct 6 inch dikes at the 231 feet

elevation in Bay B and Bay C.(4) Revise operating procedures for the

containment spray system to allow its

operation as a back up fire suppression

system with the sodium hydroxide valves

out.

By letter dated October 5, 1983, the

licensee committed to maintain an

automatically actuated closed-head

preaction system in lieu of a manually

actuated open-head system.

We have evaluated the fire protection for

the reactor coolant pump lube oil system

and conclude that the effects of a fire in an

RCP Bay will not prevent safe shutdown

capability. There are no components within

the RCP Bay that are required for safe

shutdown. The effects of any fire within an

RCP Bay will be prevented from affecting

the safe shutdown equipment outside the

RCP Bay by the suppression system

inside the RCP Bay and the ContainmentSpray System outside the Bay.

It is the staff's conclusion that: 1)installation of a re ctor coolant pump oil

collection system ih this facility would not

significantly enhance fire safety, and 2) the

existing fire protection system in the

Reactor Coolant Pump Bays with the

addition of the proposed modifications



Fire Protection Program & Design Elementsprovides an acceptable level of safety to

that achieved by compliance with the

requirements of Section 111.0 of Appendix

R to 10 CFR 50. Therefore, the licensee's

request for an exemption should be

granted."




requirements.



basis for approval.

Section (1): No oil collection system is

provided for the reactor coolant pumps at

RNP. See Section 3.3.12 above for

discussion of acceptability for lack of oil

collection system.


NLS-85-176,RCP Oil Collection Exemption ALL

Chapter 3 Requirement: (2) Leakage shall be collected and drained to a vented closed container that can hold the inventory ofthe reactor coolant pump lubricating oil system.


Section (2): Complies via Previous NRC Section (2): No oil collection system isApproval provided for the reactor coolant pumps at

HBRSEP. See Section 3.3.12 above fordiscussion of acceptability for lack of oil

collection system.

Chapter 3 Requirement: (3) A flame arrestor is required in the vent if the flash point characteristics of the oil present the hazardof a fire flashback.


Section (3): Complies via Previous NRC Section (3): No oil collection system isApproval provided for the reactor coolant pumps at

HBRSEr. See Section 3.3.12 above fordiscuss on of acceptability for lack of oilcollection system.


Duke EnergyC-hapter 3Requirement:


(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.


Section (4): Complies via Previous NRCApproval

Compliance Basis

Section (4): No oil collection system is

provided for the reactor coolant pumps at

HBRSEP. See Section 3.3.12 above for


collection system.

Chapter 3 Requirement: (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.


Section (5): Complies via Previous NRC Section (5): No oil collection system is

Approval provided for the reactor coolant pumps atHBRSEP. See Section 3.3.12 above for


collection system.


Chapter 3 Reference: 3.4 Industrial Fire Brigade.




requirements.


Chapter 3 Reference: 3.4.1 On-Site Fire-Fighting Capability.

Chapter 3 Requirement: 3.4.1 On-Site Fire-Fighting Capability.

All of the followi g requirements shall apply.

(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)



Section (a): Complies


Section (a): No Additional Clarification

Section (a) (1): Complies via Engineering

Evaluation

Section (a) (1): HBRSEP complies with

NFPA 600 as evaluated in NED-M/BMRK-

0002.



NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR ALLNFPA 600, STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000EDITION

Chapter 3 Requirement: (2) NFPA 1500, Standard on Fire Department Occupational Safety and Health Program

Compliance Statement Compliance BasisSection (a) (2): N/A Section (a) (2): NFPA 1500 is not

applicable to HBRSEP as the site utilizes afire brigade, not an organized fire

department. Fire Brigade requirements arereviewed using NFPA 600.

Chapter 3 Requirement: (3) NFPA 1582, Standard on Medical Requirements for Fire Fighters and Information for FireDepartment Physicians.


Section (a) (3): N/A Section (a) (3): NFPA 1582 is notapplicable to HBRSEP as the site utilizes afire brigade, not an organized fire

department. Fire Brigade requirements arereviewed using NFPA 600.

Chapter 3 Requirement: (b) * Industrial fire brigade members shall have no other assigned normal plant duties that wouldprevent immediate response to a fire or other emergency as required.


Section (b): Complies Section (b): No Additional Clarification


OMM-002,Fire Protection Manual Section 3.12 & 8.6

Chapter 3 Requirement: (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



Fire Protection Program & Design Elementsnuclear safety performance

Exception: Sufficient training and knowledge shall be permitted to be provided by an operations advisor

dedicated to industrial fire brigade support criteria.


Section (c): Complies

Compliance Basis

Section (c): No Additional Clarification



FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL

Chapter 3 Requirement: (d) * The industrial fire brigade shall be notified immediately upon verification of a fire.

Compliance Statement Compliance Basus

Section (d): Complies Section (d): No Additional Clarification


AOP-041,Response to Fire Event Section 2 Step 4

Chapter 3 Requirement: (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 fire-fighting operations. The physical

examination shall determine the ability of each member to use respiratory protection equipment.


Section (e): Complies Section (e): No Additional Clarification

Reference Documentc D±amls

FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.4.1



Chapter 3 Reference: 3.4.2 Pre-Fire Plans.

Chapter 3 Requirement: 3.4.2* Pre-Fire Plans.

Current and detailed pre-fire plans shall be available to the in qustrial fire brigade for all areas in which

a fire could jeopardize the ability to meet the performance criteria described in Section 1.5.




Duke EnergyReference Document

FIR-NGGC-0008,NFPA 805 Pre-Fire Plans


HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60


Doc Detanis

Sections 9.2 and 9.5

Section 8.8

ALL


Chapter 3 Reference: 3.4.2.1 [Pre-Fire Plan Contents]

Chapter 3 Requirement: 3.4.2.1*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.


Complies

Compliance Basis


Reference Document

FIR-NGGC-0008,NFPA 805 Pre-Fire Plans

Doc Details

Section 9.2


Chapter 3 Reference: 3.4.2.2 [Pre-Fire Plan Updates]

Chapter 3 Requirement: 3.4.2.2 Pre-fire plans shall be reviewed and updated as necessary.




FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 4.0

AP-043,RNP Procedure Biennial Review Process Section 8.1

EGR-NGGC-0005,Engineering Change Section 4.5

RNP/94-1890,PROPOSED CHANGE TO QUALITY ASSURANCE ALL

PROGRAM



Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDukeater 3Referencen 3.4.2.3 [Pre-Fire Plan Locations]


Pre-fire plans shall be available in the control room and made available to the plant industrial fire

brigade.




FIR-NGGC-0008,NFPA 805 Pre-Fire Plans Section 9.5


Chapter 3 Reference: 3.4.2.4 [Pre-Fire Plan Coordination Needs]


Pre-fire plans shall address coordination with other plant groups during fire emergencies.


Complies with Clarification HBRSEP has procedure FP-001, "Fire

Emergency" which is not specifically a fire

pre-plan, however FP-001 provides

specific instructions for actions required

from key groups at HBRSEP supporting

the fire brigade/fire emergency actions.

There are detailed response coordination

actions specified for Control Room, RC,

and the Security group. Any other

coordination actions would be initiated by

the Control Room personnel as needed for

any plant emergency.


FP-001 ,Fire Emergency ALL


Chapter 3 Reference: 3.4.3 Training and Drills.

Chapter 3 Requirement: 3.4.3 Training and Drills.

Industrial fire brigade members and other plant personnel who would respond to a fire in conjunction



Fire Protection Program & Design Elementswith 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.



Section (a) (1): Complies via EngineeringEvaluation

Compliance Basis


Section (a) (1): HBRSEP complies with

NFPA 600 as evaluated in the applicable

portions of NED-M/BMRK-0002.

Reference Document DDetails


NED-M/BMRK-0002,CODE COMPLIANCE EVALUATION FOR ALLNFPA 600, STANDARD ON INDUSTRIAL FIRE BRIGADES, 2000EDITION

Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(2) Industrial firebrigade members shall be given quarterly training and practice in fire fighting, including radioactivityand health physics considerations, to ensure that each member is thoroughly familiar with the steps to

be taken in the event of a fire.


Section (a) (2): Complies Section (a) (2): No Additional Clarification



GNR01 N,Plant Access Annual Requalification, CBT ALL

Chapter 3 Requirement: (a) Plant Industrial Fire Brigade Training. All of the following requirements shall apply.(3) A writtenprogram shall detail the industrial fire brigade training program.

Compliance Statement Compliance BasisSection (a) (3): Complies Section (a) (3): No Additional Clarification


FIR-Nt 3GC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM ALL

Chapter 3 Requirements (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.


Duke Energy


Section (a) (4): Complies


Compliance Basis

Section (a) (4): No Additional Clarification



TAP-404,Training Documentation and Records ALL

Chapter 3 Requirement: (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.


Section (b): Complies with Clarification

Compliance Basis

Section (b): Guidance for non-industrial

fire brigade members is found in FP-001.

The procedure defines the actions needed

to be taken by personnel discovering a fire,

security personnel actions, and duty health

physics contact actions.

Reference Document DoDetamls

FP-001 ,Fire Emergency ALL

Chapter 3 Requirement: (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.


Section (c) (1): Complies Section (c) (1): No Additional Clarification


FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM Section 9.10.3.3.a

Chapter 3 Requirement: (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.





FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM


Doc DetailsSection 9.10

Chapter 3 Requirement: (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 operationand to contain significant fire hazards.


Section (c) (3): Complies

Compliance Basis

Section (c) (3): No Additional Clarification



Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(4) Drill records shall be maintained detailing thedrill scenario, industrial fire brigade member response, and ability of the industrial fire brigade to

perform as a team.


Section (c) (4): Complies

Compliance Basis

Section (c) (4): No Additional Clarification




Chapter 3 Requirement: (c) * Drills. All of the following requirements shall apply.(5) A critique shall be held and documented

after each drill.







Chapter 3 References 3.4.4 Fire-Fighting Equipment.

Chapter 3 Requirement: 3.4.4 Fire-Fighting Equipment.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.


Complies with Clarification Per FAQ 06-0020, the following guidance

applies as to which NFPA standards

referenced in Chapter 3 are applicable:

"Where used in NFPA 805, Chapter 3, the

term, "applicable NFPA Standards" is

considered to be equivalent to those NFPA

standards identified in the current license

basis (CLB) for procedures and systems in

the Fire Protection Program that are

transitioning to NFPA 805."

Firefighting equipment is provided. A

monthly inspection/inventory "of Fire

Protection equipment and supplies located

in the Fire Equipment Staging areas to

meet the demands of the site Fire

Brigade..." is conducted per OST-639.

Personnel dosimeters are issued in

accordance with the plant radiationprotection program and DOS-NGGC-

0002. HP personnel, who provide fire

brigade support, provide radiation

monitoring equipment in accordance with

FP-001.

HBRSEP makes use of Duke Energy fleet

procedure AD-EG-ALL-1531, Selection,

Care and Maintenance of Fire Fighting

Ensembles which follows the guidance

found in NFPA standards associated with

firefighting Personal Protective Equipment.

Standards and requirements associated

with installed fire protection equipment

such as hoses, nozzels, fire extinguishers

and other equipment are maintained in

accordance with NFPA standards as

described elsewhere in Chapter 3 and

evaluated for compliance under the

various NFPA Code Compliance

Calculations listed for those sections of

Chapter 3.

Reference Document Doe Details


Duke EnergyFAQ 06-0020,Identification of "applicable NFPA standards"

OST-639,Fire Equipment Inventory (Monthly)

FP-001,Fire Emergency


DOS-NGGC-0002,Dosimetry Issuance

AD-EG-ALL-1531 ,Selection, Care, and Maintenance of Fire FightingEnsembles


ALL

ALL

Section 3.7

Section 3.27

ALL

ALL


Chapter 3 Reference: 3.4.5 Off-Site Fire Department Interface.




requirements.


Chapter 3 Reference: 3.4.5.1 Mutual Aid Agreement.

Chapter 3 Requirement: 3.4.5.1 Mutual Aid Agreement.

Off-site fire authorities shall be offered a plan for their interface during fires and related emergencies on

site.




PLP-007, Robinson Emergency Plan Attachment 6.2


Chapter 3 Reference: 3.4.5.2 Site-Specific Training.

Chapter 3 Requirement: 3.4.5.2* Site-Specific Training.

Fire fighters from the off-site fire authorities who are expected to respond to a fire at the plant shall be



Fire Protection Program & Design Elementsoffered site-specific training and shall be invited to participate in a drill at least annually.


Complies

Compliance Basis


Reference Document

FIR-NGGC-0007,NFPA 805 FIRE BRIGADE TRAINING PROGRAM

Doco Dta9is

Section 9.7.1


Chapter 3 Reference: 3.4.5.3 Security and Radiation Protection.

Chapter 3 Requirement: 3.4.5.3* Security and Radiation Protection.

Plant security and radiation protection plans shall address off-site fire authority response.





PLP-007,Robinson Emergency Plan Table 5.3.2-1 Notes


Chapter 3 Reference: 3.4.6 Communications.

Chapter 3 Requirement: 3.4.6* Communications.An effective emergency communications capability shall be provided for the industrial fire brigade.




PLP-007,Robinson Emergency Plan Attachment 6.1

OST-639,Fire Equip ent Inventory (Monthly) ALL


HBRSEP LAIR Rev 1 Page A-36


Chapter 3 Reference: 3.5 Water Supply




requirements.


Chapter 3 Reference: 3.5.1 [Water Supply Flow Code Requirements]

Chapter 3 Requirement: 3.5.1

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 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 Basis

HBRSEP complies with Section (b) ofNFPA 805 Ch. 3 Section 3.5.1 as detailedin RNP-M/BMRK-1011, RNP-M/MECH-1727, and RNP-M/MECH-1728.

Reference Document

RNP-M/BMRK-1 01 1,Code Compliance Evaluation for NFPA 15,Water Spray Fixed Systems

RNP-M/MECH-1 727,Hydraulic Analysis of the Hydrogen Seal OilWater Spray System

RNP-M/MECH-1 728,Hydraulic Analysis of the Auxiliary & Start-UpTransformer Water Spray System

NLU -78-71, License Amendment 31

Code Section 3012

Section 5

Section 5

Section 4.3.1.1



Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDukeapter 3Reference: 3.5.2 [Water Supply Tank Code Requirements]

Chapter 3 Requirement: 3.5.2*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 BasisFire water is obtained directly from Lake

Robinson via a single intake structure.

Two physically separated automatic fire

pumps are provided with separate suction

lines.

As such, HBRSEP complies with

Exception No. 1


NLU-78-71 ,License Amendment 31 Section 4.3.1.1


Chapter 3 Reference: 3.5.3 [Water Supply Pump Code Requirements]

Chapter 3 Requirement: 3.5.3*

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.


Complies via Engineering Evaluation HBRSEP complies with NFPA 20 as

evaluated in RNP-M/BMRK-1012, RNP-

M/MECH-1725, an• RNP-M/MECH-1610.

Hydraulic analysis 4emonstrates the ability

of one pump to provide required flow rate

to the largest system.



Duke Energy

RNP-M/MECH-1725,Evaluation of NFPA 13 Code ComplianceVariances

RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 -Centrifugal Fire Pumps

FP-012,Fire Protection Systems Minimum Equipment andCompensatory Actions

RNP-M/MECH-1610,Hydraulic Analysis - Main Transformers


ALL

Conclusions (Page 9) & Attachment 6-Code

Section 33-Page 4 of 62

Section 8.2

ALL


Chapter 3 Reference: 3.5.4 [Water Supply Pump Diversity and Redundancy]


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.


Complies

Compliance Basis

No Additional Clarification.


RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Body of CalculationCentrifugal Fire Pumps

FP-012,Fire Protection Systems Minimum Equipment and Section 8.2Compensatory Actions


Chapter 3 Reference: 3.5.5 [Water Supply Pump Separation Requirements]

Chapter 3 Requirement: 3.5.5Each pump and its driver and controls shall be separated from the remaining fire pumps and from therest of the plant by rated fire barriers.


Complies via Previous NRC Approval In a submittal dated 6/23/77, HBRSEPprovided the following information:

"The Unit 2 fire protection water supplysystem meets the intent of the

requirements (BTP ABCSB 9.5-1) either

by itself or by virtue of cross-connection to


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy the Unit 1 system, except that the Unit 2 Fire Protection Program & Design Elements

supply pumps are not separated from

each other or remaining pumps by three-

hour rated fire walls. However, the various

fire pumps are out-of-doors and separated

by distance as well as intervening

equipment."

In the SER dated 2/21/1980, the NRC

stated:

"The staff does not agree with the

licensee's contention that the arrangement

of the propane storage tank and other

equipment is satisfactory in relation to

safety-related equipment on the intake

structure for the following reasons....

Therefore, we will require the licensee to:

-Replace the propane engine with a diesel

engine, or

-Replace the propane engine-driven fire

pump and associated equipment to alocation substantially remote from any

safety-related equipment. "

The original propane-fueled engine driver

on one of the two 100% capacity fire

pumps was changed to a diesel-fueled

driver per Modification M-445P to address

concerns raised by the NRC over the

propane storage location and

arrangement.




requirements.



basis for approval.


NLU-80-106,RFI and Requirements to Resolvp Issues Concerning Section 3.2.3

Fire Protection

NG-77-704,Fire Protection Program Review Question 15

M-445P,Fire Pump Engine Replacement & Propane Tank Relocation ALL



Table B-1-NFPA 805 Ch.3 Transition Details

Chapter 3 Reference: 3.5.6 [Water Supply Pump Start/Stop Requirements]


Fire pumps shall be provided with automatic start and manual stop only.

Compliance Statement Complsance Basis

Complies via Engineering Evaluation Fire pumps are provided with automatic

start and manual stops, as detailed in the

applicable portions of the NFPA 20 code

compliance evaluation RNP-M/BMRK-

1012.

Reference Document nD taifi

RNP-M/BMRK-1012,Code Compliance Evaluation NFPA 20 - Code Sections 515 (Attachment 6), Code Section

Centrifugal Fire Pumps 9-5 (Attachment 7), and Code Section 9-5

(Attachment 8)


Chapter 3 Reference: 3.5.7 [Water Supply Pump Connection Requirements]


Individual fire pump connections to the yard fire main loop shall be provided and separated with

sectionalizing valves between connections.


Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:

"4.3.1.3 Fire Water Piping System

The two fire pumps have a common

discharge through a twelve inch

underground main into a ten-inch

underground fire water loop which

encircles the plant.

All yard fire hydrants, automatic water

s+ppression systems and interior fire hose

lines are supplied by the fire loop.

Sectionalizing valves of the post indicator

type are provided on the fire loop to permit

partial pipeline isolation without

interruption of service to the entire system.


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy The licensee will install isolation valves at Fire Protection Program & Design Elements

the connection of the ten inch fire line from

the Unit I fire loop to the Unit 2 fire loop,

and provide separate headers for

automatic sprinkler systems to be installed

in the reactor auxiliary building. The

licensee will also provide barriers around

all hydrants and post indicator valves to

protect against vehicular damage.

Electrical supervision, to monitor the

position of fire water system controlvalves, is not provided. A means of sealing

these valves open will be provided, and

this in combination with administrative

controls and periodic inspections will be

used to assure that valves are maintained

open.

We find that, subject to the implementation

of the above described modifications, the

fire water piping systems satisfy the

objectives identified in Section 2.1 of this

report and are, therefore, acceptable."

As seen in drawings HBR2-08255-Sheets

1 and 2, isolation valves were installed at

the connection of the Unit 1 fire loop lines

to the Unit 2 fire loop and separateheaders for automatic sprinkler systems

were installed in the reactor auxiliary

building.




requirements.



basis for approval.


NLU-78-71 ,License Amendment 31 4.3.1.3

HBR2-08255 Sh 2,Fire Protection System Flow Diagram ALL

HBR2-08255 Sh 1 ,Fire Protection System Intake Structure Flow ALL

Diagram


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalI 11i1w I--l~rp1 I-|rw wrrTinn Irn•prn m & I np.inn R-I ments


Chapter 3 Reference: 3.5.8 [Water Supply Pressure Maintenance Limitations]


A method of automatic pressure maintenance of the fire protection water system shall be provided

independent of the fire pumps.




UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.2.7

RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Code Section 661

Standpipes and Hose Stations

HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow ALL

Diagram

APP-044,Fire Alarm Console (FAC) C55, C58


Chapter 3 Reference: 3.5.9 [Water Supply Pump Operation Notification]


Means shall be provided to immediately notify the control room, or other suitable constantly attended

location, of operation of fire pumps.


Complies

Compliance Basis



UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.4.3.4.1.c

APP-044,Fire Alarm Console (FAC) C55, C58

Table B-1 NFPA 805 Ch.3 TLansition Details

Chapter 3 Reference: 3.5.10 [Water Supply Yard Main Code Requirements]


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFirp Prnt.etionn,2P-mlam_&JDesiq•,Fpemrs01ukeErie=


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 BasisHBRSEP complies with NFPA 24 as

evaluated in RNP-M/BMRK-1013.

Reference Document DoDetamis

RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24 - ALL

Standard for Outside Protection

RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code Section 4.1

Compliance Variances


Chapter 3 Reference: 3.5.11 [Water Supply Yard Main Maintenance Issues]


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.

Comp•liance Statement Compliance Basis

Complies via Previous NRC Approval In the SER dated 2/28/78, the NRC stated:

"4.3.1.3 Fire Water Piping System

The two fire pumps have a common

discharge through a twelve inch

underground main into a ten-inch

underground fire water loop which

encircles the plant.

All yard fire hydrants, automatic water

suppression systems and interior hose

lines are supplied by the fire loop.

