NUREG-2171, 'Safety Evaluation Report Related to …Documents available from public and special...

Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2

Office of Nuclear Reactor Regulation

NUREG-2171

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Safety Evaluation Report Related to the License Renewal of Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Manuscript Completed: September 2014 Date Published: September 2014 Office of Nuclear Reactor Regulation

NUREG-2171

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ABSTRACT This safety evaluation report (SER) documents the technical review of the Limerick Generating Station (LGS), Units 1 and 2, license renewal application (LRA) by the United States Nuclear Regulatory Commission (NRC) staff (the staff). By letter dated June 22, 2011, Exelon Generation Company, LLC submitted the LRA in accordance with Title 10 of the Code of Federal Regulations Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants.” Exelon requests renewal of the LGS Units 1 and 2 operating licenses (Operating License Nos. NPF-39 and NPF-85) for a period of 20 years beyond the current expiration at midnight October 26, 2024, and June 22, 2029, respectively. LGS is located approximately 21 miles northwest of Philadelphia, PA. The NRC issued the LGS Units 1 and 2 construction permits on June 19, 1974, and the operating licenses for LGS Unit 1 on August 8, 1985, and LGS Unit 2 on August 25, 1989. LGS Units 1 and 2 are of a boiling-water reactor design. General Electric supplied the nuclear steam supply system and Bechtel originally designed and constructed the balance of the plant. LGS Units 1 and 2 both have a licensed power output of 3,515 megawatts thermal.

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TABLE OF CONTENTS ABSTRACT ..............................................................................................................................iii LIST OF TABLES .................................................................................................................... xiii ABBREVIATIONS ....................................................................................................................xv SECTION 1 INTRODUCTION AND GENERAL DISCUSSION .............................................. 1-1

1.1 Introduction .................................................................................................. 1-1 1.2 License Renewal Background ..................................................................... 1-2

1.2.1 Safety Review .................................................................................... 1-3 1.2.2 Environmental Review ....................................................................... 1-4

1.3 Principal Review Matters ............................................................................. 1-5 1.4 Interim Staff Guidance ................................................................................. 1-6 1.5 Summary of Open Items .............................................................................. 1-7 1.6 Summary of Confirmatory Items .................................................................. 1-8 1.7 Summary of Proposed License Conditions .................................................. 1-8

SECTION 2 STRUCTURES AND COMPONENTS SUBJECT TO AGING MANAGEMENT REVIEW .................................................................................. 2-1 2.1 Scoping and Screening Methodology .......................................................... 2-1

2.1.1 Introduction ........................................................................................ 2-1 2.1.2 Summary of Technical Information in the Application ......................... 2-1 2.1.3 Scoping and Screening Program Review ........................................... 2-1

2.1.3.1 Implementation Procedures and Documentation Sources for Scoping and Screening .................................... 2-2

2.1.3.2 Quality Controls Applied to LRA Development .................... 2-5 2.1.3.3 Training ............................................................................... 2-6 2.1.3.4 Conclusion of Scoping and Screening Program Review ...... 2-6

2.1.4 Plant Systems, Structures, and Components Scoping Methodology .................................................................................... 2-7 2.1.4.1 Application of the Scoping Criteria in 10 CFR 54.4(a)(1) ..... 2-7 2.1.4.2 Application of the Scoping Criteria in 10 CFR 54.4(a)(2) ..... 2-9 2.1.4.3 Application of the Scoping Criteria in 10 CFR 54.4(a)(3) ... 2-18 2.1.4.4 Plant-Level Scoping of Systems and Structures ................ 2-20 2.1.4.5 Mechanical Component Scoping ....................................... 2-21 2.1.4.6 Structural Component Scoping ......................................... 2-23 2.1.4.7 Electrical Component Scoping .......................................... 2-24 2.1.4.8 Conclusion for Scoping Methodology ................................ 2-25

2.1.5 Screening Methodology ................................................................... 2-25 2.1.5.1 General Screening Methodology ....................................... 2-25 2.1.5.2 Mechanical Component Screening .................................... 2-27 2.1.5.3 Structural Component Screening ...................................... 2-28 2.1.5.4 Electrical Component Screening ....................................... 2-29 2.1.5.5 Conclusion for Screening Methodology ............................. 2-29