Sectionalizing valves of the post indicator

type are provided on the fire loop to permit

partial pipeline isolation without

interruption of service to the entire system.

The licensee will install isolation valves at


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design Elementsthe connection of the ten inch fire line from

the Unit 1 fire loop to the Unit 2 fire loop,

and provide separate headers for

automatic sprinkler systems to be installed

in the reactor auxiliary building. The

licensee will also provide barriers around

all fire hydrants and post indicator valves

to protect against vehicular damage.

Electrical supervision, to monitor theposition of fire water system control

valves, is not provided. A means of sealing

these valves open will be provided, and

this in combination with administrative

controls and periodic inspections will be

used to assure that valves are maintained

open.









requirements.



basis for approval.


NLU-78-71 License Amendment 31 Section 4.3.1.3

HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow ALL

Diagram


HBR2-08255 Sh. 6,Fire Protection System Flow Diagram ALL


Chapter 3 Reference: 3.5.12 [Water Supply Compatible Thread Connections]


Threads compatible with those used by local fire departments shall be provided on all hydrants, hose



couplings, and standpipe risers.

Exception: Fire departments shall be permitted to be provided with adapters that allow interconnection

between plant equipment and the fire department equipment if adequate training and procedures are

provided.


Complies

Compliance Basis


Reference Document


Doc Details

Section 9.5.1.4.2.5


Chapter 3 Reference: 3.5.13 [Water Supply Header Options]



N/A

3.5.13

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 of 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 analyzed 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, gate valve or other approved shutoff valve.

Compliance Basis

No headers at HBRSEP are fed from each

end.

Reference Document



Chapter 3 Reference: 3.5.14 [Water Supply Control Valve Supervision]



3.5.14*

All fire protection water supply and ire 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.

Compliance Basis


Duke EnergyComplies


HBRSEP complies with Section 3.5.14 bya combination of (b) & (c).


OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL

(Monthly) and Valve Cycling (Annual)

OST-652,Unit 2 Containment Fire Water System Valves ALL

NLU-78-71,License Amendment 31 Section 4.3.1.3

Chapter 3 Requirement: (b) Locking valves in their normal position. Keys shall be made available only to authorized personnel.


Complies

Compliance Basis

HBRSEP complies with Section 3.5.14 by

a combination of (b) & (c).


OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL(Monthly) and Valve Cycling (Annual)



Chapter 3 Requirement: (c) Sealing valves in their normal positions. This option shall be utilized only where valves are locatedwithin fenced areas or under the direct control of the owner/operator.


Complies HBRSEP complies with Section 3.5.14 bya combination of (b) & (c).


OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL(Monthly) and Valve Cycling (Annual)


NLU-78-71 License Amendment 31 Section 4.3.1.3


Chapter 3 Reference: 3.5.15 [Water Supply Hydrant Code Requirements]

Chapter 3 Requirement: 3.5.15Hydrants 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 Appurtenances, 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.


Complies via Previous NRC Approval COMPLIES VIA PREVIOUS NRC

APPROVAL: In the SER dated 2/28/78,

Complies via Engineering Evaluation the NRC stated:

"Five yard fire hydrants are provided at

approximately 250 foot intervals around

the exterior of the plant except at the north

end of the plant, where the distance

between hydrants is somewhat larger. A

hose house located near each fire hydrant

contains 2-1/2 inch diameter fire hose and

other manual firefighting tools. A sixth

hose house is centrally located between

the reactor and Turbine Building, a

seventh is located at the intake structure,

and one hose house is located outside the

Unit 2 fence east of the auxiliary building.

Standard fire hose threads are used on allfire protection equipment, and the threads

are compatible with those used by the

local public fire departments.









requirements.



basis for approval.


EVALUATION: HBRSEP complies with the

applicable portions of NFPA 24 as detailed

in RNP-M/BMRK-1013.



NLU-78-71 ,License Amendment 31

RNP-M/BMRK-1013,Code Compliance Evaluation NFPA 24-

Standard for Outside Protection

RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDocDetalil

Section 4.3.1.3

ALL

ALL


Chapter 3 Reference: 3.5.16 [Water Supply Dedicated Limits]


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.


License Amendment Required

Compliance Basis

NRC approval is being requested in

Attachment L for the use of the fire

protection water supply system for

purposes other than fire protection.

Reference Document

HBR2-08255 Sh 1,Fire Protection System Intake Structure Flow

Diagram

HBR2-08255 Sh 2,Fire Protection System Flow Diagram

HBR2-08255 Sh 3,Fire Protection System Containment Flow

Diagram

HBR2-08255 Sh. 6,Fire Protection System Flow Diagram


AOP-014,Component Cooling Water System Malfunction

AOP-022,Loss of Service Water

EDMG-001,Extreme Damage Event Early Actions and Response

Determination Criteria

EDMG-002,Refueling Water Storage Tank (RWST)

EDMG-003,Condensate Storage Tank (CST)

Doc Details

ALL

ALL

ALL

ALL

Section 8.15

ALL

ALL

ALL

ALL

ALL



EDMG-005 ,Containment Vessel (CV)

EDMG-01 1 ,Spent Fuel Pool Casualty

EDMG-012,Core Cooling Using Alternate Water Source

EDMG-013,Airborne Release Scrubbing

ALL

ALL

ALL

ALL


Chapter 3 Reference: 3.6 Standpipe and Hose Stations.



N/A N/A


Chapter 3 Reference: 3.6.1 [Standpipe and Hose Station Code Requirements]


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.



Complies via Previous NRC Approval

Compliance Basis


EVALUATION: HBRSEP complies with

NFPA 14 as detailed in RNP-M/BMRK-

1010 and RNP-M/MECH-1709.

COMPLIES VIA PREVIOUS NRC

APPROVAL: HBRSEP has Class II

standpipes in lieu of Class I1l. In the

License Amendment dated 2/28/78, the

NRC stated:

"4.3.1.4 Interior Fire Hose Stations

A total of 24 interior hose stations, each

presently equipped with 50 feet of 1-1/2

inch diameter hose, have been provided

throughout all portions of the plant except

containment. There are presently several

safety-related areas containing

combustible materials that are beyond the

reach of the existing hose lines. The


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy licensee will provide additional hose Fire Protection Program & Design Elements

stations or additional lengths of hose at

existing stations so that sufficient hose

reach is provided to protect all the areas of

the auxiliary building.

Hose racks originally used for unlined

hose are being used to store rubber lined

hose. The licensee has committed to

replace these with suitable hose reels or

hose racks designed and sized for lined

hose.

The nozzles on the interior hose lines are

1-1/2" spray nozzles. In areas with

electrical hazards, "electrically safe" hose

nozzles have been provided on the hose

station nearest these areas.



interior fire hose stations satisfy the






requirements.



basis for approval.


RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, ALL

Standpipes and Hose Stations

RNP-M/MECH-1709,Evaluation of NFPA 14 and 24 Code ALL


NLU-78-71 ,License Amendment 31 4.3.1.4


Chapter 3 Reference: 3.6.2 [Standpipe and Hose Station Capability Limitations]


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.


Duke Energy




Compliance BasisHBRSEP complies with NFPA 805requirement 3.6.2 as detailed in theapplicable portions RNP-M/BMRK-1010.


RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, ALLStandpipes and Hose Stations


Chapter 3 Reference: 3.6.3 [Standpipe and Hose Station Nozzle Restrictions]

Chapter 3 Requirement: 3.6.3The proper type of hose nozzle to be supplied to each power block area shall be based on the area firehazards. The usual combination spray/straight stream nozzle shall not be used in areas where thestraight stream can cause unacceptable damage or present an electrical hazard to fire-fightingpersonnel. Listed electrically safe fixed fog nozzles shall be provided at locations where high-voltageshock hazards exist. All hose nozzles shall have shutoff capability and be able to control water flowfrom full open to full closed.

Compliance Statement Compliance BasisComplies via Engineering Evaluation HBRSEP complies with NFPA 805

requirement 3.6.3 as detailed in theapplicable portions of RNP-M/BMRK-1010.


RNP-M/BMRK-1010,Code Compliance Evaluation for NFPA 14, Code Section 451 & 452Standpipes and Hose Stations



Chapter 3 Reference: 3.6.4 [Standpipe and Hose Station Earthquake Provisions]

Chapter 3 ReqLirement: 3.6.4 1

Provisions shall be made to supply water at least to standpipes and hose stations for manual firesuppression in all areas containing systems and components needed to perform the nuclear safetyfunctions in the event of a safe shutdown earthquake (SSE).





Seismic standpipes are not an original

commitment for HBRSEP.

The Federal Register notice that

promulgated adoption of NFPA 805 makes

the following statement:

"A commenter noted that Appendix A to

BTP APCSB 9.5-1 did not require

seismically qualified standpipes and hose

stations for operating plants and plants

with construction permits issued prior to

July 1, 1976. NRC agrees that Appendix A

to BTP APCSB 9.5-1 made separate

provisions for operating plants and plants

with construction permits issued prior to

July 1, 1976, and did not require

seismically qualified standpipes and hose

stations for those plants. Therefore, the

requirement in Section 3.6.4 of NFPA 805

is not applicable to licensees with

nonseismic standpipes and hose stations

previously approved in accordance with

Appendix A to BTP APCSB 9.5-1."



basis for approval.

Reference Document

69 FR 33356,Final Rule - NFPA 805

DocDetails

Page 33544


Chapter 3 Reference: 3.6.5 [Standpipe and Hose Station Seismic Connection Limitations]


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.

COm~lanceSttment

N/A

Compliance Basis

HBRSEP is not committed to havingseismic standpipes. See the ComplianceSection for 3.6.4




Chapter 3 Reference: 3.7 Fire Extinguishers.

Chapter 3 Requirement: 3.7 Fire Extinguishers.

Where provided, fire extinguishers of the appropriate number, size, and type shall be provided inaccordance with NFPA 10, Standard for Portable Fire Extinguishers. Extinguishers shall be permitted tobe positioned outside of fire areas due to radiological conditions.



Compliance Basis

HBRSEP complies with NFPA 10 as

evaluated in RNP-M/BMRK-1001.


RNP-M/BMRK-1001,Code Compliance Evaluation NFPA 10 Portable ALLFire Extinguishers


Chapter 3 Reference: 3.8 Fire Alarm and Detection Systems.



N/A N/A - General statement; No technicalrequirements.


Chapter 3 Reference: 3.8.1 Fire Alarm.



3.8.1 Fire Alarm.Alarm initiating devices shall be installed in accordance with NFPA 72, National Fire Alarm Code®.Alarm annunciation shall allow tle proprietary alarm system to transmit fire-related alarms, supervisorysignals, and trouble signals to th control room or other constantly attended location from whichrequired notifications and response can be initiated. Personnel assigned to the proprietary alarm stationshall be permitted to have other duties. The following fire-related signals shall be transmitted:(1) Actuation of any fire detection device

Compliance Basis


Duke Energy(Main Header): Complies via Engineering

Evaluation



(Main Header): HBRSEP complies with

NFPA 72 as evaluated in the applicable

portions of RNP-M/BMRK-1014, 1005,

1006.


Reference Document

RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National

Fire Alarm Code

RNP-M/BMRK-1005,Code Compliance Evaluation NFPA 72D

RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E

RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance

Variances

APP-044,Fire Alarm Console (FAC)


Doc Details

ALL

ALL

ALL

ALL

ALL

Appendix 9.5.1B-14

Chapter 3 Requirement: (2) Actuation of any fixed fire suppression system




APP-044,Fire Alarm Console (FAC) ALL

Chapter 3 Requirement: (3) Actuation of any manual fire alarm station





Chapter 3 Requirement: (4) Starting of any fire pump


Section (4): Complies Se tion (4): No Additional Clarification




Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsChapter 3 Requirement: (5) Actuation of any fire protection supervisory device



Compliance Basis




Chapter 3 Requirement: (6) Indication of alarm system trouble condition






Chapter 3 Reference: 3.8.1.1 [Fire Alarm Communication Requirements]


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.




FP-001,Fire Emergency Section 5.2

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Sections 9.5.2


Chapter 3 Reference: 3.8.1.2 [Fire Al rm Prompt Notification Limits]


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.





Compliance Basis


Reference Document DoDetaols

AOP-041, Response to Fire Event Section 2

Chapter 3 Requirement: (2) Members of the industrial fire brigade and other groups supporting fire emergency response




AOP-041,Response to Fire Event Section 2

Chapter 3 Requirement: (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




PLP-007,Robinson Emergency Plan Attachment 6.1


Chapter 3 Reference: 3.8.2 Detection.

Chapter 3 Requirement: 3.8.2 Detection.

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 BasisHBRSEP complies with NFPA 72 as

evaluated in the applicable portions ofRNP-M/BMRK-1014, 1005, 1006.

Reference Document

RNP-M/BMRK-1014,Code Compliance Evaluation NFPA 72 National

Fire Alarm Code

Doc Details

ALL


Duke Energy,RNP-M/BM RK-1005,Code Compliance Evaluation NFPA 72D

RNP-M/BMRK-1006,Code Compliance Evaluation for NFPA 72E

RNP-M/MECH-1697,Evaluation of NFPA 72E Code Compliance

Variances


ALL

ALL

ALL


Chapter 3 Reference; 3.9 Automatic and Manual Water-Based Fire Suppression Systems.



N/A N/A - General statement; No technicalrequirements.


Chapter 3 Reference: 3.9.1 [Fire Suppression System Code Requirements]


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


Section (1): Complies via Engineering Section (1): HBRSEP complies with NFPA

Evaluation 13 as evaluated in the applicable portions

of RNP-M/BMRK-1009.


RNP-M/BMRK- 1009,Code Compliance Evaluation for NFPA 13, ALL

Sprinkler Systems

RNP-M/MECH-1725,Evaluation of NFPA 13 Code Compliance ALL

Variances

Chapter 3 Requirement: (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Protection


Section (2): Complies via Engineering

Evaluation

Compliance BasisSection (2): HBRSEP complies with NFPA

15 as evaluated in the applicable portions


Duke Energy of RNP-M/BMRK-1011.Attachment A - NEI 04-02 Table B-1 Transition of Fundamental

Fire Protection Program & Design Elements

Reference Document

RNP-M/BMRK-1011,Code Compliance Evaluation for NFPA 15,

Water Spray Fixed Systems

RNP-M/MECH-1726,Evaluation of NFPA 15 Code Compliance

Variances

RNP-M/MECH-1 727,Hydraulic Analysis of the Hydrogen Seal Oil

Water Spray System

RNP-M/MECH-1728,Hydraulic Analysis of the Auxiliary & Start-Up

Transformer Water Spray System

Doc Details

ALL

ALL

ALL

ALL

Chapter 3 Requirement: (3) NFPA 750, Standard on Water Mist Fire Protection Systems


Section (3): N/A Section (3): N/A - No Water Mist FireProtection Systems are installed at

HBRSEP.

Chapter 3 Requirement: (4) NFPA 16, Standard for the Installation of Foam-Water Sprinkler and Foam-Water Spray Systems


Section (4): N/A

Compliance Basis

Section (4): N/A - No Foam-Water

Sprinkler or Foam-Water Spray Systems

are installed at HBRSEP.


Chapter 3 Reference: 3.9.2 [Fire Suppression System Flow Alarm]


Each system shall be equipped with a water flow alarm.


Complies via Engineering Evaluation COMPLIES VIA ENGINEERINGEVALUATION: Several of the automatic

water-based fire suppressio' systems do

not have water flow alarms. ;These

systems have less than 20 sprinklers and

are not required to have water flow alarms.

The systems discussed are in the CCW

Pump Room, Turbine Generator



Fire Protection Program & Design ElementsMonitoring Room (TGMR), and the Rad

Waste Building. Per RNP-M/BMRK- 1009,

"Code Compliance Evaluation NFPA 13-

Standard for Installation for Sprinkler

Systems", and RNP-M/BMRK-1011,

"Code Compliance Evaluation NFPA 15 -

Water Spray Fixed Systems", all

requirements from NFPA 13 and NFPA 15

for water flow alarms are met.


RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Code Sections 3-16.2, 3-17.2

Sprinkler Systems




Chapter 3 Reference: 3.9.3 [Fire Suppression System Alarm Locations]


All alarms from fire suppression systems shall annunciate in the control room or other suitable

constantly attended location.


Complies via Engineering Evaluation HBRSEP complies with NFPA 805

requirement 3.9.3 as detailed in the

applicable portions of RNP-M/BMRK-

1009.


RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Code Sections 5-3.5.2, 3-17.3.3, 3-17.6.2

Sprinkler Systems





Chapter 3 Reference: 3.9.4 [Fire Suppression System Diesel Pump Sprinkler Protection]


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Enerqy 3Fire Protection Program & Design ElementsCDueterr3fRepuirement:o t.9ks

Diesel-driven fire pumps shall be protected by automatic sprinklers.



Compliance BasisCOMPLIES VIA ENGINEERING

EVALUATION: The diesel-driven fire

pump is installed outdoors at the Intake

Structure on Lake Robinson. Per RNP-

M/BMRK- 1012, "Code Compliance

Evaluation NFPA 20- Centrifugal Fire

Pumps", all requirements from NFPA 20-

1978, for outdoor diesel-driven fire pumps,

are met.


HBR2-11937,Fire Pre-Plan Drawings, Sh 1-60 Sheet 45


Chapter 3 Reference: 3.9.5 [Fire Suppression System Shutoff Controls]


Each system shall be equipped with an OS&Y gate valve or other approved shutoff valve.


Complies via Engineering Evaluation HBRSEP complies with NFPA 805

requirement 3.9.5 as detailed in the

applicable portions of RNP-M/BMRK-1009

and RNP-M/BMRK-1011.


RNP-M/BMRK-1009,Code Compliance Evaluation for NFPA 13, Sections 3-13.1.1, 3-14.1.1

Sprinkler Systems

RNP-M/BMRK-101 1 ,Code Compliance Evaluation for NFPA 15, Section 2080



Chapter 3 Reference: 3.9.6 [Fire Suppression System Valve Supervision]


All valves controlling water-based fire suppression systems required to meet the performance or



Fire Protection Program & Design Elementsdeterministic requirements of Chapter 4 shall be supervised as described in 3.5.14.


Complies

Compliance Basis



OST-602,Unit No. 2 Fire Water System Flowpath Verification ALL

(Monthly) and Valve Cycling (Annual)




Chapter 3 Reference: 3.10 Gaseous Fire Suppression Systems.




requirements.


Chapter 3 Reference: 3.10.1 [Gaseous Suppression System Code Requirements]


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 applicable NFPA codes:

(1) NFPA 12, Standard on Carbon Dioxide Extinguishing Systems



Evaluation

Compliance BasisSection (1): HBRSEP complies with NFPA

12 as evaluated in the applicable portions

of RNP-M/BMRK-1007.

Reference Document I

RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, CarbonDioxide Extinguishing Systems

RNP-M/MECH-1708,Evaluation of NFPA 12 Code ComplianceVariances

Doc Details

ALL

ALL


Duke Energy

HBR2-11992 SH00001,EDG HIGH PRESSURE C02 FIRE

EXTINGUISHING SYSTEM

HBR2-11992 Sh 02,EDG High Pressure C02 Fire Extinguishing

System

HBR2-11992 Sh 03,EDG High Pressure C02 Fire Extinguishing

System


ALL

ALL

ALL

Chapter 3 Requirement: (2) NFPA 12A, Standard on Halon 1301 Fire Extinguishing Systems


Section (2): Complies via Engineering Section (2): HBRSEP complies with NFPA

Evaluation 12A as evaluated in the applicable

portions of RNP-M/BMRK-1008.

Reference Document DocDetaols

RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, ALL

Halon 1301 Systems

Chapter 3 Requirement: (3) NFPA 2001, Standard on Clean Agent Fire Extinguishing Systems


Section (3): N/A Section (3): Clean Agent Fire

Extinguishing Systems are not utilized at

HBRSEP.


Chapter 3 Reference: 3.10.2 [Gaseous Suppression System Alarm Location]


Operation of gaseous fire suppression systems shall annunciate and alarm in the control room or other

constantly attended location identified.




RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon

Dioxide Extinguishing Systems

RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A,

Halon 1301 Systems

Section 1452 & 1-8.5

Section 1-8.4


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke Enery

APP-044,-ire Alarm Console (FAC) ALL


Chapter 3 Reference: 3.10.3 [Gaseous Suppression System Ventilation Limitations]


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.


Complies via Engineering Evaluation HBRSEP complies with NFPA 805requirement 3.10.3 as detailed in the

applicable portions of RNP-M/BMRK-1007

and RNP-M/BMRK-1008.

Reference Document Doc Detanis

RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon ALL


RNP-M/BMRK-1008,Code Compliance Evaluation for NFPA 12A, ALL

Halon 1301 Systems

RNP-M/MECH-1708,Evaluation of NFPA 12 Code Compliance ALL

Variances


Chapter 3 Reference: 3.10.4 [Gaseous Suppression System Single Failure Limits]


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 theprimary and backup fire suppression capability.

Compliance StatementN/A

Compliance Basis

No areas at HBRSEP are required to beprotected by both primary and backupgaseous fire suppression systems.



Attachment A - NEI 04-02 Table B-1 Transition of Fundamental

Chapter 3 Reference: 3.10.5 [Gaseous Suppression System Disarming Controls]


Provisions for locally disarming automatic gaseous suppression systems shall be secured and under

strict administrative control.





OP-809,Diesel Generator Carbon Dioxide Suppression System ALL

OPS-NGGC- 1308,Plant Status Control ALL


Chapter 3 Reference: 3.10.6 [Gaseous Suppression System C02 Limitations)


Total flooding carbon dioxide systems shall not be used in normally occupied areas.




HBR2-9717,Fire Area/Zone Locations ALL


Chapter 3 Reference: 3.10.7 [Gaseous Suppression System C02 Warnings]




3.10.7

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 BasisSee proposed modification pertinent to

NFPA 805 Chapter 3, Section 3.10.7

compliance in Attachment "S", Table S-2

of the Transition Report.




RNP-M/BMRK-1007,Code Compliance Evaluation NFPA 12, Carbon Code Sections 122 & 1-6.2



Chapter 3 Reference: 3.10.8 [Gaseous Suppression System C02 Required Disarming]


Positive mechanical means shall be provided to lock out total flooding carbon dioxide systems during

work in the protected space.



Reference Document DoDetails

OP-805,Carbon Dioxide Suppression System Section 8.3

OP-809,Diesel Generator Carbon Dioxide Suppression System Section 8.3


Chapter 3 Reference: 3.10.9 [Gaseous Suppression System Cooling Considerations]


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.



Reference Document Doc Malls

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1

BTP APCSB 9.5-1,Guideline for Fire Protection for Nuclear Power Appendix A Section E.4 & E.5

Plants Docketed Prior to July 1, 1976




Chapter 3 Reference: 3.10.10 [Gaseous Suppression System Decomposition Issues]

Chapter 3 Requirement: 3.10.10Particular attention shall be given to corrosive characteristics of agent decomposition products on

safety systems.

Compliance Statement Compliance Bas• s



BTP APCSB 9.5-1,Guideline for Fire Protection for Nuclear Power Appendix A Section E.4 & E.5

Plants Docketed Prior to July 1, 1976

UFSAR,HBR 2 Updated Final Safety Analysis Report (FSAR) Section 9.5.1.1


Chapter 3 Reference: 3.11 Passive Fire Protection Features

Chapter 3 Requirement: 3.11 Passive Fire Protection Features.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.


N/A

Compliance Basis

N/A - General statement; No technicalrequirements.


Chapter 3 Reference: 3.11.1 Building Separation.

Chapter 3 Requirement: 3.11.1 Building Separation.

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, Reconmended 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.


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsCompliance Statement Compliance Basis


Complies via Engineering Evaluation COMPLIES VIA ENGINEERING

EVALUATION: Design and installation

deviations pertaining to passive fire

protection features are evaluated on a fire

zone basis at HBRSEP in calculations

RNP-M/MECH-1672 through RNP-

M/MECH-1696 or EC packages as

installed in the plant.

Reference Document DocDetails

RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier ALL

Penetration Seals in Fire Zone 1









RNP-M/MECH-1677, Evaluation of Non-Standard Fire Barrier ALL








RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier ALL












RNP-M/MECH- 1688, Evaluation of Non-Standard Fire Barrier ALL



Attachment A - NEI 04-02 Table B-1 Transition of FundamentalFire Protection Program & Design ElementsDuke EnerlRNP-MECH- 1689Evaluation of Non-Standard Fire Barrier


RNP-M/MECH- 1690, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 19

RNP-M/MECH-1683, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 12

RNP-M/MECH-1691 ,Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 20


RNP-M/MECH-1693,Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 22

RNP-M/MECH- 1694, Evaluation of Non-Standard Fire BarrierPenetration Seals in Fire Zone 23





HBR2-9717,Fire Area/Zone Locations

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

Sections 4.11 & 4.14

ALL

ALL


Chapter 3 Reference: 3.11.2 Fire Barriers.

Chapter 3 Requirement: 3.11.2 Fire Barriers.

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.


Complies


Compliance BasisCOMPLIES: No Additional Clarification



deviations pertaining to passive fire

protiction features are evaluated on a fire

zone basis at HBRSEP in calculations

RNP-M/MECH-1672 through RNP-

M/MECH-1696 or EC packages as

installed in the plant.

HBRSEP LAR Rev 1

Page A-69


Attachment A - NEI 04-02 Table B-1 Transition of Fundamentalkire Prnrtion-Empn_DUke a Pnrn,,

Reference Document

RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier






RNP-M/MECH- 1675,Evaluation of Non-Standard Fire Barrier












RNP-M/MECH- 1681, Evaluation of Non-Standard Fire Barrier

























DALDL lls

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL


Duke EnergyPenetration Seals in Fire Zone 22

RNP-M/MECH-1 694,Evaluation of Non-Standard Fire Barrier







FP-014,Control of Fire Barrier Penetrations

RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration

Seals

EE-87-0166,Evaluation Of Concrete Brick "Rubble" Fire Barrier

Penetration Seals

ESR-97-405,Evaluation of Pyrocrete Fire Barrier Designs

EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting

ESR-94-1003,Discrepancy Resolution For RNP2-M-063



ALL

ALL

ALL

Sections 4.11 & 4.14

ALL

ALL

ALL

ALL

ALL

ALL

ALL


Chapter 3 Reference: 3.11.3 Fire Barrier Penetrations.

Chapter 3 Requirement: 3.11.3* Fire Barrier Penetrations.

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.


Section (1):Complies via Engineering Evaluation

Compliance Basis

Section (1):Design and installation deviationspertaining to passive fire protectionfeatures are evaluated on a fire zone basisat HBRSEP in calculations RNP-M/MECH-1672 through RNP-M/MECH-1696 or ECpackages as installed in the plant.

HBRSEP complies with NFPA 80 asevaluated in the applicable portions ofRNP-M/BMRK-1003.



RNP-M/MECH-1672, Evaluation of Non-Standard Fire Barrier














RNP- M/MECH- 1679 Evaluation of Non-Standard Fire Barrier




RNP-M/MECH-1681 ,Evaluation of Non-Standard Fire Barrier




















RNP-M/MECH-1691 Evaluation of Non-Standard Fire Barrier






HBRSEP LAR Rev 1


Doc Dtals

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

Page A-72

Duke Energy







RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80

STANDARD FOR FIRE DOORS AND WINDOWS



Seals

RNP-M/MECH-1 670,Evaluation Of Concrete Hatch Covers

EE-90-0104,Generic Evaluation Of HVAC Fire Damper And Fire

Door Installation Discrepancies


ALL

ALL

ALL

ALL

Appendix 9.5.1A-7

ALL

ALL

Chapter 3 Requirement: (2) NFPA 90A,



Evaluation

Standard for the Installation of Air-Conditioning and Ventilating Systems.

Compliance Basis

Section (2):

Design and installation deviations

pertaining to passive fire protection

features are evaluated on a fire zone basis

at HBRSEP in calculations RNP-M/MECH-

1672 through RNP-M/MECH-1696 or EC

packages as installed in the plant.

HBRSEP complies with NFPA 90A as

evaluated in the applicable portions of

RNP-M/BMRK-1004.

Reference Document









RNP-M/MECH-1676,Ev luation of Non-Standard Fire Barrier

Penetration Seals in FireZone 5





Doc Details

ALL

ALL

ALL

ALL

ALL

ALL

ALL


Duke Energy





RNP-M/MECH- 1681 ,Evaluation of Non-Standard Fire Barrier
















RNP-M/MECH- 1687 Evaluation of Non-Standard Fire Barrier
















RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR

NFPA 90A 1976 & 1985 EDITIONS AIR CONDITIONING


RNP2-M-063 Selection of 3 Hour Fire Rated Barrier Penetration

Seals


EE-90-0104,Generic Evaluation Of HVAC Fire Damper And Fire

Door Installation Discrepancies


ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

Appendix 9.5.1A-7

ALL

ALL

ALL


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalIii 11W-Pperav E-re P-rotecton P-roprlms JC. I IPCIFl -ments

Chapter 3 Requirement: (3) NFPA 101, Life Safety Code

Exception: Where fire area boundaries are not wall-to-wall, floor-to-ceiling boundaries with allpenetrations sealed to the fire rating required of the boundaries, a performance-based analysis shall berequired to assess the adequacy of fire barrier forming the fire boundary to determine if the barrier will

withstand the fire effects of the hazards in the area. Openings in fire barriers shall be permitted to beprotected by other means as acceptable to the AHJ.


Section (3): Complies with Clarification

Compliance BasisSection (3): HBRSEP complies with

clarification with NFPA 101. HBRSEP

complies with NFPA 101 with regards to

fire rated door assemblies since NFPA

101, Section 8.3.3.1 refers to NFPA 80,

which is evaluated in RNP-M/BMRK-1003.

HBRSEP complies with NFPA 101 with

regards to rated fire dampers since NFPA

101, Section 9.2.1 refers to NFPA 90A,

which is evaluated in RNP-M/BMRK-1004


RNP-M/BMRK-1003,CODE COMPLIANCE EVALUATION NFPA 80 ALL

STANDARD FOR FIRE DOORS AND WINDOWS

RNP-M/BMRK-1004,CODE COMPLIANCE EVALUATION FOR ALL

NFPA 90A 1976 & 1985 EDITIONS AIR CONDITIONING

NFPA 101,Life Safety Code, 2009 Edition Sections 8.3.3.1 & 9.2.1


Chapter 3 Reference: 3.11.4 Through Penetration Fire Stops.

Chapter 3 Requirement: 3.11.4* Through Penetration Fire Stops.

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.


Section (a): Section (a):


Complies via Engineering Evaluation COMPLIES VIA ENGINEERING



EVALUATION: Design and installationdeviations pertaining to passive fireprotection features are evaluated on a firezone basis at HBRSEP in calculationsRNP-M/MECH-1672 through RNP-M/MECH-1696 or EC packages asinstalled in the plant.

Engineering evaluations were developedto analyze the acceptability of the typicalpenetration seal designs utilized atHBRSEP.

Reference Document Doe Details

RNP-M/MECH-1672,Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 1








RNP-M/MECH-1679, Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 8


RNP-M/MECH-1681 Evaluation of Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 10




RNP-M/MECH-1685,Evaluation kf Non-Standard Fire Barrier ALLPenetration Seals in Fire Zone 14




Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsPenetration Seals in Fire Zone 16



















HBR2-09716,Fire Barrier Penetrations ALL

GID/R87038/0014,Design Basis Document; Fire Barrier System Sections 4.1.3 & 4.1.4

FP-014,Control of Fire Barrier Penetrations ALL

RNP2-M-063,Selection of 3 Hour Fire Rated Barrier Penetration ALL

Seals

EE-93-0095,Evaluation of Reduced Pre Soak Time for Grouting ALL

ESR-94-1003,Discrepancy Resolution For RNP2-M-063 ALL

EE-90-0025,Past And Present Operability Of Steam Generator ALL

Blowdown Line Penetration Seals (Penetrations CP-2674 And CP-

5612)

EE-93-0043,Evaluation Of Temporary Fire Barrier Penetration Seals ALL

Between Fire Zone 11 And 24

ESR-98-0221, Penetration Seals Containing Copper Piping and ALL

Tubing

RNP-M/MECH-1671, Evaluation Of Large Bore Piping Penetrations ALL

Chapter 3 Requirement: (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 ir diameter shall be sealed at the fire barrier

with a fire-rated internal seal unless the conduit extends greater than 5 ft (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.


Attachment A - NEI 04-02 Table B-1 Transition of FundamentalDuke Energy Fire Protection Program & Design ElementsCompliance Statement Compliance Basis

Section (b): Complies Section (b):

COMPLIES: No Additional Clarification




deviations pertaining to passive fireprotection features are evaluated on a fire

zone basis at RNP in calculations RNP-

M/MECH-1672 through RNP-M/MECH-

1696 or EC packages as installed in the

plant.

Reference Document DocDtails















RNP-M/MECH- 1679,Evaluation of Non-Standard Fire Barrier ALL












RNP-M/ME!•H- 1685,Evaluation of Non-Standard Fire Barrier ALL






Duke EnergyPenetration Seals in Fire Zone 16



















GID/R87038/0014,Design Basis Document; Fire Barrier System

CTL# CRE093-4324,Conduit Fire Test of One Hundred One

Electrical Conduit Penetrations


Seals


ESR-94-0930,Fire Barriers

ESR-94-1103,Evaluation for 12" of Dow Corning Foam For Conduit

Seals

NED-B/MECH- 1001, Fire Resistance of Capped Conduits


ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

ALL

Section 2.2.2.1, Table A.4.0-1

ALL

ALL

ALL

ALL

ALL

ALL


Chapter 3 Reference: 3.11.5 Electrical Raceway Fire Barrier Systems (ERFBS).

Chapter 3 Requirement: 3.11.5* Electrical Raceway Fire Barrier Systems (ERFBS).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 meet the acceptance criteria of NRC Generic Letter

86-10, Supplement 1, "Fire Endurance ITest 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 Shutdown Training Within the

Same Fire Area," Supplement 1, 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, Supplement1, are acceptable providing that the system successfully met the limiting end point temperature

requirements as specified by the AHJ at the time of acceptance.




GID/R87038/0014,Design Basis Document; Fire Barrier System Section 4.4



Duke Energy Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review

B. NEI 04-02 Table B-2 - Nuclear Safety Capability Assessment -Methodology Review

96 Pages Attached

HBRSEP LAR Rev I 7age B-1

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy

Table B-2 Nuclear Safety Capability Assessment Methodology Review

NFPA 805 Section 2.4.2.1 Nuclear Safety Capability System and Equipment Selection

NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire

Requirement event shall be developed. The equipment list shall contain an inventory of those critical componentsrequired 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 Guidance

This section discusses a generic deterministic methodology and criteria that licensees can use to3 Deterministic perform a post-fire safe shutdown analysis to address regulatory requirements. The plant-specific

Methodology 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 inthis 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 ill.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.

Additional information is provided in Appendix B to this document.

Applicability Comments

Applicable

AlignmentAlignment Basis

Statement

Robinson Nuclear Plant's (HBRSEP) Safe Shutdown Methodology was reviewed against theAligns requirements of Appendix R Sections IlI.G, liI.J, and 1l1.L as required by 10CFR50.48(b). NRC review

and approval of the HBRSEP safe shutdown methodology is contained in a series of Safety

Evaluation Reports.

For the re-validation of the Safe Shutdown Analysis (SSA) performed prior to and in conjunction with

the transition process, NEI 00-01, Revision 1 (which has since been updated to revision 2) was oneof the references used in developing the circuit analysis procedure, which is now captured in FIR-

NGGC-0101, Revision 2. Except as noted in this document, the plant's methodology meets the

guidelines of NEI 00-01.

Comments

HBRSEP LAR Rev 1 Page B-2

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuk••Energy Capabil . dolommReview

Reference Document

FIR-NGGC-0101, Fire Protection Nuclear Safety Capability

Assessment (NSCA)

NLS-84-516, Fire Protection Rule - Alternate Safe Shutdown

Capability - Sections III.G.3 and III.L of Appendix R to 10 CFR 50 -

H.B. Robinson Steam Electric Plant Unit No. 2

NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10

CFR 50, Items IlI.G.3 and Il.L; Alternate Safe Shutdown Capability -

H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106

Doc Detail

Section 3.2, 3.34



NFPA 805 A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireRequirement 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




NE IRef NEI 00-01 GuidanceThis section discusses the identification of systems available and necessary to perform the required

3.1[Itro]n Safems safe shutdown functions. It also provides information on the process for combining these systems

Shutdown Systems into safe shutdown paths. Appendix R Section III.G.1.a requires that the capability to achieve and

and Path maintain hot shutdown be free of fire damage. It is expected that the term "free of fire damage" will

Development be further clarified in a forthcoming Regulatory Issue Summary. Appendix R Section IIl.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.

[Refer to hard copy of NEI 00-01 for Figure 3-11


Applicable

AlignmentSttment

Aligns

Alignment Basg

RNP-E/ELEC-1216 identifies the systems and components necessary achieve and maintan safeshutdown.

For the re-4lidation of the Safe Shutdown Analysis (SSA) performed prior to a4d in conjunction withthe transition process, NEI 00-01, Revision 1 was one of the references used in developing the circuitanalysis procedure, which is now captured in FIR-NGGC-0101, revision 1. Except as noted in thisdocument, the plant's methodology meets the guidelines of NEI 00-01.



Comments

Reference Document Doc Detail

FIR-NGGC-0101, Fire Protection Nuclear Safety CapabilityAssessment (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.0Robinson Nuclear Plant




required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI00-01 Ref NEI 00-01 GuidanceThe goal of post-fire safe shutdown is to assure that one train of shutdown systems, structures, and

3.1 [B, Goals] Safe components remains free of fire damage for a single fire in any single plant fire area. This goal isShutdown Systems accomplished by determining those functions important to achieve and maintain hot shutdown. Safeand Path shutdown systems are selected so that the capability to perform these required functions is a part ofDevelopment each safe shutdown path. The functions important to post-fire safe shutdown generally include, but

are not limited to the following:

Reactivity controlPressure control systemsInventory control systemsDecay heat removal systemsProcess monitoringSupport systems- Electrical systems- Cooling systems

These functions are of importance because they have a direct bearing on the safe shutdown goal ofbeing able to achieve and maintain hot shutdown which ensures the integrity of the fuel, the reactorpressure vessel, and the primary containment. If these functions are preserved, then the plant will besafe because the fuel, the reactor and the primary containment will not be damaged. By assuringthat this equipment is not damaged and remains functional, the protection of the health and safety ofthe public is assured.


Applicable

AlianmentAlignmentBasis


Duke Energytatement

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review

RNP-E/ELEC-1216 identifies the safe shutdown functions.Aligns

Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.1

Robinson Nuclear Plant




Requirement 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 toachieve and maintain the nuclear safety functions and components whose fire-induced failure could



0-01Ref NEI 00-01 Guidance

In addition to the above listed functions, Generic Letter 81-12 specifies consideration of associated3.1 [C, Spurious circuits with the potential for spurious equipment operation and/or loss of power source, and the

Operations] Safe common enclosure failures. Spurious operations/actuations can affect the accomplishment of the

Shutdown Systems post-fire safe shutdown functions listed above. Typical examples of the effects of the spurious

and Path operations of concern are the following:

Development

- 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 reactorpressure vessel or the primary containment. Common power source and common enclosure

concerns could also affect these and must be addressed.

Applicability CommentsApplicable

AlignmentStateent

Alignment Basi

AlignsHBRSEP has considered spurious operation, common power sources, and common enclosure

concerns that would cause a circuit to be considered an associated circuit.

RCS isolation valves (such as the RHR Pump suction valves) are defined as high/low pressure


Duke EnergyAttachment B - NEI 04-02 Table B-2 Nuclear Safety

Capability Assessment Methodology Reviewinterface boundary valves if their spurious operation could lead to the rupture of low pressure piping

or a loss of RCS inventory that exceeds the RCS makeup capability. Such interface boundary valves

are subject to more stringent circuit analysis criteria, and are identified in FSSPMD by the HLP flag.

This high/low pressure interface boundary valve definition is conservative with respect to that in in

Appendix C of NEI 00-01 and NFPA-805 FAQ 06-0006.

During the re-validation, the definition from the previous SSA was carried forward for conservatism.Thus, some components are classified as high-low interfaces which do not meet the above definition

since their spurious opening will not result in a rupture of downstream piping and a subsequent

intersystem LOCA. Robinson may choose to remove the classification of these components as high-

low interfaces at a future date.

Comments

Reference Document DoDetal

FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.4, 3.34

Assessment (NSCA)

FSSPMD, Fire Safe Shutdown Program Manager Database










The following criteria and assumptions may be considered when identifying systems available and3. 1.1 Criteria/ necessary to perform the required safe shutdown functions and combining these systems into safe

Assumptions shutdown paths.


Applicable

AlignmentAligmentAlignment BasesStatement

This is generic introductory information and contains no specific requirements.N/A

Comments



Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review





achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear


NEI00-01Ref NEI 00-01 Guidance[BWR] GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths For The

3.1.1.1 [GE BWR BWR" addresses the systems and equipment originally designed into the GE boiling water reactors

Paths] (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.


Not Applicable

Alignment

Statement

HBRSEP is a PWR. This guidance is specific to BWRs.N/A

Comments








NEL00-IRef

3.1.1.10 [Manual I

Automatic Initiation

of Systems]

NEI 00-01 Guidance

Manual initiation from the main control room or emergency control stations of systems required to

achieve and maintain safe shutdown s acceptable where permitted by current regulations or

approved by NRC; automatic initiatio r of systems selected for safe shutdown is not required but may

be included as an option.

ApplicabilityApplicable



AlignmenttAligimitAlignment Basis

Statement

Reliance on the automatic logic for safe shutdown systems is not required, but if credited needs to beAligns appropriately evaluated as being free of fire damage. The only automatic logics evaluated at

HBRSEP are the Emergency Diesel Generator (EDG) Automatic Sequencing Logic and the Fast Bus

Transfer logic. These logics are incorporated in the overall SSD fault tree for HBRSEP.

Comments

Reference Document DocDetaal

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.4.3 and 2.2.4.4










NE[D00-1Ref NEI 00-01 Guidance

Where a single fire can impact more than one unit of a multi-unit plant, the ability to achieve and3.1.1.11 [Multiple maintain safe shutdown for each affected unit must be demonstrated.

Affected Units]


Not Applicable

AlignmentAlignment BasisStatement

Robinson is a single unit site.N/A

Comments



NFPA 805

Requirement

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



Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria of Section 1.5. Components required to




NE[ 0-1Ref

3.1.1.2 [SRVs / LP

Systems]

NEI 00-01 Guidance

[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 lII.G.2. The NRC has accepted the

BWROG position and issued an SER dated Dec. 12, 2000.

A4pplicabilit Comments

Not Applicable

Alignment Alignment Basis

StatementHBRSEP is a PWR. This guidance is specific to BWRs.