2.1.6 Summary of Evaluation Findings ..................................................... 2-30 2.2 Plant-Level Scoping Results ...................................................................... 2-30

2.2.1 Introduction ...................................................................................... 2-30 2.2.2 Summary of Technical Information in the Application ....................... 2-30 2.2.3 Staff Evaluation ............................................................................... 2-30 2.2.4 Conclusion ....................................................................................... 2-31

2.3 Scoping and Screening Results: Mechanical Systems ............................... 2-31 2.3.1 Reactor Vessel, Internals, and Reactor Coolant System .................. 2-32

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2.3.1.1 Reactor Coolant Pressure Boundary ................................. 2-33 2.3.1.2 Reactor Pressure Vessel .................................................. 2-34 2.3.1.3 Reactor Vessel Internals ................................................... 2-35

2.3.2 Engineered Safety Features ............................................................ 2-36 2.3.2.1 Containment Atmosphere Control System ........................ 2-36 2.3.2.2 Core Spray System ........................................................... 2-37 2.3.2.3 High Pressure Coolant Injection System ........................... 2-37 2.3.2.4 Reactor Core Isolation Cooling System ............................. 2-38 2.3.2.5 Residual Heat Removal System ........................................ 2-39 2.3.2.6 Standby Gas Treatment System ....................................... 2-41

2.3.3 Auxiliary Systems ............................................................................ 2-41 2.3.3.1 Auxiliary Steam System .................................................... 2-43 2.3.3.2 Closed Cooling Water System .......................................... 2-44 2.3.3.3 Compressed Air System ................................................... 2-45 2.3.3.4 Control Enclosure Ventilation System ............................... 2-45 2.3.3.5 Control Rod Drive System ................................................. 2-47 2.3.3.6 Cranes and Hoists ............................................................ 2-48 2.3.3.7 Emergency Diesel Generator Enclosure Ventilation

System.............................................................................. 2-48 2.3.3.8 Emergency Diesel Generator System ............................... 2-50 2.3.3.9 Fire Protection System ...................................................... 2-52 2.3.3.10 Fuel Handling and Storage System ................................. 2-57 2.3.3.11 Fuel Pool Cooling and Cleanup System .......................... 2-57 2.3.3.12 Nonsafety-Related Service Water System ...................... 2-58 2.3.3.13 Plant Drainage System ................................................... 2-59 2.3.3.14 Primary Containment Instrument Gas System ................. 2-60 2.3.3.15 Primary Containment Leak Testing System..................... 2-62 2.3.3.16 Primary Containment Ventilation System ........................ 2-62 2.3.3.17 Process Radiation Monitoring System ............................. 2-64 2.3.3.18 Process and Post-Accident Sampling System ................. 2-65 2.3.3.19 Radwaste System ........................................................... 2-66 2.3.3.20 Reactor Enclosure Ventilation System ............................ 2-67 2.3.3.21 Reactor Water Cleanup System ...................................... 2-69 2.3.3.22 Safety-Related Service Water System ............................ 2-69 2.3.3.23 Spray Pond Pump House Ventilation System.................. 2-71 2.3.3.24 Standby Liquid Control System ....................................... 2-73 2.3.3.25 Traversing Incore Probe System ..................................... 2-74 2.3.3.26 Water Treatment and Distribution System ....................... 2-74

2.3.4 Steam and Power Conversion Systems ........................................... 2-75 2.3.4.1 Circulating Water System .................................................. 2-76 2.3.4.2 Condensate System .......................................................... 2-77 2.3.4.3 Condenser and Air Removal System ................................ 2-77 2.3.4.4 Extraction Steam System .................................................. 2-78 2.3.4.5 Feedwater System ............................................................ 2-79 2.3.4.6 Main Steam System .......................................................... 2-80 2.3.4.7 Main Turbine ..................................................................... 2-81

2.4 Scoping and Screening Results: Structures ............................................... 2-82 2.4.1 220 and 500 kV Substations ............................................................ 2-83

2.4.1.1 Summary of Technical Information in the Application ........ 2-83 2.4.1.2 Staff Evaluation ................................................................. 2-84

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2.4.1.3 Conclusion ........................................................................ 2-84 2.4.2 Admin Building Shop and Warehouse .............................................. 2-84

2.4.2.1 Summary of Technical Information in the Application ........ 2-84 2.4.2.2 Staff Evaluation ................................................................. 2-85 2.4.2.3 Conclusion ........................................................................ 2-85

2.4.3 Auxiliary Boiler and Lube Oil Storage Enclosure .............................. 2-85 2.4.3.1 Summary of Technical Information in the Application ........ 2-85 2.4.3.2 Staff Evaluation ................................................................. 2-86 2.4.3.3 Conclusion ........................................................................ 2-86

2.4.4 Circulating Water Pump House ........................................................ 2-86 2.4.4.1 Summary of Technical Information in the Application ........ 2-86 2.4.4.2 Staff Evaluation ................................................................. 2-86 2.4.4.3 Conclusion ........................................................................ 2-87

2.4.5 Component Supports Commodities Group ....................................... 2-87 2.4.5.1 Summary of Technical Information in the Application ........ 2-87 2.4.5.2 Staff Evaluation ................................................................. 2-87 2.4.5.3 Conclusion ........................................................................ 2-87

2.4.6 Control Enclosure ............................................................................ 2-88 2.4.6.1 Summary of Technical Information in the Application ........ 2-88 2.4.6.2 Staff Evaluation ................................................................. 2-88 2.4.6.3 Conclusion ........................................................................ 2-89

2.4.7 Cooling Towers ................................................................................ 2-89 2.4.7.1 Summary of Technical Information in the Application ........ 2-89 2.4.7.2 Staff Evaluation ................................................................. 2-89 2.4.7.3 Conclusion ........................................................................ 2-90

2.4.8 Diesel Oil Storage Tank Structures .................................................. 2-90 2.4.8.1 Summary of Technical Information in the Application ........ 2-90 2.4.8.2 Staff Evaluation ................................................................. 2-90 2.4.8.3 Conclusion ........................................................................ 2-91

2.4.9 Emergency Diesel Generator Enclosure .......................................... 2-91 2.4.9.1 Summary of Technical Information in the Application ........ 2-91 2.4.9.2 Staff Evaluation ................................................................. 2-92 2.4.9.3 Conclusion ........................................................................ 2-92

2.4.10 Piping and Component Insulation Commodity Group ..................... 2-92 2.4.10.1 Summary of Technical Information in the Application ...... 2-92 2.4.10.2 Staff Evaluation ............................................................... 2-92 2.4.10.3 Conclusion ...................................................................... 2-92

2.4.11 Primary Containment ..................................................................... 2-93 2.4.11.1 Summary of Technical Information in the Application ...... 2-93 2.4.11.2 Staff Evaluation ............................................................... 2-93 2.4.11.3 Conclusion ...................................................................... 2-93

2.4.12 Radwaste Enclosure ...................................................................... 2-93 2.4.12.1 Summary of Technical Information in the Application ...... 2-93 2.4.12.2 Staff Evaluation ............................................................... 2-94 2.4.12.3 Conclusion ...................................................................... 2-94

2.4.13 Reactor Enclosure ......................................................................... 2-94 2.4.13.1 Summary of Technical Information in the Application ...... 2-94 2.4.13.2 Staff Evaluation ............................................................... 2-95 2.4.13.3 Conclusion ...................................................................... 2-95

2.4.14 Service Water Pipe Tunnel ............................................................ 2-95

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2.4.14.1 Summary of Technical Information in the Application ...... 2-95 2.4.14.2 Staff Evaluation ............................................................... 2-95 2.4.14.3 Conclusion ...................................................................... 2-96

2.4.15 Spray Pond and Pump House ........................................................ 2-96 2.4.15.1 Summary of Technical Information in the Application ...... 2-96 2.4.15.2 Staff Evaluation ............................................................... 2-96 2.4.15.3 Conclusion ...................................................................... 2-97

2.4.16 Turbine Enclosure .......................................................................... 2-97 2.4.16.1 Summary of Technical Information in the Application ...... 2-97 2.4.16.2 Staff Evaluation ............................................................... 2-97 2.4.16.3 Conclusion ...................................................................... 2-97

2.4.17 Yard Facilities ................................................................................ 2-97 2.4.17.1 Summary of Technical Information in the Application ...... 2-97 2.4.17.2 Staff Evaluation ............................................................... 2-98 2.4.17.3 Conclusion ...................................................................... 2-98