N/A

Comments








NEI0 Ref NEI 00-01 Guidance

[PWR] Generic Letter 86-10, Enclosure 2, Section 5.3.5 specifies that hot shutdown can be3.1.1.3 [Pressurizer maintained without the use of pressurizer heaters (i.e., pressure control is provided by controlling the

Heaters] 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.


Applicable

AlignmentStatement

Aliganment Basis


Duke Energy

Aligns with intent


In most fire areas, HBRSEP does not rely on the use of pressurizer heaters to maintain hot

shutdown.

RCS pressure is controlled during hot shutdown and cooldown by controlling the rate of charging to

the RCS. Pressurizer heaters and/or auxiliary spray reduces operator burden. Neither component is

required to provide adequate pressure control if charging is available. 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.

Use of the SI Pumps in lieu of the Charging Pumps may be required for certain shutdown scenarios.

The RCS would have to be depressurized to less than operating pressure of the SI pumps.

Pressurizer heaters are credited to stabilize pressure transients when SI pumps are operated

intermittently.

The NEI guidance does not prevent the use of pressurizer heaters, but only serves to note that they

are generally not required.

Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.1.2 (2)Robinson Nuclear Plant





NE[D00-01 R NEI 00-01 GuidanceThe classification of shutdown capability as alternative shutdown is made independent of the

3.1.1.4 [Alternative selection of systems used for shutdown. Alternative shutdown capability is determined based on anShutdown inability to assure the availability of a redundant safe shutdown path. Compliance to the separationCapability] 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),exeml~tions, deviations, GL 86-10 fire hazards analyses or fire protectionl design change evaluations,as apl ropriate. These may also be used in conjunction with alternative shutdown capability.




Alignment

StatementAlignment Basis

AlignsThe plant's alternate and dedicated safe shutdown systems and strategies were reviewed and

approved in the supplemental SER.

Comments


NLS-85-732, Supplemental Safety Evaluation for Appendix R to 10

CFR 50, Items Ill.G.3 and ll.L; Alternate Safe Shutdown Capability -

H.B. Robinson Steam Plant, Unit 2 - TAC No. 60106









NEID0-1Ref NEI 00-01 Guidance

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

Conditions] 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.


Applicable

Alignment Basis

Statement

This is a basic assumption for all safe shutdown analyses and applies to HBRSEP.Aligns

Comments

Reference Document DoDetail

FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1 (2)

Assessment (NSCA)


HBRSEP LAR Rev 1 Page B-1 1

Duke EnergyNFPA 805 Section


2.4.2.1 Nuclear Safety Capability System and Equipment Selection

NFPA 805

Requirement







NE-I00-0Ref NEI 00-01 Guidance

No Final Safety Analysis Report accidents or other design basis events (e.g. loss of coolant accident,3.1.1.6 [Other earthquake), single failures or non-fire induced transients need be considered in conjunction with theEvents in fire.Conjunction withFire]


Applicable

Alignm entAlg m n BAliganment Basis

StatementThis is a basic assumption for all safe shutdown analyses and applies to HBRSEP.

Aligns

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.1 (6,7,8)Assessment (NSCA)





3.1.1.7 [ Offsite

Power]

NEI 00-01 Guidance

For the case of redundant shutdown, offsite power may be credited if demonstrated to be free of firedamage. Offsite power should be assumed to remain available for those cases where its availabilitymay adversely impact safety (i.e., reliance cannot be placed on fire causing a loss of offsite power ifthe consequences of offsite power availability are more severe than its presumed loss). No creditshould be taken for a fire causing a loss of offsite power. For areas where train separation cannot be


Duke Energy

Applicability

Applicable


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.

Comments

AlignmentStatement

Aligns

Alignment Basis

For fire areas that use redundant shutdown capabilities offsite power is credited unless the fireimpacts equipment required to support offsite power. If the fire impacts offsite power, at least oneonsite power source is available to provide the required power.

For areas that use alternative / dedicated shutdown, a LOOP is assumed.

In the analysis the LOOP is not credited for preventing or terminating spurious operations orpositioning SSE in its required position. Steps in the procedures insure that the appropriate actionsare taken to line up SSE and deal with potential spurious equipment operations.


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.






required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure could



NE[O0-1Ref NEI 00-01 Guidance

Post-fire safe shutdown systems and components are not required to be safety-related.3.1.1.8 [Safety-

Related Equipment]


AlignmentStateme~nt

Alignment Basis




Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.1.2Robinson Nuclear Plant




required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear



The post-fire safe shutdown analysis assumes a 72-hour coping period starting with a reactor3.1.1.9 [72 Hour scram/trip. Fire-induced impacts that provide no adverse consequences to hot shutdown within thisCoping] 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.


Applicable

AlignmentAligmentAlignment BasisStatement

NFPA 805 does not require a plant to transition to cold shutdown in the event of a fire. The fire area-Aligns by-fire area assessment documents the method of accomplishment of the NFPA 805 performance

goals, including an optional transition to cold shutdown. For all fires at HBRSEP, the systems andequipment required to place the plant in a safe and stable condition are available following a fireoccurring while the plant is at power without regard to a specific mission time or event copingduration.

Comments

Reference Document DocDetail

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 1.5.1 and 1.5.2Robinson Nuclear Plant


NFPA 805 Section

HBRSEP LAR Rev 1

2.4.2.1 Nuclear Safety Capability System and Equipment Selection

Page B-14


Capability Assessment Methodology Review

NFPA 805

Requirement







NEI -0Ref NEI 00-01 Guidance

The following discussion on each of these shutdown functions provides guidance for selecting the3.1.2 Shutdown systems and equipment required for safe shutdown. For additional information on BWR systemFunctions selection, refer to GE Report GE-NE-T43-00002-00-01-R01 entitled "Original Safe Shutdown Paths

for the BWR."


Applicable


This is an introductory section with no specific requirements. The GE information does not apply toAligns HBRSEP.

Comments









NE[D00-01Ref

3.1.2.1 Reactivity

Control

Applicability

Applicable

HBRSEP LAR Rev 1

NEI 00-01 Guidance

[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.

Comments

Page B-15


Aligment Alignment BasisStatement

Reactivity control is provided by insertion of control rods via a reactor trip. Long term reactivityAligns control is provided by boron addition via the charging pumps or safety injection pumps taking suction

from the Reactor Water Storage Tank.

Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.2.2.5








NEI 0-01Rf NEI 00-01 Guidance

The systems discussed in this section are examples of systems that can be used for pressure control.3.1.2.2 Pressure This does not restrict the use of other systems for this purpose.

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.


Applicable

AlignmentStatemetnt

Alignment Bas

The Reactor Coolant Pressure Control function uses the same components as the RCS Inventory


Duke EnergyAligns


Control function. RCS Pressure is controlled by controlling the rate of charging to the RCS. 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.

Pressurizer heaters are credited for pressure control when the SI pumps are used for makeup. When

the SI pumps are utilized for makeup, the RCS is de-pressurized to the SI Pump operating pressure

by cycling a Pressurizer PORV.

Comments

Reference Document Doc DetailRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 2.2.1.2










[BWR] Systems selected for the inventory control function should be capable of supplying sufficient3.1.2.3 Inventory reactor coolant to achieve and maintain hot shutdown. Manual initiation of these systems is

Control 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.


AlignmentStatementf

Aligns

Alignment Basis

The RCS Inventory Control function is required to restore and maintain RCS integrity and reactor

coolant makeup capability to compensate for RCS fluid losses (i.e. RCP seal leak-off) and shrinkage

during cooldovn. Reactor Coolant Inventory Control is accomplished by the following actions:

- RCS Isolation - RV Head Vents and Pressurizer PORVs (RCS)

- Normal Letdown Isolation (CVCS)

- Excess Letdown Isolation (CVCS)

- RHR Isolation (RHR)



Capability Assessment Methodology Review- Charging/makeup from the RWST via the charging pumps or SI pumps. The SI pumps require the

CCW system for cooling. (CVCS, SI, CCW, SW)

- Use of the SI Pumps for RCS Makeup requires RCS depressurization via the Pressurizer PORVs.

- RCP Seal Cooling via Seal Injection and/or Thermal Barrier Cooling (CVCS, CCW, SW)

Comments













[BWR] Systems selected for the decay heat removal function(s) should be capable of:3.1.2.4 Decay Heat

Removal - 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.

[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.


Applicable

AlignmentStatement

Alignment Basis

HBRSEP uses the Auxiliary Feedwater System (AFW) and the Main Steam System (MS) to remove

decay heat from the reactor through the steam generators for Hot Shutdown. The Residual HeatAligns

HBRSEP LAR Rev 1 Page B-1 8

Duke EnergyAttachment 8 - NEI 04-02 Table B-2 Nuclear Safety

Capability Assessment Methodology ReviewRemoval (RHR) System is available at temperatures below 350F and pressures less than 375 psig to

remove decay heat and continue the reactor cooldown to cold shutdown.

Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.3 and 2.2.1.4










NEL 01Ref

3.1.2.5 ProcessMonitoring

NEI 00-01 Guidance

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 theNRC for meeting the process monitoring function. This instrumentation is that which monitors theprocess variables necessary to perform and control the functions specified in Appendix R SectionIII.L.1. Such instrumentation must be demonstrated to remain unaffected by the fire. The IN 84-09 listof process monitoring is applied to alternative shutdown (III.G.3). IN 84-09 did not identify specificinstruments for process monitoring to be applied to redundant shutdown (III.G.1 and III.G.2). Ingeneral, process monitoring instruments similar to those listed below are needed to successfully useexisting operating procedures (including Abnormal Operating Procedures).

BWR- Reactor coolant level and pressure- Suppression pool level and temperature

- Emergency or isolation condenser level

- Diagnostic instrumentation for safe shutdown systems- Level indication for tanks needed for safe shutdown

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


Duke Energy



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.

Comments

AlignmentStatement

Aligns

Alignment Bas'

The process monitoring function is capable of providing direct reading of those process variablesnecessary for plant operators to perform and/or control identified safe shutdown functions. Plantmonitoring instrumentation, in the context of post-fire safe shutdown operation, consists of thoseminimal instrument channels or local gauges/indicators necessary to monitor the operation of primary

shutdown components and systems, and the operation of those components or systems that providerequired support functions. The parameters to be monitored during post-fire shutdown operations,along with the credited instruments, are summarized on Table 2-2 of RNP-E/ELEC-1216.

Comments


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.8, 2.2.2.12, and Table 2-2








prevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEI01 Ref NEI 00-01 Guidance

[Blank Heading - No specific guidance]3.1.2.6 Support

Systems

Notlicbi CommentsNot Applicable

N/A

Comments

Alignment Basis I

Support system requirements will be addressed under the corresponding NEI 00-0 1 sub-section.


Attachment B - NEI 04-02 Table B-2 Nuclear Safetyf'gn.hilifw A . .. m- - f KA~fhn-nlný I D i-WniL, C:-~~r~

-J =.l~l.l -_ L:- , llLy • O O l•lL V•AV•.IrI •









VV

NiE[00-01 Ref

3.1.2.6.1 Electrical

Systems

ApplicabilityApplicableI

NEI 00-01 Guidance

AC Distribution System

Power for the Appendix R safe shutdown equipment is typically provided by a medium voltage systemsuch as 4.16 KV Class 1E busses either directly from the busses or through step downtransformers/load centers/distribution panels for 600, 480 or 120 VAC loads. For redundant safeshutdown performed in accordance with the requirements of Appendix R Section III.G.1 and 2, powermay be supplied from either offsite power sources or the emergency diesel generator depending onwhich has been demonstrated to be free of fire damage. No credit should be taken for a fire causinga 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 controlpanels including switchgear breaker controls. The 125VDC distribution panels may also supply powerto the 120VAC distribution panels via static inverters. These distribution panels typically supplypower 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 arenecessary to supply any required control power during the interim time period required for the dieselgenerators to become operational. Once the diesels are operational, the 125 VDC distributionsystem can be powered from the diesels through the battery chargers.[BWR] Certain plants are also designed with a 250VDC Distribution System that supplies power toReactor Core Isolation Cooling and/or High Pressure Coolant Injection equipment.The DC control centers may also supply power to various small horsepower Appendix R safeshutdown system valves and pumps. If the DC system is relied upon to support safe shutdownwithout battery chargers being available, it must be verified that sufficient battery capacity exists tosupport the necessary loads for sufficient time (either until power is restored, or the loads are nolonger required to operate).

Comments

AlignmentStatement

Alignment Basis

AlignsThe Electrical Distribution System provides 4160VAC, 480VAC, 120VAC and 125VDC power from

off-site (115KV Grid) and onsite sources (EDGs and DSDG) to safe shutdown electrical loads. The



Capability Assessment Methodology Reviewfuel oil systems associated with the onsite power supplies are also included in the analysis.


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.7, 2.2.2.11Robinson Nuclear Plant




required to achieve the nuclear safety performance criteria of Section 1.5. Components required toachieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclear


NEI 00-01 Ref NEI 00-01 GuidanceHVAC Systems

3.1.2.6.2 CoolingSystems [HVAC] 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 testmethods, and to assure protection for plant operations staff from the effects of fire (smoke, heat, toxicgases, 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.


AlignmentStatementt

ý ligns

Alignment Basi

Plant ventilation systems are required fun tional to provide environmental conditions that supportcontinuous occupancy or safe shutdown dquipment operation in the following area:

- EDG Rooms- Main Control Room- Motor Driven AFW Pump Room



- SI Pump Room Capability Assessment Methodology Review

The Main Control Room ventilation system includes refrigerant units that are cooled by the SW

system.


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.2.1.9, 2.2.2.13, Table 2-3Robinson Nuclear Plant





NEI 0-01Ref NEI 00-01 GuidanceVarious cooling water systems may be required to support safe shutdown system operation, basedon plant-specific considerations. Typical uses include:

Systems [Main - RHR/SDC/DH Heat Exchanger cooling waterSection] - Safe shutdown pump cooling (seal coolers, oil coolers)

- Diesel generator cooling

- HVAC system cooling water


Applicable

AlignmentAlignment BStatement

The Component Cooling Water and Service Water Systems provide cooling to the various safeAligns shutdown loads.

Comments

Reference Document DocDetailRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis fir H.B. Sections 2.2.1.5, 2.2.1.6Robinson Nuclear Plant












Refer to Figure 3-2 for a flowchart illustrating the various steps involved in selecting safe shutdown3.1.3 Methodology systems and developing the shutdown paths.

for Shutdown

System Selection The following methodology may be used to define the safe shutdown systems and paths for an

Appendix R analysis:

[Refer to hard copy of NEI 00-01 for Figure 3-2]


Applicable

Alignment

StatementThis is an introductory statement and provides no requirements. The sub-paragraphs with specific

N/A requirements are addressed separately as required.

Comments









NE[D00-01Ref

3.1.3.1 Identify safe

shutdown functions

NEI 00-01 Guidance'

Review available doumentation to obtain an understanding of the available plant systers and the

functions required to achieve and maintain safe shutdown. Documents such as the following may be

reviewed:

- Operating Procedures (Normal, Emergency, Abnormal)



Capability Assessment Methodology Review- 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"

Applicabllity

Applicable

Comments

Alignment Alignment BasisStatement

Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire Hazard

Aligns Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system

comprising the safe shutdown paths, the mechanical or electrical equipment required for the

operation of the system and the equipment whose spurious operation could affect the performance of

the safe shutdown systems were identified.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.1

Assessment (NSCA)










NEI00-01Ref

3.1.3.2 Identify

Combinations of

Systems that

Satisfy Each Safe

Shutdown Function

Applicability

HBRSEP LAR Rev 1

NEI 00-01 GuidanceGiven the criteria/assumptions defined in Section 3.1.1, identify the available combinations ofsystems capable of achieving the safe shutdown functions of reactivity control, pressure control,inventory control, decay heat removal, process monitoring, and support systems such as electricaland coolin systems (refer to Section 3.1.2). This selection process does notlrestrict the use of othersystems. In addition to achieving the required safe shutdown functions, consider spurious operationsand power supply issues that could impact the required safe shutdown function.

Comments

Page B-25


Applicable

AlignmentStatment

Alignment Basis

AlignsIn accordance with the provisions of 10CFR50, Appendix R, Section lIl.G, at least one means of

achieving and maintaining safe shutdown conditions must remain available in the event of a fire in

any fire area. In developing an appropriate shutdown equipment complement to support this

requirement, it is necessary to categorize equipment into logical train-oriented groupings, identified as

shutdown categories; these categories are further defined as Alternate A and Alternate B. Although,

in many cases, the equipment complement selected corresponds closely to safety-related train

divisions, it should not be construed that Alternate A and B equipment automatically corresponds to

safety-related Train A and B equipment. In general, the Alternate A division constitutes the

equipment credited for safe shutdown outside the control room (i.e. dedicated shutdown) and the

Alternate B division constitutes the equipment credited for control room shutdown scenarios.












NEI0-01 Ref NEI 00-01 Guidance

Select combinations of systems with the capability of performing all of the required safe shutdown3.1.3.3 Define functions and designate this set of systems as a safe shutdown path. In many cases, safe shutdown

Combinations of paths may be defined on a divisional basis since the availability of electrical power and other support

Systems for Each systems must be demonstrated for each path.

Safe Shutdown

Path

ApplcabCommentsApplicable

AlmanmentStatement

Algnment BI

AlignsComponents have been grouped into Appendix R fire-safe shutdown systems according to theAppendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is



Capability Assessment Methodology Reviewdivided into one or more "success paths". Each success path represents a functionally independent

method of accomplishing a unique fire-safe shutdown function. Each success path is divided into two

or more redundant Appendix R fire-safe shutdown trains or "success path trains". Each success

path train will often be comprised of components from different plant systems necessary to

accomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safe

shutdown fault tree developed during the safe shutdown re-validation project.


RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.1.2.1 - 2.1.2.4, 2.2.4.1Robinson Nuclear Plant





NEI.0-1Ref NEI 00-01 GuidanceAssign a path designation to each combination of systems. The path will serve to document the

3.1.3.4 Assign combination of systems relied upon for safe shutdown in each fire area. Refer to Attachment 1 to thisShutdown Paths to document (NEI 00-01) for an example of a table illustrating how to document the variousEach Combination combinations of systems for selected shutdown paths.of Systems

Applicabfility Comments

Applicable


Components have been grouped into Appendix R fire-safe shutdown systems according to theAligns Appendix R fire-safe shutdown function they support. Each Appendix R fire-safe shutdown system is

divided into one or more "success paths". Each success path represents a functionally independentmethod of accomplishing a unique fire-safe shutdown function. Each success path is divided into twoor more redundant Appendix R fire-safe shutdown trains or "success path trains". Each successpath train will often be comprised of components from different plant systems necessary toaccomplish an Appendix R fire-safe shutdown function. These combinations are reflected in the safeshutdown fault tree developed during the safe shutdown re-validation project.

Comments






Sections 2.2.2.1 - 2.2.2.4, Table 2-1










The previous section described the methodology for selecting the systems and paths necessary to3.2 Safe Shutdown achieve and maintain safe shutdown for an exposure fire event (see Section 5.0 DEFINITIONS for

Equipment "Exposure Fire"). This section describes the criteria/assumptions and selection methodology for

Selection 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.


Applicable

Alignment BasisStatement

This is an introductory statement and provides no requirements. The sub-paragraphs with specificN/A requirements are addressed separately as required.

Comments



NFPA 805Reqluirement

A comprehensive list of systems and equipme t and their interrelationships to be analyzed for a fire

event shall be developed. The equipment list hall contain an inventory of those critical components







NEI 00-01 Ref NEI 00-01 GuidanceConsider the following criteria and assumptions when identifying equipment necessary to perform the

3.2.1 Criteria / required safe shutdown functions:Assumptions


Applicable

Agnment Alignment BasisStatement


Comments





achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

3.2.1.1 [PrimarySecondaryComponents]

NEI 00-01 Guidance

3.2.1.1 Safe shutdown equipment can be divided into two categories. Equipment may be categorizedas (1) primary components or (2) secondary components. Typically, the following types of equipmentare 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, turbinespeed 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, distributionpanels, etc.).

Secondary components are typically items found within the circuitry for a primary component. Theseprovide a supporting role to the overall circuit function. Some secondary components may provide anisolation 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 elemenis, 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 anoption, include secondary components with a primary component(s) that would be affected by firedamage to the secondary component. By doing this, the SSEL can be kept to a manageable size


Duke Energy



and the equipment included on the SSEL can be readily related to required post-fire safe shutdown

systems and functions.

Comments

Alignment ~ nmm~Alignment Basis

Statement

Components are not identified as primary or secondary. Components providing a "secondary"

Aligns function are either identified as safe shutdown components and included in the safe shutdown

equipment list, or have their applicable cables assigned to the primary component.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2

Assessment (NSCA)









NEL 00-01Ref NEI 00-01 Guidance

3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact3.2.1.2 [Fire their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping

Damage to materials, including tubing with brazed or soldered joints, are not included in this assumption). Fire

Mechanical damage should be evaluated with respect to the ability to manually open or close the valve should

Components (not this be necessary as a part of the post-fire safe shutdown scenario.

electrically

supervised)]


Applicable

AlegnmentStatement

Aligns

Alignment Basis

Due to the substantial nature of equipment and the nature and location of combustibles, fire will not

not impact the pressure boundary function. A fire does not cause a manual valve to change itsposition, Manual stroking of a valve once the fire is extinguished is evaluated as part of the manualaction feasibility study.



Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.1.3, 9.4.1

Assessment (NSCA)









N.EI00-01 Ref NEI 00-01 Guidance

Assume that manual valves are in their normal position as shown on P&lDs or in the plant operating3.2.1.3 [Manual procedures.

Valve Positions]


Applicable

Alignment Basis

Statement


Comments



Assessment (NSCA)





NFPA 805

Requirement

A comprehensive list of systems and equipment and their interrelationships to be analyzed for a fireevent shall be developed. The equipment list shall contain an inventory of those critical components



Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria of Section 1.5T Components required to

achieve and maintain the nuclear safety functions and components whose fire-induced failure couldprevent the operation or result in the maloperation of those components needed to meet the nuclearsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.


Assume that a check valve closes in the direction of potential flow diversion and seats properly with3.2.1.4 [Check sufficient leak tightness to prevent flow diversion. Therefore, check valves do not adversely affect the

Valves] flow rate capability of the safe shutdown systems being used for inventory control, decay heat

removal, equipment cooling or other related safe shutdown functions.


Applicable

AIflgnment Alignment Basis

Statement

FIR-NGGC-001 identifies that properly oriented check valves credited as system boundaries shouldAligns be included in the SSEL, and that those in the flow path need not be included. Check Valves

credited as boundaries are included in the SSEL, but the assumption that they are leak tight is

inherent in the analysis.

Comments



Assessment (NSCA)









NEI 00-01 Ref

3.2.1.5 [InstrumentFailures]

Applicability

Applicable

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Instruments (e.g., resistance temperature detectors, thermocouples, pressure transmitters, and flowtransmitters) are assumed to fail upscale, midscale, or downscale as a result of fire d mage,whichever is worse. An instrument performing a control function is assumed to provide an undesiredsignal to the control circuit.

Comments

Page B-32

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Canability Assessment Methodology Review

Alignment

Statement

Instruments exposed to the fire are assumed to fail. It is a generally accepted practice (that can beAligns verified based on a review of the fire area analysis) that instruments are assumed to fail to their worst

case position.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.3.2 and 9.4.1

Assessment (NSCA)








NEL 00-0 ef NEI 00-01 Guidance

Identify equipment that could spuriously operate or mal-operate and impact the performance of3.2.1.6 [Spurious equipment on a required safe shutdown path during the equipment selection phase. Consider Bin 1

Components] of RIS 2004-03 during the equipment identification process.


Applicable


Statement

FIR-NGGC-0101 states, "Electrically operated or controlled valves or dampers in the flow pathsAligns whose spurious operation could adversely affect system operation shall be included on the SSEL."

This is affirmed in the Section 2.1.2.5 of RNP-E/ELEC-1216.

RIS 2004-03 was a reference for the procedures used in the safe shutdown re-validation, and is also

referenced in FIR-NGGC-0101.

Comments I

Reference Document


Assessment (NSCA)

DocDetailSection 2.1, 9.1.3


Duke EnerqyRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant


Sections 2.1.2.5, _3.









NEI00-1Ref NEI 00-01 Guidance

Identify instrument tubing that may cause subsequent effects on instrument readings or signals as a3.2.1.7 [Instrument result of fire. Determine and consider the fire area location of the instrument tubing when evaluating

Tubing] the effects of fire damage to circuits and equipment in the fire area.


Applicable

Alignment Alignment B

Statement

FIR-NGGC-0101 provides direction for evaluating the fire effects on instrument tubing and theAligns potential impact on spurious operation. FSSPMD documents tubing routing to ensure the impact of

this issue is evaluated.

Comments

Reference Document DoclDtail


Assessment (NSCA)





NFPA 805Requirement







Capability Assessment Methodology Reviewsafety criteria shall be included. Availability and reliability of equipment selected shall be evaluated.

NEL00-01Ref

3.2.2 Methodologyfor EquipmentSelection

Applicability

Applicable

NEI 00-01 Guidance

Refer to Figure 3-3 for a flowchart illustrating the various steps involved in selecting safe shutdownequipment.

Use the following methodology to select the safe shutdown equipment for a post-fire safe shutdownanalysis:


Comments

Alignment Alignment BassStatement


Comments






Mark up and annotate a P&ID to highlight the specific flow paths for each system in support of each3.2.2.1 Identify' the shutdown path. Refer to Attachment 2 for an example of an annotated P&ID illustrating this concept.System Flow Pathfor Each ShutdownPath


Applicable

AlignmentStatement

Alignment Basis.

The safe shutdown flow paths at Robinson are depicted on the HBR2-11390 Series of drawings.Aligns



Comments



Assessment (NSCA)

HBR2-11390, Appendix R and Station Blackout Safe-Shutdown

Analysis Flowpath/Boundary Diagrams









NEI•00-01ReJ f NEI 00-01 GuidanceReview the applicable documentation (e.g. P&IDs, electrical drawings, instrument loop diagrams) to

3.2.2.2 Identify the assure that all equipment in each system's flow path has been identified. Assure that any equipment

Equipment in Each that could spuriously operate and adversely affect the desired system function(s) is also identified. If

Safe Shutdown additional systems are identified which are necessary for the operation of the safe shutdown system

System Flow Path under review, include these as systems required for safe shutdown. Designate these new systems

Including with the same safe shutdown path as the primary safe shutdown system under review (Refer to

Equipment That Figure 3-1).

May Spuriously

Operate and Affect

System Operation


Applicable


Statement

Using Operating Procedures (Normal, Emergency, and Abnormal), System Descriptions, Fire HazardAligns Analyses, Piping and Instrument Diagrams (P&IDs) and Single Line Diagrams for each system

comprising the safe shutdown paths, the mechanical or electrical equipment required for the

operation of the system and the equipment whose spurious operation could affect the performance of

the safe shutdown systems were identified.

Reference Document


Assessment (NSCA)

Doc Detali

Section 9.1.3


Duke EnergyRNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.



Section 2.1.2.5









NEI 0-01Ref NEI 00-01 GuidancePrepare a table listing the equipment identified for each system and the shutdown path that it

3.2.2.3 Develop a supports. Identify any valves or other equipment that could spuriously operate and impact the

List of Safe operation of that safe shutdown system. Assign the safe shutdown path for the affected system to

Shutdown this equipment. During the cable selection phase, identify additional equipment required to support

Equipment and the safe shutdown function of the path (e.g., electrical distribution system equipment). Include this

Assign the additional equipment in the safe shutdown equipment list. Attachment 3 to this document provides an

Corresponding example of a (SSEL). The SSEL identifies the list of equipment within the plant considered for safe

System and Safe shutdown and it documents various equipment-related attributes used in the analysis.

Shutdown Path(s)

Designation to

Each.


Applicable

Aignment BasisStatement

System and component identification is discussed in section 2.2.2, and section 2.2.3 refers to theAligns SSEL maintained in FSSPMD. Attachment 24 of is a printout of the SSEL. This includes valves and

pumps whose spurious operation may impact a safe shutdown system from performing its function.

Information in FSSPMD includes the component's power supply, fire zone location, normal and

required positions, required cables, and associated circuits.

The components and the safe shutdown function(s) they support are depicted in the safe shutdown

fault tree. Development of the fault tree is described in FIR-NGGC-0101, revision 2.

Comments

Reference Document


Doc DetailSection 9.2


Duke EneryFSSPMf, Fire Safe Shutdown Program Manager Database




Sections 2.2.2, 2.2.3, and Att. 24









NE IRef NEI 00-01 Guidance

Collect additional equipment-related information necessary for performing the post-fire safe shutdown3.2.2.4 Identify analysis for the equipment. In order to facilitate the analysis, tabulate this data for each piece of

Equipment equipment on the SSEL. Refer to Attachment 3 to this document for an example of a SSEL.

Information Examples of related equipment data should include the equipment type, equipment description, safe

Required for the shutdown system, safe shutdown path, drawing reference, fire area, fire zone, and room location of

Safe Shutdown equipment. Other information such as the following may be useful in performing the safe shutdown

Analysis analysis: normal position, hot shutdown position, cold shutdown position, failed air position, failed

electrical position, high/low pressure interface concern, and spurious operation concern.


Applicable


The information identified as needed for performing safe shutdown analysis on the componentsAligns identified on the SSEL is contained in the FSSPMD. This can be verified on a component basis

through reports that can be generated through the FSSPMD.

Information in FSSPMD includes the component's power supply, fire zone location, normal and

required positions, required cables, and associated circuits.

Comments


FSSPMD, Fire Safe Shutdown Program Mana er Database

RNP-E/ELEC-1216, The Fire Safe Shutdown Inalysis for H.B. Attachment 24













In the process of defining equipment and cables for safe shutdown, identify additional supporting3.2.2.5 Identify equipment such as electrical power and interlocked equipment. As an aid in assessing identified

Dependencies impacts to safe shutdown, consider modeling the dependency between equipment within each safe

Between shutdown path either in a relational database or in the form of a Safe Shutdown Logic Diagram

Equipment, (SSLD). Attachment 4 provides an example of a SSLD that may be developed to document these

Supporting relationships.

Equipment, Safe

Shutdown Systems

and Safe Shutdown

Paths.


Applicable

Alignment

Statement

Power supplies are identified and documented in the FSSPMD. Cables that are associated with aAligns component because of interlocks or permissive are documented with the component in the FSSPMD.

Comments



Assessment (NSCA)



NFPA 805 Section 2.4.2.2 Nuclear Safety Capability Circuit Analysis

NFPA 805Requirement

2.4.2.2.1 Cilcuits Required in Nuclear Safety Functions. Circuits required for th_ 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.

Page B-39HBRSEP LAR Rev 1


Capability Assessment Methodology ReviewThis 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 00-01Ref

3.3 Safe Shutdown

Cable Selection and

Location

NEI 00-01 Guidance

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.


Applicable


Statement

This is an introductory statement and provides no requirements. The sub-paragraphs with specific


Comments



NFPA 805

Requirement

2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety


tl e operation, or that result in the maloperation of the equipment ide tified in 2.4.2.1. This evaluation

sl}all consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and



This will ensure that a comprehensive population of circuitry is evaluated.





Capability Assessment Methodology Reviewtheir impact on the ability to achieve nuclear safety performance criteria.











To identify an impact to safe shutdown equipment based on cable routing, the equipment must have3.3.1 Criteria /cables that affect it identified. Carefully consider how cables are related to safe shutdown equipment

Assumptions 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:


Applicable

AlignmentAlnment Basis

Statement


Comments



NFPA 805

Requirement









encl'sure with circuits required to achieve nuclear safety performance criteria shall be evaluated for

theirlimpact on the ability to achieve nuclear safety performance criteria!






coordinated with the downstream protection device. Capability Assessment Methodology Review






NEI 00-01 Ref

3.3.1.1 [Cable

Selection]


NEI 00-01 Guidance

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.

Comments

Alignment

Statement

FIR-NGGC-01 01 Section 9.3.1 provides direction for assigning cables to components. This processAligns is documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.

Comments



Assessment (NSCA)


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.0




NFPA 805Requirement


functions shall be identified. This includes circuits that are recuired for operation, that could prevent








Capability Assessment Methodology Review2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common












NEI 0-1Rf NEI 00-01 Guidance

In cases where the failure (including spurious actuations) of a single cable could impact more thanone piece of safe shutdown equipment, include the cable with each piece of safe shutdown

Affecting Multiple equipment.

Components]


Applicable

AigmnAlignment BasisStatement

Circuit analysis is performed independently on individual components, so cables affecting more thanAligns one component will be identified with each applicable component.

Comments

Reference Document DocDDtail


Assessment (NSCA)






INFPA 895Requirement

2.4.2.2.1 Circuits Required in Nuclear Safety Funclions. Circuits required for the nuclear safety





















NEI 00-01Ref NEI 00-01 Guidance

Electrical devices such as relays, switches and signal resistor units are considered to be acceptable3.3.1.3 [Isolation isolation devices. In the case of instrument loops, review the isolation capabilities of the devices in

Devices] 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.


Applicable

Aliginment Alignment Basis

Statement

Isolation devices are defined in FIR-NGGC-0101, Section 3.Aligns

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 3, Item 43.

Assessment (NSCA)





NFPA 805

Requirement








Capability Assessment Methodology Reviewrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.This will ensure that a comprehensive population of circuitry is evaluated.













NEI 00-01 Ref NEI 00-01 GuidanceScreen out cables for circuits that do not impact the safe shutdown function of a component (i.e.,

3.3.1.4 [Identify annunciator circuits, space heater circuits and computer input circuits) unless some reliance on these"Not Required" circuits is necessary. However, they must be isolated from the component's control scheme in such aCables] way that a cable fault would not impact the performance of the circuit.


Applicable


In FSSPMD cables that are not required for safe shutdown have an "A" or an "NA" entered in theAligns FMEA section of the circuit information form in FSSPMD. The "A" indicates that the component

"achieves" its safe shutdown function even if that cable is damaged by fire. The "NA" indicates that

the cable is not part of a safe shutdown circuit.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2 and Attachment 1Assessment (NSCA)



NFPA 805 Section 2.4.2.2 Nuclear Safety Capa ility Circuit Analysis

NFPA 805

Requirement

2.4.2.2.1 Circuits Required in Nuclear Safety Functions. CircUits required for the nuclear safety




HBRSEP LAR Rev 1 Page B--45


Capability Assessment Methodology Reviewshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.









(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achievethe nuclear safety performance criteria and whose fire-induced failure could cause the loss of the




NEI 00-01 Ref

3.3.1.5

[Identification of

Power Supplies]

Applicability

Applicable

NEI 00-01 Guidance

For each circuit requiring power to perform its safe shutdown function, identify the cable supplyingpower to each safe shutdown and/or required interlock component. Initially, identify only the powercables 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 electricaldistribution system to capture the remaining equipment from the electrical power distribution systemnecessary to support delivery of power from either the offsite power source or the emergency dieselgenerators (i.e., onsite power source) to the safe shutdown equipment. Add this equipment to thesafe shutdown equipment list. Evaluate the power cables for this additional equipment for associatedcircuits concerns.

Comments

AlignmentStatement

Aligns

Alignment Bas*

The power cables for individual components are listed in the circuit analysis for that component ifpower is needed for the component to perform its safe shutdown function. Power supplies are linkedto their components in FSSPMD in the "Power Supplies, Related, Auxiliary, and Other ImportantCircuits" portion of the Circuit Information Form. A standard note "A" entered for a power supply inthis section indicates that the power supply is required for the component to perform its safeshutdown function. The power supply requirement is modeled in the fault tree.


FIR-NGGC-0101, Fire Protection Nuclear S fety Capability Section 9.3.2 and Attachment 1

Assessment (NSCA)







NFPA 805

Requirement




















NEI 00-01Ref

3.3.1.6 [ESFAS

Initiation]

Applicability

Applicable

NEI 00-01 Guidance

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.

Comments

AlignmentStaement

Aligns

Alignment Basis

Reliance on the automatic logic for safe shutdown systems is not credited at HBRSEP. Althoughautomatic ESFAS signals are not credited, they have been included in the SSA to assure that anactuation of the logic does not cause any adverse consequences.

The Fast Bus Transfer scheme and EDG Auto Sequencing fault trees have been modeled so thatthey maA be credited when available. I

Comments



FR-NGC 0101, Fire Protection Nuclear Safety Capability

Assessment (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant


Sections 9.2.3, 9.3.1, 9.3.2, and 9.3.7

Sections 2.1.3.2 and 2.2.4.4



NFPA 805

Requirement




















NEI 00-01 Ref

3.3.1.7 [Circuit

Coordination]

Applicability

Applicable

NEI 00-01 GuidanceCabling for the electrical distribution system is a concern for those breakers that feed associatedcircuits and are not fully coordinated with upstream breakers. With respect to electrical distributioncabling, two types of cable associations exist. For safe shutdown considerations, the direct powerfeed 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 loadcenter to an MCC supports the MCC. However, for cases where sufficient branch-circuit coordinationis not provided, the same cables discussed above would also support the power supply. Forexample, the power feed to the pump discussed above would support the bus from which it is fedbecause, for the case of a common power source analysis, the concern is the loss of the upstreampower source and not the connected load. Similarly, the cable feeding the MCC from the load centerwould also be necessary to support the load center.

Comments

AlignmentStatement Alignment Basis



The guidelines that will be used in the evaluation of the common power supplies are as follows (Ref.Aligns FIR-NGGC-0101):

- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for

electrical coordination will be developed.

- For each safe shutdown power supply, the existing short circuit calculations, load studies,

coordination calculations, protective device setting sheets, and time current curves as appropriate to

confirm proper coordination between upstream and downstream protective devices will be reviewed.

- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be

considered.

- For cases in which coordination between series protective devices cannot be demonstrated, a

common power supply associated circuit will be assumed to exist. These circuits will be dispositioned

by one of the following means:

1) Demonstrate coordination by refining the available short circuit current and/or trip device

characteristics.

2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,

equipment located in same fire area as power supply).

3) Identify readily achievable protective device setting changes (including changes in fuse size and/or

clearing characteristics) that will establish coordination.

4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 1OCFR50

Appendix R safe shutdown when protection devices do not provide the desired coordination.

5) Existing short circuit and coordination calculations will be updated as necessary to fully document

where coordination is credited for 10CFR50 Appendix R safe shutdown.

The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault

on the same breaker's load cable would prevent the breaker from tripping on over-current and could

result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV

breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV

power supply circuit analysis as Associated Circuits and Cables.

Comments



Assessment (NSCA)

FPP-RNP-200, 10CFR50, Appendix R, Section Ill.G, Associated

Circuits Analysis

RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common

Power Supply Analysis

RNP-E-8.053, Non-Safety Overcurrent Protection Coordination

RNP-E-9.021, 10CFR50 Appendix R Fuse Analysis for DS Bus

RNP-E/ELEY-1216, The Fire Safe Shutdown Analysis for H.B. Secti n 3.1.2.1

Robinson N ýclear Plant





NFPA 805

Requirement




















3.3.2 Associated

Circuit Cables

Applicability

Applicable

NEI 00-01 Guidance

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 tohot shorts, open circuits, or shorts to ground, of redundant trains of systems necessary to achievehot shutdown. The three types of associated circuits were identified in Reference 6.1.5 and furtherclarified in a NRC memorandum dated March 22, 1982 from R. Mattson to D. Eisenhut, Reference6.1.6. They are as follows:- Spurious actuations- Common power source- Common enclosure.

Comments



Comments

Ta le B-2 Nuclear Safety Capability Assessment Methodology Revie


HBRSEP LAR Rev I Page B-50

Duke Energy

NFPA 805

Requirement












(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated 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 therequired components shall be identified. The concern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.

3.3.2 [A] Associated

Circuit Cables -

Cables Whose

Failure May Cause

Spurious Actuations

NEI 00-01 Guidance

Safe shutdown system spurious actuation concerns can result from fire damage to a cable whosefailure could cause the spurious actuation/mal-operation of equipment whose operation could affectsafe shutdown. These cables are identified in Section 3.3.3 together with the remaining safeshutdown cables required to support control and operation of the equipment.


Applicable

AligmentAlignment BasisStatement

Cables that can cause an undesired spurious actuation are identified by an "S" in the FMEA code of

Aligns the circuit information form in FSSPMD. They are evaluated in the SSA in the same manner as"required" cables. RNP-E/ELEC-1216 evaluates throughout for spurious operation of valves, pumps,

and breakers.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.0, 9.1.3, 9.3.2, and Attachment 1

Asse sment (NSCA)

FSS MD, Fire Safe Shutdown Program Manager Database

RNP-EIELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 3.1.2.3


HBRSEP LAR Rev I Page B-51

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke EnergNTable 0- Nuclear Safety Capability Assessment Methodology Review


NFPA 805

Requirement











(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This








NEI 00-01 Ref

3.3.2 [B] Associated

Circuit Cables -

Common Power

Source Cables

Applicability

Applicable

NEI 00-01 Guidance

The concern for the common power source associated circuits is the loss of a safe shutdown powersource due to inadequate breaker/fuse coordination. In the case of a fire-induced cable failure on anon-safe shutdown load circuit supplied from the safe shutdown power source, a lack of coordinationbetween the upstream supply breaker/fuse feeding the safe shutdown power source and the loadbreaker/fuse supplying the non-safe shutdown faulted circuit can result in loss of the safe shutdownbus. This would result in the loss of power to the safe shutdown equipment supplied from that powersource 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 foranalyzing the impact of these cables on post-fire safe shutdown.

Comments

AlignmentStatement

Aligns

Alignment Basis

The guidelines that will be used in th• evaluation of the common power supplies are as follows (Ref.

FIR-NGGC-0101): I

- Using the single-line drawings, a list of the safe shutdown power supplies to be reviewed for

electrical coordination will be developed.

- For each safe shutdown power supply, the existing short circuit calculations, load studies,



Capability Assessment Methodology Reviewcoordination calculations, protective device setting sheets, and time current curves as appropriate to

confirm proper coordination between upstream and downstream protective devices will be reviewed.

- In reviewing coordination, electrical system line-ups credited in the safe shutdown analysis will be

considered.

- For cases in which coordination between series protective devices cannot be demonstrated, a

common power supply associated circuit will be assumed to exist. These circuits will be dispositioned

by one of the following means:

1) Demonstrate coordination by refining the available short circuit current and/or trip device

characteristics.

2) Demonstrate that the lack of coordination does not adversely affect safe shutdown (e.g.,

equipment located in same fire area as power supply).


clearing characteristics) that will establish coordination.

4) Incorporate the Associated Circuits and Cables into the safe shutdown analysis as 1OCFR50Appendix R safe shutdown when protection devices do not provide the desired coordination.

5) Existing short circuit and coordination calculations will be updated as necessary to fully document

where coordination is credited for 10CFR50 Appendix R safe shutdown.

The loss of control power or fire induced cable damage to a 4KV breaker's trip circuit prior to a fault

on the same breaker's load cable would prevent the breaker from tripping on over-current and could

result in loss of the 4KV power supply (i.e. switchgear). Therefore, all non-safe shutdown 4KV

breakers that are associated with a safe shutdown 4KV power supply will be included in the 4KV

power supply circuit analysis as Associated Circuits and Cables.

Comments



Assessment (NSCA)

RNP-E-8.005, 10CFR50 Appendix R Associated Circuit, Common

Power Supply Analysis






NFPA 805

Requirement


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 evlluation






Attachment B - NEt 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology Review

2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for









the immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.

NE[I•000IRef NEI 00-01 Guidance

The concern with common enclosure associated circuits is fire damage to a cable whose failure could3.3.2 [C] Associated propagate to other safe shutdown cables in the same enclosure either because the circuit is not

Circuit Cables - properly protected by an isolation device (breaker/fuse) such that a fire-induced fault could result in

Common Enclosure ignition along its length, or by the fire propagating along the cable and into an adjacent fire area. This

Cables 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.


Applicable

AlignmentAliganment Basis

Statement

The following guidelines were used in the evaluation of common enclosure associated circuits (Ref.Aligns FIR-NGGC-0101):

- Perform an evaluation of the common enclosure associated circuits by reviewing design and

installation criteria for cable and electrical penetrations. Confirm that cables are adequately protectedagainst short circuits and will not propagate a fire from one fire area to another. In evaluating

common power supply circuits the acceptance criteria shall not be limited to standard cable damage

temperatures, which are based on not degrading cable insulation (typically 2500 C for thermoset

cable). Rather, the criteria will be based on not exceeding temperatures at which self ignition or

damage to surrounding cables could occur.

- If a common enclosure associated circuit is determined to exist, the concern shall be resolved by

one of the following means:

1) Demonstrate by analysis that the cable does not pose a risk to cables within the common

enclosure under fault conditions (i.e., the cable exceeds its recommended tempera ure rise but does

not represent a hazard to surrounding cables),2) Demonstrate that the lack of fault protection does not adversely affect safe shutdown,


time-current characteristics) that will establish cable protection without affecting other performance

requirements, or



Capability Assessment Methodology Review4) Incorporate the cables of concern into the safe shutdown analysis as post-fire safe shutdown

cables for the affected power supply.

5) Existing short circuit and electrical protection calculations will be updated as necessary.

Comments



Assessment (NSCA)

FPP-RNP-200, 10CFR50, Appendix R, Section Ill.G, Associated Section 4.0

Circuits Analysis





NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safety

Requirement functions shall be identified. This includes circuits that are required for operation, that could prevent














the nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concern is that the effects of a fire can extend outside of



N-EL00-01IRef

3.3.3 Methodologyfor Cable Selectionand Location

NEI 00-01 Guidance

Refer to Figure 3-4 for a flowchart illustrating the various steps involved i selecting the cablesnecessary for performing a post-fire safe shutdown analysis.Use the following methodology to define the cables required for safe shutdown including cables thatmay cause associated circuits concerns for a post-fire safe shutdown analysis:



Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewApplicability Comments

Applicable

Alignment Basis

Statement


Comments





















NEI 00-01 Ref

3.3.3.1 IdentifyCircuits Requiredfor the Operation ofthe Safe Shutdown

Equipment

NEI 00-01 Guidance

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:

- Single-line electrical diagrams

- Elementary wiring diagrams

- Electrical connection diagrams

- Instrument loop diagrams.

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,



Capability Assessment Methodology Reviewinclude the additional upstream power sources up to either the offsite or the emergency power

source.


Applicable


FIR-NGGC-01 01 Section 9.3.2 provides direction for assigning cables to components. This processAligns is further documented in Section 3.0 of RNP-E/ELEC-1216 and in the FSSPMD.

Comments


FIR-NGGC-01.01, Fire Protection Nuclear Safety Capability Section 9.3.2Assessment (NSCA)





NFPA 805 2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyRequirement 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 evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired 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 commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir 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 ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achievethe nuclear safety performance criteria and whose fire-induced failure could cause the loss of therequired components shall be identified. The concerh is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrica/faults on inadequately protected cables or viainadequately sealed fire area boundaries.

NE[ 00-0 ef

HBRSEP LAR Rev 1

NEI 00-01 Guidance

In reviewing each control circuit, investigate interlocks that may lead to additional circuit schemes,

Page B-57

Duke Energy3.3•.3.2 Identify

Interlocked Circuits

and Cables Whose

Spurious Operation

or Mal-operation

Could Affect

Shutdown

ApplicabiletyApplicable


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 Figure 3-3) if they can impact the operation of the equipment under consideration.

Comments

Alignment

Statement

Aligns with intent

Alignment Basis

As an alternative to adding the interlocked equipment to the SSEL, it is acceptable to include the

cables that are required for the interlocking function (or that could cause the spurious actuation) with

the main component that was originally under consideration. Adding them to the components may

ease the development of a suitable mitigating strategy in areas where the interlocked cables may be

damaged by the fire. Interlocked circuits were either included in the analysis, or the interlocked

contact or relay was assumed to be in its worst-case position. Associated circuits identified for each

component are either included in the main circuit analysis, or are included by listing the applicable

circuit in the "Power Supplies, Related, Auxiliary, and Other Important Circuits" on the Circuit

Information Form.

Comments



Assessment (NSCA)




NFPA 805

Requirement








2.4.2.2.2 Other Required Circuits. Other cýircuits that share common power supply and/or common

enclosure with circuits required to achiev• nuclear safety performance criteria shall be evaluated for







Capability Assessment Methodology Reviewcoordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve


NEI 00-01 Ref

3.3.3.3 Assign

Cables to the Safe

Shutdown

Equipment

NEI 00-01 Guidance

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.


Applicable

AinntAlignment Basis

Statement

The circuit analysis results for each electrically operated safe shutdown component are contained inAligns the FSSPMD. FSSPMD contains various forms and reports for presenting SSD cables and

associated circuits. Refer to Progress Energy procedure FIR-NGGC-0101 for a description of the

circuit analysis nomenclature used in the FSSPMD.

Comments

Reference Document Doc DetaflFIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.2

Assessment (NSCA)




NFPA 805 Section 2.4.2.3 Nuclear Safety Equipment and Cable Location.

NFPA 805 Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be

Requirement identified.

NEIBR -L0 Revf

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Identify the routing for each cable including all raceway and cable endpoints. Typically, this

Page B-59

Duke Energ3.3.3.4 Iycentify

Routing of Cables

Applicability

Applicable


information is obtained from joining the list of safe shutdown cables with an existing cable and

raceway database

Comments

igmnt AlignmentB

Statement

Equipment location and Cable raceway routing information (i.e. cable raceway to fire area/zone

Aligns correlation) was migrated in 2004 from the safe shutdown analysis of record at the time (FPP-RNP-

150, Revision 7A) to the FSSPMD. Additional cables were added to FSSPMD based on revised

component selection.

Comments



Assessment (NSCA)








NEI 00-01 Ref NEI 00-01 GuidanceIdentify the fire area location of each raceway and cable endpoint identified in the previous step and

3.3.3.5 Identify join this information with the cable routing data. In addition, identify the location of field-routed cable

Location of by fire area. This produces a database containing all of the cables requiring fire area analysis, their

Raceway and locations by fire area, and their raceway.

Cables by Fire Area


Applicable


Cable to raceway infbrmation is contained in the Cable Information Forms of the FSSPID. RacewayAligns and endpoint locations for all required cables are also contained in FSSPMD.

Comments




Assessment (NSCA)





DoctDeti9l

Section 9.3.9

Section 4.0


NFPA 805 Section 2.4.2.4 Fire Area Assessment.

NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with the

Requirement 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).

0-0 NEI 00-01 Guidance

By determining the location of each component and cable by fire area and using the cable to3.4 Fire Area equipment relationships described above, the affected safe shutdown equipment in each fire area

Assessment and can be determined. Using the list of affected equipment in each fire area, the impacts to safe

Compliance shutdown systems, paths and functions can be determined. Based on an assessment of the number

Assessment 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.

Applocabfflet Comments

Applicable

Alignment Basis

Statement


Comments



NFPA 805

Requirement

HBRSEP LAR Rev 1

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


Page B-61




NE[D00-01Ref NEI 00-01 Guidance

The following criteria and assumptions apply when performing fire area compliance assessment to3.4.1 Criteria / mitigate the consequences of the circuit failures identified in the previous sections for the requiredAssumptions safe shutdown path in each fire area.


Applicable

Almommit Alignment BasisStatement


Comments



NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with theRequirement 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 Chapter4 for methods of achieving these performance criteria (performance-based or deterministic).

NEI.0001Ref NEI 00-01 Guidance

Assume only one fire in any single fire area at a time.3.4.1.1 [Number ofPostulated Fires]


Applicable

AligenmentStatement

RNP-E/ELEC-1216 postulates only one fire occurring at a time.Aligns

Comments

Reference Document I


Assessment (NSCA)



Doc Detail

Section 9.4.1

Section 1.4









NEI.00-01Ref NEI 00-01 Guidance

Assume that the fire may affect all unprotected cables and equipment within the fire area. This3.4.1.2 [Damage to assumes that neither the fire size nor the fire intensity is known. This is conservative and bounds the

Unprotected exposure fire that is required by the regulation.

Equipment and

Cables]


Applicable


RNP-E/ELEC-1216 postulates all electrical equipment and cables in a given fire area are damagedAligns and unavailable unless NRC exemption or appropriate evaluation (GL-86-10) has been completed.

Comments



Assessment (NSCA)







act vities on the ability to achieve the nuclear safety performance crite ia of Section 1.5. [See Chapter4 fr methods of achieving these performance criteria (performance-b sed or deterministic).

NEI 00-01Ref

3.4.1.3 [Assess

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Address all cable and equipment impacts affecting the required safe shutdown path in the fire area.

Page B-63

Duke Energy

Impacts to Required

Components]



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.

Comments

Aignment Alignment Basis

Statement

All potential impacts of the fire are identified in the fault tree. Potential damage to equipmentAligns required to show success in each area is addressed with an appropriate compliance strategy. The

results are documented in FSSPMD and in RNP-E/ELEC-1216.

Comments



Assessment (NSCA)









NEIO0-1Ref NEI 00-01 Guidance

Use manual actions where appropriate to achieve and maintain post-fire safe shutdown conditions in3.4.1.4 [Manual accordance with NRC requirements.

Actions]


Applicable

Alignment

Statement

Aligns

AlignmentBasis

Manual actions credited in the shutdown analysis are sumrrmarized on a fire area basis in Attachment

26 of RNP-E/ELEC-1216. RNP-E-8.050 documents the feasibility of the manual actions. The

current regulatory guidance, as reflected in FAQs 06-0012 and 07-0030 was used as the basis for

determining the acceptability of the manual actions.

Comments




Reference Document Doc DetailFIR-NGGC-0101, Fire Protection Nuclear Safety Capability Attachment 2Assessment (NSCA)

RNP-E-8.050, Appendix R Transient Analysis and TimelineEvaluation for H.B. Robinson - Unit No. 2

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 5.2 and Alt. 26Robinson Nuclear Plant



NFPA 805 Fire Area Assessment. An engineering analysis shall be performed in accordance with theRequirement requirements of Section 2.3 for each fire area to determine the effects of fire or fire suppression


NEI 00-01 Ref NEI 00-01 GuidanceWhere appropriate to achieve and maintain cold shutdown within 72 hours, use repairs to equipment

3.4.1.5 [Repairs] required in support of post fire shutdown.


Applicable


Repairs are identified where necessary for cold shutdown equipment. A list of credited repairAligns procedures can be found in RNP-E/ELEC-1216.

No repairs are required to achieve safe and stable conditions.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.15 and 9.4.2Assessment (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Section 1.7.2.2Robinson Nuclear Plant

Table •-12 Nuclear Safety Capability Assessment Methodology Review



Duke Energy

NFPA 805Requirement






EIEL00-01Ref

3.4.1.6 [Assess

Compliance with

Deterministic

Criteria]

Applicability

Applicable

NEI 00-01 Guidance

Appendix R compliance requires that one train of systems necessary to achieve and maintain hotshutdown conditions from either the control room or emergency control station(s) is free of firedamage (lll.G.l.a). When cables or equipment, including associated circuits, are within the same firearea outside primary containment and separation does not already exist, provide one of the followingmeans of separation for the required safe shutdown path(s):- Separation of cables and equipment and associated non-safety circuits of redundant trains withinthe same fire area by a fire barrier having a 3-hour rating (lll.G.2.a)- Separation of cables and equipment and associated non-safety circuits of redundant trains withinthe same fire area by a horizontal distance of more than 20 feet with no intervening combustibles orfire hazards. In addition, fire detectors and an automatic fire suppression system shall be installed inthe fire area (llI.G.2.b).- Enclosure of cable and equipment and associated non-safety circuits of one redundant train withina fire area in a fire barrier having a one-hour rating. In addition, fire detectors and an automatic firesuppression system shall be installed in the fire area (lll.G.2.c).For fire areas inside noninerted containments, the following additional options are also available:- Separation of cables and equipment and associated non-safety circuits of redundant trains by ahorizontal distance of more than 20 feet with no intervening combustibles or fire hazards (llI.G.2.d);- Installation of fire detectors and an automatic fire suppression system in the fire area (llI.G.2.e); or- Separation of cables and equipment and associated non-safety circuits of redundant trains by anoncombustible radiant energy shield (llI.G.2.f).Use exemptions, deviations and licensing change processes to satisfy the requirements mentionedabove and to demonstrate equivalency depending upon the plant's license requirements.

Comments

AlignmentStatement

This section of NEI 00-01 repeats the requirements of Appendix R III.G.2. RNP-E/ELEC-1216Aligns documents how each fire area has adequate systems to comply with these requirements and the

requirements of NFPA 805 Sections 4.2.3 and 4.2.4 for the post-transition configuration.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4.2Assessment (NSCA)










NEI00-01 Ref NEI 00-01 GuidanceConsider selecting other equipment that can perform the same safe shutdown function as the

3.4.1.7 [Consider impacted equipment. In addressing this situation, each equipment impact, including spurious

Additional operations, is to be addressed in accordance with regulatory requirements and the NPP's current

Equipment] licensing basis.

App1icability Comments

Applicable


Statement

This consideration is not clearly stated but is inherent in performing a safe shutdown analysis. RNP-Aligns E/ELEC-1216 only documents the systems and components that were actually selected and not

those that were considered but not necessary.

Any plant system that supports meeting the safe shutdown performance goals may be considered for

inclusion in the SSEL. However, the intent is to minimize the systems and components identified in

the SSEL for configuration control purposes. If the system or component cannot directly assist in

demonstrating compliance with the deterministic requirements of NFPA 805, its inclusion in the SSEL

may not be warranted.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.1 and 9.4

Assessment (NSCA)





NFPA 805 Fire Area Assessment. An ek.gineering analysis shall be performed in accordance with the





Duke Ene81RefNEI 00-1Re

3.4.1.8 [ConsiderInstrument TubingEffects]

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyNEI 00-01 Guidance Capability Assessment Methodology Review

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.


Applicable

AlignmentSatemnt Alignment BasisStatement

RNP-E/ELEC-1216 documents the consideration of fire effects on instrument tubing for HBRSEP.Aligns

Comments

Reference Document DoDetael


Assessment (NSCA)

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Sections 2.1.2.7, 2.2.2.12, Att. 22, and Att. 25








NEI 00-01 Re NEI 00-01 GuidanceRefer to Figure 3-5 for a flowchart illustrating the various steps involved in performing a fire area

3.4.2 Methodology assessment.

for Fire Area Use the following methodology to assess the impact to safe shutdown and demonstrate Appendix R

Assessment compliance:



Applicable

augmo n

Staement

N/A

Alignment Basis


requirements are addressed separately as required.



Comments







NEIA0-1ef NEI 00-01 Guidance

Identify the safe shutdown cables, equipment and systems located in each fire area that may be3.4.2.1 Identify the potentially damaged by the fire. Provide this information in a report format. The report may be sorted

Affected Equipment by fire area and by system in order to understand the impact to each safe shutdown path within each

by Fire Area fire area (see Attachment 5 for an example of an Affected Equipment Report).


Applicable


Statement

RNP-E/ELEC-1216 lists the components potentially affected in each fire area. These reports areAligns available in the FSSPMD.

Comments

Reference Document


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Attachment 22






activities on the ability to achieve the nuclear safety performance criteria o0 Section 1.5. [See Chapter

4 for m. thods of achieving these performance criteria (performance-based or deterministic).

NEI 00-01 Ref

3.4.2.2 Determine

HBRSEP LAR Rev 1

NEI 00-01 Guidance

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.

Page B-69

Duke Energythe Shut-down Paths

Least Impacted By

a Fire in Each 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 mal-

operation 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.

A~ppicabilt CommentApplicable

AlignmentStatemenlt

Algnment ais

AlignsRNP-E/ELEC-1216 identifies both the equipment that is selected for a given fire area and the

equipment that is not selected. This shows that the division selected for a given safe shutdown

system is the train that was generally that least affected by the fire.

The use of a fault tree in the analysis helps to ensure that components that may have support

systems affected, such as cooling or power supplies, are not credited without taking these failures

into account.

Reference Document Detal

FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4Assessment (NSCA)


RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B. Att. 22Robinson Nuclear Plant

Table B-2 Nuclear SaIfety Capability Assessment Methodology Review

NFPA 805 Section

HBRSEP LAR Rev 1

2.4.2.4 Fire Area Assessment.

Page B-70


NFPA 805Requirement





NEI NEI 00-01 Guidance

Using the circuit analysis and evaluation criteria contained in Section 3.5 of this document, determine3.4.2.3 Determine the equipment that can impact safe shutdown and that can potentially be impacted by a fire in the fire

Safe Shutdown area, and what those possible impacts are.

Equipment Impacts


Applicable

Alignment BStatement

RNP-E/ELEC-1216 identifies the equipment used for safe shutdown and what the potential impact ofAligns the fire on the safe shutdown equipment. This information is also contained in FSSPMD.

Comments



Assessment (NSCA)









NE[lI00-01Ref

3.4.2.4 Develop a

Compliance

Strategy or

Disposition to

Mitigate the Effects

Due to Fire

Damage to Each

HBRSEP LAR Rev 1

NEI 00-01 Guidance

The available deterministic methods for mitigating the effects of circuit failures are summarized asloollows (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 demonstratethat there are no intervening combustibles within the 20 foot separation distance.- Reroute or relocate the circuit/equipment, or perform other modifications to resolve vulnerability.

Page B-71

Duke Energy

Required

Component or

Cable


Attachment B - NE 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology Review

- 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 Ill.G.2.d, e and f.

Comments

AlignmentStatment

Aligns

Alignment Basis

RNP-E/ELEC-1216 verifies that appropriate separation is used for redundant cables. This can take

the form of 3-hour fire barriers, 1-hour fire barriers with suppression and detection, or 20 feet of

separation with suppression and detection. In some cases, exemptions have been requested from

and granted by the NRC for configurations that did not meet these requirements. Also some fire

protection evaluations have determined that the protection in place provides adequate separation for

the hazards of the area.



Assessment (NSCA)









NE[OI 0-1Ref

3.4.2.5 Documentthe ComplianctStrategy or IDispositionDetermined toMitigate the EffectsDue to Fire

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Assign compliance strategy statements or codes to components or cables to identify the justificationor mitigating actions proposed for achieving safe shutdown. The justification should address thecumulative effect of the actions relied upon by the license to mitigate a fire in the area. Provideeach piece of safe shutdown equipment, equipment not in the path whose spurious operation or mal-operation could affect safe shutdown, and/or cable for the required safe shutdown path with a specificcompliance strategy or disposition. Refer to Attachment 6 for an example of a Fire Area AssessmentReport documenting each cable disposition.

Page B-72

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewDamage to EachRequiredComponent orCable


Applicable

AlignmentStatmentAlignment Basis

Statement

The fire area by fire area separation reports contained in Attachment 22 of RNP-E/ELEC-1216Aligns identify all the analyzed circuits and components and the credited compliance strategies.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.4Assessment (NSCA)






the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired 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 commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir 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 ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properlycoordinated 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 wl ose fire-induced failure could cause the loss of therequired components shall be identified. The cZoncern is that the effects of a fire can extend outside ofthe immediate fire area due to fire-induced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Page B-73

Duke Energy

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 Il.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.


Applicable

AlignmtAlignment BasisStatement


N/A requirements are addressed separately as required by NFPA 805 Sections 4.2.3 and 4.2.4 for the

post-transition configuration.

Comments




















the immediate fire area due to fire-i duced electrical faults on inadequately protected cables or viainadequately sealed fire area boundaries.

NE-•00-01RQef

3.5.1 Criteria /

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Apply the following criteria/assumptions when performing fire-induced circuit failure evaluations.

Page B-74

Duke EnergyAssumptions

Applicability

Applicable


Comments

Ali nment Alignment Basis

Statement


Comments





the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, and







(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This







3.5.1.1 [Circuit

Failure Types andImpact]

NEI 00-01 Guidance

Consider the following circuit failure types on each conductor of each unprotected safe shutdowncable to determine the potential impact of a fire on the safe shutdown equipment associated with that

conductor.

- A hot short may result f[om a fire-induced insulation breakdown between conductors of th• samecable, 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 circuitcontinuity. 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]


Duke Energy



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.

Comments

AlignmentStatement

Aligns

Alignment Basis

The safe shutdown circuit analysis shall be reviewed and updated as necessary for credible circuit

failures as a deterministic analysis utilizing the Current Design Method (CDM). These failures include:

- Multiple shorts to ground or grounded conductor.

- Multiple open circuits.

- One hot short per affected component or multiple hot shorts for high/low pressure interface

components.

- Cable-to-cable shorts are postulated to occur.



Assessment (NSCA)





NFPA 805

Requirement






required to actIieve the nuclear safety performance criteria, including spurious op ration and signals.







Capability Assessment Methodology Review(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the [oss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This









Assume that circuit contacts are positioned (i.e., open or closed) consistent with the normal3.5.1.2 [Circuit mode/position of the safe shutdown equipment as shown on the schematic drawings. The analyst

Contacts and must consider the position of the safe shutdown equipment for each specific shutdown scenario when

Operational Modes] determining the impact that fire damage to a particular circuit may have on the operation of the safe

shutdown equipment.


Applicable


Components are assumed to be in their normal position at the time of the fire. This includes electricalAligns contacts and switches.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 9.3.2 and 9.4.1

Assessment (NSCA)



NFPA 805Requirement






This will ensure that a comprehensive population of circuitry is evaluated./







Capability Assessment Methodology Reviewa power supply required to achieve the nuclear safety performance criteria shall be identified. This








NEI000Ref NEI 00-01 Guidance

Assume that circuit failure types resulting in spurious operations exist until action has been taken to3.5.1.3 [Duration of isolate the given circuit from the fire area, or other actions have been taken to negate the effects of

Circuit Failures] 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.

Appolicability Comments

Applicable


Statement

"Hot Short" duration is considered to exist until action has been taken to isolate the given circuit fromAligns the fire area, or other actions as appropriate have been taken to negate the effects of the spurious

actuation. HBRSEP does not postulate that the fire will eventually clear the "Hot Short."

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3.6

Assessment (NSCA)



NFPA 805

Requirement














Capability Assessment Methodoloqy Reviewa power supply required to achieve the nuclear safety performance criteria shall be identified. This








N.EI0-01 Ref NEI 00-01 Guidance

When both trains are in the same fire area outside of primary containment, all cables that do not3.5.1.4 [Cable meet the separation requirements of Section II.G.2 are assumed to fail in their worst caseFailure configuration.Configurations]


Applicable

Alignment Alignment Basisstatement

The following damage is assumed to occur due to the postulated fire:Aligns a. Fire damage occurs throughout the fire area under consideration.

b. Fire damage results in an unusable cable that cannot be considered functional with regard toensuring proper circuit operation.

Electrical equipment located in a fire area is assumed to fail as a result of the postulated fire in thefire area, and is considered unavailable to ensure completion of safe shutdown functions unless it

meets the separation criteria of 10 CFR 50 Appendix R or is shown to be acceptable as-is based onan approved exemption. This electrical equipment includes motors, instruments, UiP converters,controllers, switches, MCC's, switchgear, transformers, generators, batteries, panel boards, etc.

Comments

1Reference Document Doc Detail


Assessment (NSCA)



NFPA 89L5

Require, ent

2.4.2.2.1 Circuits Required in Nuclear Safety Funclions. Circuits required for the nuclear safety





















NEI 00-01 Ref

3.5.1.5 [A, Circuit

Failure Risk

Assessment

Guidance]

NEI 00-01 Guidance

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.


Not Applicable

Alignmet Alignment Basis

Statement



Comments



NFPA 805

Requirement

2.4.2.2.1 Circuits Required in Nuclear Safety Functions. Circuits required for the nuclear safetyfunctions shall be identified. This includes circuits that are required for operation, that could preventthe operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that ;re required to support the proper operation of componentsrequired to achieve the nuclear safety p rformance 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 commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.



Capability Assessment Methodology Review(a) Common Power Supply Circuits. Those circuits whose fire-induced failure could cause the loss ofa power supply required to achieve the nuclear safety performance criteria shall be identified. This








NEI 00-01 Ref

3.5.1.5 [B, Cable

Failure Modes]

NEI 00-01 Guidance

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 likelythan 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 aninspection 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 instru ent 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

necessary to achieve and maintain hot shutdown, then the instrument circuit may be considered an

associated circuit and handled accordingly.


Attachment B - NEI 04-02 Table B-2 Nuclear SafetyDuke Energy Capability Assessment Methodology ReviewApplicability Comments

Not Applicable

AlinmentAlignment BStatement

Section 9.3.3 of FIR-NGGC-0101 describes the specific cable failure modes to be considered inAligns conducting a circuit analysis. Section 3.0 of FIR-NGGC-0101 contains general definitions for the

cable failure modes. Configurations considered included the following, and were applied to all safe

shutdown cables.

a) For any multiconductor cable (including thermoset, thermoplastic, and armored), any and all

potential spurious actuations that may result, including possible combinations of conductors within the

cable, may be postulated to occur concurrently regardless of the number.

b) Inter-cable shorting of thermoset cables, or thermoset and thermoplastic cables, are considered

to be credible events.

c) Compatible polarity hot shorts for DC circuits were considered to the degree specified in the cases

below:

- Case 1 - Intra-Cable Shorts within a Single Cable

For this case, a single cable must contain both a source and target conductor for both polarities. It is

postulated that intra-cable shorts within the cable will result in compatible polarity connections for

both polarities (e.g., a plus-to-plus and a minus-to-minus connection for a DC control circuit). Given

the relatively high probability of intra-cable conductor-to-conductor shorting, this failure mode was

considered.

- Case 2 - Intra-Cable Shorts on Separate Cables

For this case, two independent but coincident hot shorts of the proper polarity (without grounding) in

separate cables must occur. Given the relative high probability of intra-cable conductor-to-conductor

shorting, this failure mode was considered.

- Case 3 - Inter-Cable Shorts on Separate Cables

For this case two independent but coincident hot shorts of the proper polarity (without grounding)must occur. This case differs from Case 1 and 2 in that one or both of the hot shorts must involve

inter-cable shorting. Given the low likelihood of coincident proper polarity shorts combined with the

low likelihood of inter-cable hot shorting, this failure mode was only considered for components

identified as "high-low pressure interface" or Fire PRA "high consequence equipment."

In the plant's review of multiple spurious actuations, the following were considered.

a. Multiple spurious operations resulting from a fire-induced circuit failure affecting a single conductor.

b. Multiple fire-induced circuit failures affecting multiple conductors within the same multi-conductor

cable with the potential to cause a spurious operation of a component were assumed to exist

concurrently.

c. Multiple fire-induced circuit failure affecting separate conductors in separate cables with the

potential to cause a spurious operation of a component must be assumed to exist concurrently when

the effect of the fire-induced circuit is sealed-in or latched. There was no specific limit to the number

of cables that were considered to be damaged.

Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.0, 9.3.2, 9.3.3, 9.3.10, 9.4.3, 9.4.6

Assessment (NSCA)











NE-IL•DD•Re• NEI 00-01 Guidance

Determination of the potential consequence of the damaged associated circuits is based on the3.5.1.5 [C, examination of specific NPP piping and instrumentation diagrams (P&IDs) and review of components

Likelihood of that could prevent operation or cause maloperation such as flow diversions, loss of coolant, or other

Undesired scenarios that could significantly impair the NPP's ability to achieve and maintain hot shutdown.

Consequences] 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.


Applicable


RNP-E/ELEC-1216 limits the evaluation of multiple spurious operations, implementing the designAligns with intent strategy of any and all potential spurious operation, on a one at a time basis. As part of the manual

action feasibility study, two concurrent spurious operations were evaluated.

Multiple spurious operations (MSO) were considered for a variety of scenarios by the MSO Expert

panel. Components were identified for consideration and possible inclusion in the Safe Shutdown

Analysis and the Fire PRA. Any MSOs that were determined to be risk-significant by the PRA were

analyzed accordingly.

Comments


RNP-E-8.050, Appendix R Transient Analysis and Timeline

Evaluation for H.B. Robinson - Unit No. 2





Duke EnergyNFPA 805 Section


2.4.2.2 Nuclear Safety Capability Circuit Analysis

NFPA 805

Requirement












a power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation could occur if the upstream protection device (i.e., breaker or fuse) is not properly






NE[.0_0.IRef

3.5.2 Types ofCircuit Failures

Applicability

Applicable

NEI 00-01 Guidance

Appendix R requires that nuclear power plants must be designed to prevent exposure fires fromdefeating the ability to achieve and maintain post-fire safe shutdown. Fire damage to circuits thatprovide control and power to equipment on the required safe shutdown path and any other equipment

whose spurious operation/mal-operation could affect shutdown in each fire area must be evaluatedfor the effects of a fire in that fire area. Only one fire at a time is assumed to occur. The extent of firedamage is assumed to be limited by the boundaries of the fire area. Given this set of conditions, itmust be assured that one redundant train of equipment capable of achieving hot shutdown is free offire 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 bedesigned for the fire-induced effects of a hot short, short-to-ground, and open circuit. With respect tothe 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.

Comments

Aliianment

Statement


Comments





NFPA 805

Requirement




















NEI 00-01 Ref

3.5.2.1 Circuit

Failures Due to an

Open Circuit

NEI 00-01 Guidance

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

0oher device. This loss of power may change the state of the equipn ent. Evaluate this to determine

iftequipment fails safe.

Open circuit on a high voltage (e.g., 4.16 kV) ammeter current transformer (CT) circuit may result in

secondary damage.



Capability Assessment Methodology ReviewFigure 3.5.2-1 shows an open circuit on a grounded control circuit.

[Refer to hard copy of NEI 00-01 for Figure 3.5.2-1]

Open circuit No. 1:

An open circuit at location No. 1 will prevent operation of the subject equipment.

Open circuit No. 2:

An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not

impact the ability to close/stop the equipment.

CommentsApplicability

Applicable

AlignmenStatement

Alignment Basis

AlignsCircuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method

(RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress

-Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes

multiple open circuits.

An evaluation considering the potential for secondary fires resulting from an open circuit on all CT

secondary circuits at RNP has been performed and is documented in EC 93120. This evaluation

concludes that there is little or no potential for adverse impacts to Safe Shutdown equipment resulting

from this postulated failure mode.

Comments


EC 93120, Evaluation of Possible Secondary Fire Caused by Open

Circuited CT at RNP

FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Section 9.3.3, 9.3.4Assessment (NSCA)




NFPA 805

Requirement



the operation, or that result in the maloperation of the equ pment identified in 2.4.2.1. This evaluation

shall consider fire-induced failure modes such as hot shorts (external and internal), open .circuits, and






Capability Assessment Methodology Review2.4.2.2.2 Other Required Circuits. Other circuits that share common power supply and/or common












NEI ~00-1Ref

3.5.2.2 CircuitFailures Due to aShort-to-Ground [A,General]


NEI 00-01 Guidance

This section provides guidance for addressing the effects of a short-to-ground on circuits for safeshutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation systemresulting in the potential on the conductor being applied to ground potential. A short-to-ground cancause 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 whereproper coordination does not exist.

Consider the following consequences in the post-fire safe shutdown analysis when determining theeffects 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 mayresult in loss of power to an interlocked relay or other device that may cause one or more spuriousoperations.

Comments

AlignmenttAliganment Basis

Statement


Comments

Table 8-2 Nuclear Safety Capability Assessment Methodology Review


NFPA 805

Requirement

HBRSEP LAR Rev 1




Page B-87



shorts to ground, to identify circuits that are required to support the proper operation of componentsrequired to achieve the nuclear safety performance criteria, including spurious operation and signals.



enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.








This section provides guidance for addressing the effects of a short-to-ground on circuits for safe3.5.2.2 Circuit shutdown equipment. A short-to-ground is a fire-induced breakdown of a cable insulation systemFailures Due to a resulting in the potential on the conductor being applied to ground potential. A short-to-ground canShort-to-Ground [B, cause a loss of power to or control of required safe shutdown equipment. In addition, a short-to-

Grounded Circuits] ground may affect other equipment in the electrical power distribution system in the cases where

proper coordination does not exist.

Short-to-Ground on Grounded Circuits

Typically, in the case of a grounded circuit, a short-to-ground on any part of the circuit would present

a concern for tripping the circuit isolation device thereby causing a loss of control power.

Figure 3.5.2-2 illustrates how a short-to-ground fault may impact a grounded circuit.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-21

Short-to-ground No. 1:

A short-to-ground at location No. 1 will result in the control power fuse blowing and a loss of power to

the control circuit. This will result an inability to operate the equipment using the control switch.

Depending on the coordination characteristics between the protective device on this circuit and

upstream circuits, the power supply to other circuits could be affected.


A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch

is closed. Should this occur, the effect would be identical to that for the short-to-ground at location

No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop

control switch, the equipment will still be able to be opened/started.

Aplical_•bilat Comments

Applicable

Alignment


Duke Energy

Statement

Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewAlignment Basis

Circuits are evaluated using both the Current Design /Method (CDM) and Revised Design Method

(RDM). Fire induced circuit failures (CDM and RDM) to be considered are described in Progress

Energy Procedure FIR-NGGC-0101. Fire induced circuit failure analysis utilizing CDM includes

multiple shorts to ground.

Aligns









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 fortheir 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 ofa power supply required to achieve the nuclear safety performance criteria shall be identified. Thissituation 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



3.5.2.2 Circuit

Failures Due to a

Short-to-Ground [C,

Ungrounded

Circuits]

HBRSEP LAR Rev 1

NEI 00-01 Guidance

Short-to-Ground on Ungrounded Circuits

In the case of an ungrounded circuit, postulating only a single short-to-ground on any part of the

circuit may not result in tripping the circuit isolation device. Another short-to-ground on the circuit or

another circuit from the same source would need to exist to cause a loss of control power to the

circuit.

Page B-89


Capability Assessment Methodology ReviewFigure 3.5.2-3 illustrates how a short to ground fault may impact an ungrounded circuit.

[Refer to hard copy of NEI 00-01 Rev. 1 for Figure 3.5.2-3]

Short-to-ground No. 1: A short-to-ground at location No. 1 will result in the control power fuse

blowing and a loss of power to the control circuit if short-to-ground No. 3 also exists either within the

same circuit or on any other circuit fed from the same power source. This will result in an inability to

operate the equipment using the control switch. Depending on the coordination characteristics

between the protective device on this circuit and upstream circuits, the power supply to other circuits

could be affected.


A short-to-ground at location No. 2 will have no effect on the circuit until the close/stop control switch

is closed. Should this occur, the effect would be identical to that for the short-to-ground at location

No. 1 described above. Should the open/start control switch be closed prior to closing the close/stop

control switch, the equipment will still be able to be opened/started.

CommentsApplicability

Applicable

AlmawmntStatement

Alignment Basis

AlignsA single ground fault on an ungrounded AC or DC control circuit has no immediate functional affect.

Thus, ungrounded systems are more resilient to functional failures. Nonetheless, multiple ground

faults are credible and must be considered. For ease of analysis, an existing - but unspecified -

ground fault from the same power source will be assumed when analyzing ungrounded systems.

Furthermore, multiple shorts-to-ground are to be evaluated for their impact on ungrounded circuits.

As noted in FIR-NGGC-0101, it is likely that over the course of a fire at least one conductor from

each polarity of a circuit (positive and negative polarity) will eventually become grounded. Thus, the

circuit analysis should not try to take credit for a circuit remaining functional simply because two

conductors must short to ground to render the circuit inoperable (i.e., blow the fuse or trip the circuit

breaker).

Comments

Reference Document Doc Qetail




NFPA 805

Requirement

HBRSEP LAR Rev 1




Page B-90

















NEI 0-01Ref NEI 00-01 Guidance

This section provides guidance for analyzing the effects of a hot short on circuits for required safe3.5.2.3 Circuit shutdown equipment. A hot short is defined as a fire-induced insulation breakdown between

Failures Due to a conductors of the same cable, a different cable or some other external source resulting in an

Hot Short [A, undesired impressed voltage on a specific conductor. The potential effect of the undesired impressed

General] 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.


Applicable

AAgenment

Statement

The Current Design Method is the safe shutdown circuit analysis method used for applying failures toAligns circuits. One key atribute of this method is that a hot short is applied independent of the cable

configuration and is applied as a hot probe. The probe's power is postulated to be present, and its

source is not identified.

Unless otherwise documented, all cables at HBRSEP were assumed to be thermo-plastic. Hot shorts

are postulated to occur regardless of the cable insulation type.



Comments


FIR-NGGC-0101, Fire Protection Nuclear Safety Capability Sections 3.17 and 9.3.3Assessment (NSCA)




the operation, or that result in the maloperation of the equipment identified in 2.4.2.1. This evaluationshall consider fire-induced failure modes such as hot shorts (external and internal), open circuits, andshorts to ground, to identify circuits that are required to support the proper operation of componentsrequired 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 commonenclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated fortheir impact on the ability to achieve nuclear safety performance criteria.



coordinated with the downstream protection device.(b) Common Enclosure Circuits. Those circuits that share enclosures with circuits required to achieve


3.5.2.3 CircuitFailures Due to aHot Short [B,Grounded Circuits]

NEI 00-01 Guidance

A Hot Short on Grounded Circuits

A short-to-ground is another failure mode for a grounded control circuit. A short-to-ground asdescribed above would result in de-energizing the circuit. This would further reduce the likelihood forthe circuit to change the state of the equipment either from a control switch or due to a hot short.Nevertheless, a hot short still needs to be considered. Figure 3.5.2-4 shows a typical groundedcontrol circuit that might be used for a motor-operated valve. However, the protective devices andposition indication lights that would normally be included in the control circuit for a motor-operatedvalve ha e been omitted, since these devices are not required to understany the concepts beingexplainec# in this section. In the discussion provided below, it is assumed that a single fire in a givenfire area could cause any one of the hot shorts depicted. The following discussion describes how toaddress the impact of these individual cable faults on the operation of the equipment controlled bythis circuit.


Page B-92HBRSEP LAR Rev 1



Hot short No. 1:

A hot short at this location would energize the close relay and result in the undesired closure of a

motor-operated valve.

Hot short No. 2:

A hot short at this location would energize the open relay and result in the undesired opening of a

motor-operated valve.

Applicability

Applicable

Comments


Statement

Hot shorts on grounded circuits were considered. Cables susceptible to grounds were identified withAligns the associated equipment.

Comments



Assessment (NSCA)



NFPA 805

Requirement













coordinated with the downstream protection device. I(b) Common Enclosure Circuits. Those circuits that share enclosLres with circuits required to achieve






Attachment B - NEI 04-02 Table B-2 Nuclear Safetyn 6 =LuI eI~~ LndedUIIL Capain I iissent ivietii u IUUIU Rxeview~

NEL0-01Ref

3.5.2.3 Circuit

Failures Due to a

Hot Short [C,

Ungrounded

Circuits]

Applicability

Applicable

NEI 00-01 Guidance

A Hot Short on Ungrounded Circuits

In the case of an ungrounded circuit, a single hot short may be sufficient to cause a spuriousoperation. A single hot short can cause a spurious operation if the hot short comes from a circuit fromthe positive leg of the same ungrounded source as the affected circuit.

In reviewing each of these cases, the common denominator is that in every case, the conductor inthe circuit between the control switch and the start/stop coil must be involved.

Figure 3.5.2-5 depicted below shows a typical ungrounded control circuit that might be used for amotor-operated valve. However, the protective devices and position indication lights that wouldnormally be included in the control circuit for a motor-operated valve have been omitted, since thesedevices are not required to understand the concepts being explained in this section.

In the discussion provided below, it is assumed that a single fire in a given fire area could cause anyone of the hot shorts depicted. The discussion provided below describes how to address the impactof these cable faults on the operation of the equipment controlled by this circuit.


Hot short No. 1:A hot short at this location from the same control power source would energize the close relay andresult in the undesired closure of a motor operated valve.

Hot short No. 2:A hot short at this location from the same control power source would energize the open relay and

result in the undesired opening of a motor operated valve.

Comments

AAlAlignment B

Statement

Hot shorts on ungrounded circuits were considered. Cables susceptible to grounds were identifiedAligns with the associated equipment. Hot shorts between two cables were considered credible in the RNP

analysis.

Comments



Assessment (NSCA)

RNP-E/ELE4_-1216, The Fire Safe Shutdown Analysis for H.B. Sectiin 3.1.1






NFPA 805

Requirement

Nuclear Safety Equipment and Cable Location. Physical location of equipment and cables shall be

identified.

NE[I0001Ref

3.5.2.4 Circuit

Failures Due to

Inadequate Circuit

Coordination

NEI 00-01 Guidance

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.

For the example shown in Figure 3.5.2-6, the circuit powered from load breaker 4 supplies power to a

non-safe shutdown pump. This circuit is damaged by fire in the same fire area as the circuit

providing power to from the Train B bus to the Train B pump, which is redundant to the Train A pump.

To assure safe shutdown for a fire in this fire area, the damage to the non-safe shutdown pump

powered from load breaker 4 of the Train A bus cannot impact the availability of the Train A pump,

which is redundant to the Train B pump. To assure that there is no impact to this Train A pump due

to the associated circuits' common power source breaker coordination issue, load breaker 4 must be

fully coordinated with the feeder breaker to the Train A bus.


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 protectiverelaying 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.


Attachment B - NEI 04-02 Table B-2 Nuclear SafetyCapability Assessment Methodology ReviewDuke Energy - 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.

CommentsApplicabilityApplicable

AlignmentStatement AlignmentBasi

AlignsPower cables for Safe Shutdown equipment have been selected for evaluation for all components

that are required to change states. Coordination of electrical breakers and fuses assure that other

power cables from loads on the same electrical bus or distribution center will not adversely impact

safe shutdown equipment. FIR-NGGC-0101 provides guidance on verifying that circuit coordination

exists, as well as methods for addressing cases where coordination is not readily apparent.

Circuit breaker and fuse coordination are verified by calculations.

Comments

Reference Document


RNP-E-8.005, 1OCFR50 Appendix R Associated Circuit, CommonPower Supply Analysis


RNP-E-9.021, 1OCFR50 Appendix R Fuse Analysis for DS Bus

RNP-E/ELEC-1216, The Fire Safe Shutdown Analysis for H.B.Robinson Nuclear Plant

Doc Detail

Section 9.3.6

Section 3.2.2





I

3.5.2.5 CircuitFailures Due toCommon Enclosure

HBRSEP LAR Rev 1

NEI 00101 Guidance

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.

Page B-96

Duke EnergyConcerns


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.

Applicability

Applicable

Comments

AlgnmentStatement

Circuit breaker and fuse coordination are verified by calculations. Adequate coordination exists to

Aligns assure that a common enclosure issue is not credible.

Comments

Reference Document DocDetall

FPP-RNP-200, 1OCFR50, Appendix R, Section lIl.G, Associated Section 4.0Circuits Analysis




Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval

L. NFPA 805 Chapter 3 Requirements for Approval10 CFR 50.48(c)(2)(vii)

13 Pages Attached

HBRSEP LAR Rev I Page L-1 IIHBRSEP LAR Rev I Page L-1 I


Approval Request 1

NFPA 805 Section 3.3.5.1

NFPA 805 Section 3.3.5.1 states:"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 inmetallic conduit, or routed in cable trays with solid metal top and bottom covers."

HBRSEP has wiring above suspended ceilings that may not comply with therequirements of this code section.

Suspended ceilings are noncombustible and exist only in the Control Room (FZ 23),Inside AO Office and old Turbine Building RCA Entrance (FZ 25A). Combustibles inconcealed spaces are minimal.

The three areas currently with suspended ceilings inside the NFPA 805 defined powerblock are in the Control Room (FZ 23), Inside AO Office and old Turbine Building RCAEntrance (FZ 25A). The Inside AO Office and old Turbine Building RCA Entrance (FZ25A) are not risk significant. Neither of the rooms nor the cables are safety-related.

Most electrical wiring above the Control Room partial suspended ceiling is in conduitexcept for short flexible connectors to lighting fixtures. There is one eight-foot length ofeight-inch diameter UL approved flexible air duct with flame spread rating of 25 or less.The quantity of cabling above the suspended ceilings in the Control Rooms is very lowand results in limited combustible loading. The existing fire detection capability and/orthe Control Room Operators who are continuously present in the area would identify thepresence of smoke. In addition, no equipment important to nuclear safety is located inthe vicinity of these cables.

Video/communication/data cables that have been field routed above suspended ceilingsare low voltage. Existing cables for video, communication, and networking may not beplenum rated, but are not generally susceptible to shorts that would result in a fire.

Basis for Request:

The basis for the approval request of this deviation is:" All electrical wiring above the control room partial suspended ceiling is in

conduit except for short flexible connectors to lighting fixtures. According toFAQ 06-0021, cable air drops of limited length (-3 feet) are consideredacceptable.

* No equipment important to nuclear safety is located in the vicinity of thesecables.

* Minimum amount of cables exist above the Control Room ceiling, which resultsin limited combustible loading.

* Smoke Detectors are installed both above and below the partial suspendedceiling in the Control Room.

" The Inside AO Office and old Turbine Building RCA Entrance (FZ 25A) are notrisk significant. Neither of the rooms nor the cables are safety related.

I

HBRSEP LAR Rev I Page L-2 I


* Existing fleet procedures will be used to ensure that changes moving forwardare considered for NFPA 805 impacts. (FIR-NGGC-0010)

Acceptance Criteria Evaluation:

Nuclear Safety and Radiological Release Performance Criteria:

The location of wiring above suspended ceilings does not affect nuclear safety. Powerand control cables comply with this section. No equipment important to nuclear safetyis located in the vicinity of these cables. Therefore, there is no impact on the nuclearsafety performance criteria.

The location of cables above suspended ceilings has no impact on the radiologicalrelease performance criteria. The radiological release review was performed based onthe manual fire suppression activities in areas containing or potentially containingradioactive materials and is not dependent on the type of cables or locations ofsuspended ceilings. The location of cables does not change the radiological releaseevaluation performed that potentially contaminated water is contained and smokemonitored. The cables do not add additional radiological materials to the area orchallenge system boundaries that contain such.


Duke Energy Aftachment L - NFPA 805 Chapter 3 Requirements for Approval

Safety Margin and Defense-in-Depth:

Power and control cables meet the requirements of this requirement. The use of thesematerials has been defined by the limitations of the analytical methods used in thedevelopment of the FPRA. Therefore, the inherent safety margin and conservatisms inthese methods remain unchanged.

The three echelons of defense-in-depth are 1) to prevent fires from starting(combustible/hot work controls), 2) rapidly detect, control and extinguish fires that dooccur thereby limiting damage (fire detection systems, automatic fire suppression,manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protectionfor systems and structures so that a fire will not prevent essential safety functions frombeing performed (fire barriers, fire rated cable, success path remains free of firedamage, recovery actions). The prior introduction of non-listedvideo/communication/data cables routed above suspended ceilings does not impact fireprotection defense-in-depth. Echelon 1 is maintained by the current cable installationprocedures documenting the requirements of NFPA 805 Section 3.3.5.1. The controlroom is a continuously manned area of the plant. The introduction of cables abovesuspended ceilings does not affect echelons 2 and 3. The video/communication/datacables routed above suspended ceilings does not result in compromising automatic firesuppression functions, manual fire suppression functions, fire protection for systemsand structures, or post-fire safe shutdown capability.

Conclusion:

HBRSEP determined that the performance based approach satisfies the followingcriteria:

* Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release

* Defense in Depth

" Safety Margin

HBRSEP LAR Rev I Page L-4 I!HBRSEP LAR Rev I Page L-4 I

Duke Energy Affachment L - NFPA 805 Chapter 3 Requirements for Approval

Approval Request 2

NFPA 805 Section 3.3.5.2

NFPA 805 Section 3.3.5.2 states:"Only metal tray and metal conduits shall be used for electrical raceways. Thin wallmetallic tubing shall not be used for power, instrumentation, or control cables. Flexiblemetallic conduits shall only be used in short lengths to connect components."

The use of PVC piping for underground embedded conduit is permitted by HBRSEP perHBR2-0B060 Sht D6 for electrical raceway installations. Polyvinyl Chloride (PVC) orHigh Density Polyethylene (HDPE) type ducts (conduits) are permitted when embeddedin compacted sand or reinforced concrete. In addition, some PVC conduit was found inreinforced concrete wall. The PVC/HDPE conduit is embedded within a noncombustibleenclosure which provides protection from mechanical damage and from damageresulting from either an exposure fire or from a fire within the conduit impacting othertargets.

Basis for Request:

" The PVC/HDPE conduit, while a combustible material, is not subject toflame/heat impingement from an external source which would result in structuralfailure, contribution to fire load, and damage to the circuits contained withinwhere the conduit is embedded in concrete or compacted sand.

" Failure of circuits within the conduit resulting in a fire would not result in damageto external targets.



The use of PVC/HDPE conduit in embedded locations does not affect nuclear safety asthe material in which conduits are run within an embedded location is not subject to thefailure mechanisms potentially resultant in circuit damage or resultant damage toexternal targets. Therefore there is no impact on the nuclear safety performance criteria.

The use of PVC/HDPE conduits in embedded installations has no impact on theradiological release performance criteria. The radiological release review wasperformed based on the manual fire suppression activities in areas containing orpotentially containing radioactive materials and is not dependent on the type of conduitmaterial. The conduit material does not change the radiological release evaluationperformed that concluded that potentially contaminated water is contained and smoke ismonitored. The conduits do not add additional radiological materials to the area orchallenge systems boundaries that contain such as the PVC/HDPE conduits areembedded.

HBRSEP LAR Rev I Page L-5

Duke Energy Affachment L - NFPA 805 Chapter 3 Requirements for Approval

Safety Margin and Defense-in-Depth:

The PVC/HDPE conduit material is embedded in a non-combustible configuration. Thematerial is protected when embedded from mechanical damage and from damageresulting from either an exposure fire or from a fire within the conduit impacting othertargets. The areas with PVC/HDPE conduit have been analyzed in their currentconfiguration. The precautions and limitations on the use of these materials do notimpact the analysis of the fire event. Therefore, the inherent safety margin andconservatisms in these analysis methods remain unchanged.

The three echelons of defense-in-depth are 1) to prevent fires from starting(combustible/hot work controls), 2) rapidly detect, control and extinguish fires that dooccur thereby limiting damage (fire detection systems, automatic fire suppression,manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protectionfor systems and structures so that a fire will not prevent essential safety functions frombeing performed (fire barriers, fire rated cable, success path remains free of firedamage, recovery actions). The use of PVC/HDPE conduits in embedded installationsdoes not impact fire protection defense-in-depth. The PVC/HDPE conduit in embeddedinstallations does not affect echelons 1, 2, and 3. The PVC/HDPE conduits do notdirectly result in compromising automatic fire suppression functions, manual firesuppression functions, or post-fire safe shutdown capability.

Conclusion:

HBRSEP determined that the performance based approach satisfies the followingcriteria"

* Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release

* Defense in Depth

" Safety Margin

HBRSEP LAR Rev I Page L-6 IIHBRSEP LAR Rev 1 Page L-6 I


Approval Request 3

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 toprovide backup to nuclear safety systems, provided the fire protection water supplysystems are designed and maintained to deliver the combined fire and nuclear safetyflow demands for the duration specified by the applicable analysis.Exception No. 2: Fire protection water storage can be provided by plant systems servingother functions, provided the storage has a dedicated capacity capable of providing themaximum fire protection demand for the specified duration as determined in thissection.

The review of plant flow diagrams show no hard connections to other plant systems,besides those for fire protection use. It should be noted that although there are no hardpipe connections to other plant systems, there are procedures that utilize the fireprotection water supply. They are as follows:

" AOP-014 - Loss of CCW

" AOP-022 - Loss of Service Water

* EDMG-001 - Extreme Damage Event Early Actions

* EDMG-002 - Refueling Water Storage Tank (RWST)

* EDMG-003 - Condensate Storage Tank (CST)

" EDMG-005 - Containment Vessel (CV)

" EDMG-01 1 - Spent Fuel Pit Casualty

* EDMG-012 - Core Cooling Using Alternate Water Source

* EDMG-01 3 -Airborne Release Scrubbing

" SAM-1 - Inject into the Steam Generator

* SAM-3 - Inject into the RCS

• SAM-4 - Inject into Containment

" SAM-6 - Control Containment Conditions

" SAM-8 - Flood Containment

The use of the fire protection water for these non-fire protection system water demandswould have no adverse impact on the ability of the fire protection system to providerequired flow and pressure. OMM-002, Section 8.15, details restrictions and allowancesfor use of the fire protection water supply system at HBRSEP.



Basis for Request:

The use of the fire protection water for these non-fire protection system water demandswould have no adverse impact on the ability of the fire protection system to providerequired flow and pressure. This is based on how fire water usage is restricted (CR 99-01247), in the following ways:

1. Fire service related activities (emergency, testing and training).

2. When the use of fire water is specifically called out in approved plantprocedures (i.e., AOPs).

3. During plant emergencies when fire water is needed to protect safety relatedequipment.

4. When usage is deemed necessary AND sufficient justification is provided toshow that the use of the fire water system for the proposed activity does notcause the fire water system to be in a condition outside of its design basis(i.e., the quantity of water needed for the proposed activity does not dropsupply and pressure below that required/defined in UFSAR Section 9.5.1).Permission shall have the approval of the Shift Manager (CR 96-00729 andCR 96-00730).

The water supply system is capable of maintaining the pressure in the main plant loopat 70 psi or higher with the largest deluge system in operation and with the systemsupplying an additional 1000 gpm to hoses.



The use of fire protection water for non-fire protection plant evolutions is an occurrencethat requires Shift Manager review and concurrence. The flow limitations to those non-fire protection functions ensure that there is no impact in the ability of the automaticsuppression systems to perform Therefore, there is no impact on the nuclear safetyperformance criteria.

The use of fire protection water for plant evolutions other than fire protection has noimpact on the radiological release performance criteria. The radiological releaseperformance criteria is satisfied based on the determination of limiting radioactiverelease (Attachment E), which is not affected by impacts on the fire protection systemdue it's use for non-fire protection purposes.

Safety Margin and Defense-i n-Depth:

The use of the fire water system, including the use of hydrants and hose, for non-fireprotection uses does not impact fire protection defense-in-depth. The fire pumps havethe excess capacity to supply the demands of the fire protection system as well as thenon-fire protection uses identified above. This does not compromise automatic ormanual fire suppression functions, fire suppression for systems and structures, or thenuclear safety capability assessment. Since both the automatic and manual firesuppression functions are maintained, defense-in-depth is maintained.

HBRSEP LAR Rev I -Fage L-8


The methods, input parameters, and acceptance criteria used in this analysis werereviewed and found to be in accordance with NFPA 805 Chapter 3. The methods, inputparameters, and acceptance criteria used to calculate flow requirements for theautomatic and manual suppression systems were not altered. Therefore, the safetymargin inherent in the analysis for the fire event has been preserved.

Conclusion:

HBRSEP determined that the performance based approach satisfies the followingcriteria:

" Satisfies the performance goals performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release

" Defense in Depth

* Safety Margin



Approval Request 4

NFPA 805 Section 3.2.3(1)In accordance with 10 CFR 50.48(c)(2)(vii), "Performance-based methods," the fireprotection program elements and minimum design requirements of Chapter 3 may besubject to the performance-based methods permitted elsewhere in the standard.

In accordance with NFPA 805 Section 2.2.8, the performance-based approach to satisfythe nuclear safety, radiation release, life safety, and property damage/businessinterruption performance criteria requires engineering analyses to evaluate whether theperformance criteria are satisfied.

In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shalldetermine that the performance-based approach utilized to evaluate a variance from therequirements of NFPA 805 Chapter 3:

A. Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release;

B. Maintains safety margins; andC. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire

suppression, mitigation, and post-fire nuclear safety capability).Duke Energy, HBRSEP requests formal approval of performance-based exception tothe requirements in Chapter 3 of NFPA 805 as follows:

NFPA 805, Section 3.2.3(1)"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:Inspection, testing, and maintenance for fire protection systems and featurescredited by the fire protection program."

Duke Energy, HBRSEP requests the ability to utilize performance-based methods toestablish the appropriate inspection, testing, and maintenance frequencies for fireprotection systems and features required by NFPA 805. Performance-based inspection,testing, and maintenance frequencies will be established as described in Electric PowerResearch Institute (EPRI) Technical Report TR-1006756, "Fire Protection SurveillanceOptimization and Maintenance Guide for Fire Protection", Final Report, July 2003.

Basis for Request:NFPA 805 Section 2.6, "Monitoring," requires that

"A monitoring program shall be established to ensure that the availability andreliability of the fire protection systems and features are maintained and to assessthe performance of the fire protection program in meeting the performance criteria.Monitoring shall ensure that the assumptlons in the engineering analysis remainvalid."

HBRSEP LAR Rev I Page 1-10 I

HBRSEP LAR Rev 1 Page L-10 I


NFPA 805 Section 2.6.1, "Availability, Reliability, and Performance Levels," requiresthat "Acceptable levels of availability, reliability, and performance shall beestablished."

NFPA 805 Section 2.6.2, "Monitoring Availability, Reliability, and Performance,"requires that "Methods to monitor availability, reliability, and performance shall beestablished. The methods shall consider the plant operating experience andindustry operating experience."

The scope and frequency of the inspection, testing, and maintenance activities for fireprotection systems and features required in the fire protection program have beenestablished based on the previously approved Technical Specifications / LicenseControlled Documents and appropriate NFPA codes and standard. This request doesnot involve the use of the EPRI Technical Report TR-1006756 to establish the scope ofthose activities as that is determined by the required systems review identified inAttachment C

This request is specific to the use of EPRI Technical Report TR-1006756 to establishthe appropriate inspection, testing, and maintenance frequencies for fire protectionsystems and features credited by the fire protection program. As stated in EPRITechnical Report TR-1 006756 Section 10.1, "The goal of a performance-basedsurveillance program is to adjust test and inspection frequencies commensurate withequipment performance and desired reliability." This goal is consistent with the statedrequirements of NFPA 805 Section 2.6. The EPRI Technical Report TR-1006756provides an accepted method to establish appropriate inspection, testing, andmaintenance frequencies which ensure the required NFPA 805 availability, reliability,and performance goals are maintained.

The target tests, inspections, and maintenance will be those activities for the NFPA 805required fire protection systems and features. The reliability and frequency goals will beestablished to ensure the assumptions in the NFPA 805 engineering analysis remainvalid. The failure criterion will be established based on the required fire protectionsystems and features credited functions and will ensure those functions are maintained.Data collection and analysis will follow the EPRI Technical Report TR-1 006756document guidance. The failure probability will be determined based on EPRI TechnicalReport TR-1 006756 guidance and a 95% confidence level will be utilized. Theperformance monitoring will be performed in conjunction with the Monitoring Programrequired by NFPA 805 Section 2.6 and it will ensure site specific operating experience isconsidered in the monitoring process. The following is a flow chart that identifies thebasic process that will be utilized.


Duke Energy Attachment L - NFPA 805 Chapter 3 Requirements for ApprovalDuke Energy Attachment L - NFPA 805 Chapter 3 Requirements for Approval

Program Framework

Identify Target Tests and Inspections

Establish Reliability and Frequency GoalsSet Failure Criteria

Assess Licensing Impact and Other Constraints

Data Collection and Evaluation

Establish Data Collection Guidelines

Collect Required Surveillance Data

Assemble Data in Spreadsheet or DatabaseAnalyze .Data to Identify Failures

Reliability and Uncertainty Analysis

Compute Failure ProbabilitiesCompute Uncertainty Limits

Confirm That Reliability Supports Target Frequency

Program ImplementationModify Program Documents

Revise -Surveil lance ProceduresConduct Ongoing Performance Monitoring

Refine and Modify Frequencies as Appropriate

rP RI TR-1 006756 - Figure 10-1Flowchart fo&Performance-Based Surveillance Program



Duke Energy, HBRSEP does not intend to revise any fire protection surveillance, test orinspection frequencies until after transitioning to NFPA 805. Existing fire protectionsurveillance, test and inspection will remain consistent with applicable station, Insurer,and NFPA Code requirements. HBRSEP's intent is to obtain approval via the NFPA 805Safety Evaluation to use EPRI Technical Report TR1 006756 guideline in the future asopportunities arise. Duke Energy, HBRSEP reserves the ability to evaluate fireprotection features with the intent of using the EPRI performance-based methods toprovide evidence of equipment performance beyond that achievable under traditionalprescriptive maintenance practices to ensure optimal use of resources while maintainingreliability.

Nuclear Safety and Radiological Release Performance Criteria:Use of performance-based test frequencies established per EPRI Technical ReportTR-1006756 methods combined with NFPA 805 Section 2.6, Monitoring Program, willensure that the availability and reliability of the fire protection systems and features aremaintained to the levels assumed in the NFPA 805 engineering analysis. Therefore,there is no adverse impact to Nuclear Safety Performance Criteria by the use of theperformance-based methods in EPRI Technical Report TR-1006756.

The radiological release performance criteria are satisfied based on the determination oflimiting radioactive release. Fire Protection Systems and Features may be credited aspart of that evaluation. Use of performance-based test frequencies established per theEPRI Technical Report TR-1 006756 methods combined with NFPA 805 Section 2.6,Monitoring Program, will ensure that the availability and reliability of the fire protectionsystems and features are maintained to the levels assumed in the NFPA 805engineering analysis which includes those assumptions credited to meet theRadioactive Release performance criteria. Therefore, there is no adverse impact toRadioactive Release performance criteria.

Safety Margin and Defense-in-Depth:Use of performance-based test frequencies established per EPRI Technical ReportTR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, willensure that the availability and reliability of the fire protection systems and features aremaintained to the levels assumed in the NFPA 805 engineering analysis which includesthose assumptions credited in the Fire Risk Evaluation safety margin discussions. Inaddition, the use of these methods in no way invalidates the inherent safety marginscontained in the codes and standards used for design and maintenance of fireprotection systems and features. Therefore, the safety margin inherent and credited inthe analysis has been preserved.

HBRSEP LAR Rev I Page L-13 IIHBRSEP LAR Rev I Page L-13 i


The three echelons of defense-in-depth described in NFPA 805 Section 1.2 are1) to 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 firesuppression, pre-fire plans), and

3) provide adequate level of fire protection for systems and structures so that afire will not prevent essential safety functions from being performed (firebarriers, fire rated cable, success path remains free of fire damage, recoveryactions).

Echelon 1 is not affected by the use of the EPRI Technical Report TR-1 006756methods. Use of performance-based test frequencies established per EPRI TechnicalReport TR-1006756 methods combined with NFPA 805 Section 2.6, MonitoringProgram, will ensure that the availability and reliability of the fire protection systems andfeatures credited for defense-in-depth are maintained to the levels assumed in theNFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2and 3 for defense-in-depth.

Conclusion:NRC approval is requested for use of the performance-based methods contained in theElectric Power Research Institute (EPRI) Technical Report TR-1006756, "FireProtection Equipment Surveillance Optimization and Maintenance Guide", Final Report,July 2003 to establish the appropriate inspection, testing, and maintenance frequenciesfor fire protection systems and features required by NFPA 805. As described above, thisapproach is considered acceptable because it:

A. Satisfies the performance goals, performance objectives, and performancecriteria specified in NFPA 805 related to nuclear safety and radiological release;

B. Maintains safety margins; andC. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire

suppression, mitigation, and post-fire safe shutdown capability).


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