2.5 Scoping and Screening Results: Electrical ................................................. 2-98 2.5.1 Electrical and Instrumentation and Controls Commodity Groups ..... 2-99

2.5.1.1 Summary of Technical Information in the Application ........ 2-99 2.5.1.2 Staff Evaluation ................................................................. 2-99 2.5.1.3 Conclusion ...................................................................... 2-101

2.6 Conclusion for Scoping and Screening .................................................... 2-101 SECTION 3 AGING MANAGEMENT REVIEW RESULTS .................................................... 3-1

3.0 Applicant’s Use of the Generic Aging Lessons Learned Report ................... 3-1 3.0.1 Format of the License Renewal Application ....................................... 3-2

3.0.1.1 Overview of Table 1s .......................................................... 3-2 3.0.1.2 Overview of Table 2s .......................................................... 3-3

3.0.2 Staff’s Review Process ...................................................................... 3-4 3.0.2.1 Review of AMPs .................................................................. 3-5 3.0.2.2 Review of AMR Results ...................................................... 3-6 3.0.2.3 UFSAR Supplement ............................................................ 3-7 3.0.2.4 Documentation and Documents Reviewed .......................... 3-7

3.0.3 Aging Management Programs ........................................................... 3-8 3.0.3.1 AMPs Consistent with the GALL Report ............................ 3-11 3.0.3.2 AMPs Consistent with the GALL Report with Exceptions

or Enhancements .............................................................. 3-80 3.0.4 QA Program Attributes Integral to Aging Management Programs .. 3-169

3.0.4.1 Summary of Technical Information in the Application ...... 3-169 3.0.4.2 Staff Evaluation ............................................................... 3-170 3.0.4.3 Conclusion ...................................................................... 3-171

3.0.5 Operating Experience for Aging Management Programs ............... 3-171 3.0.5.1 Summary of Technical Information in Application ............ 3-171 3.0.5.2 Staff Evaluation ............................................................... 3-171 3.0.5.3 UFSAR Supplement ........................................................ 3-178 3.0.5.4 Conclusion ...................................................................... 3-179

3.1 Aging Management of Reactor Vessel, Internals and Reactor Coolant System .................................................................................................... 3-180 3.1.1 Summary of Technical Information in the Application ..................... 3-180 3.1.2 Staff Evaluation ............................................................................. 3-180

3.1.2.1 AMR Results Consistent with the GALL Report ............... 3-203

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3.1.2.2 AMR Results Consistent with the GALL Report for Which Further Evaluation is Recommended ................... 3-217

3.1.2.3 AMR Results Not Consistent with or Not Addressed in the GALL Report ............................................................. 3-225

3.1.3 Conclusion ..................................................................................... 3-229 3.2 Aging Management of Engineered Safety Features Systems .................. 3-229

3.2.1 Summary of Technical Information in the Application ..................... 3-230 3.2.2 Staff Evaluation ............................................................................. 3-230

3.2.2.1 AMR Results Consistent with the GALL Report ............... 3-241 3.2.2.2 AMR Results Consistent with the GALL Report for

Which Further Evaluation is Recommended ................... 3-249 3.2.2.3 AMR Results Not Consistent with or Not Addressed in

the GALL Report ............................................................. 3-251 3.2.3 Conclusion ..................................................................................... 3-255

3.3 Aging Management of Auxiliary Systems ................................................. 3-255 3.3.1 Summary of Technical Information in the Application ..................... 3-256 3.3.2 Staff Evaluation ............................................................................. 3-256




3.4 Aging Management of Steam and Power Conversion Systems ............... 3-319 3.4.1 Summary of Technical Information in the Application ..................... 3-319 3.4.2 Staff Evaluation ............................................................................. 3-320




3.5 Aging Management of Structures and Component Supports ................... 3-341 3.5.1 Summary of Technical Information in the Application ..................... 3-342 3.5.2 Staff Evaluation ............................................................................. 3-342




3.6 Aging Management of Electrical and Instrumentation and Controls System .................................................................................................... 3-393 3.6.1 Summary of Technical Information in the Application ..................... 3-393 3.6.2 Staff Evaluation ............................................................................. 3-394



the GALL Report ............................................................. 3-410

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3.6.3 Conclusion ..................................................................................... 3-413 3.7 Conclusion for Aging Management Review Results ................................. 3-413

SECTION 4 TIME-LIMITED AGING ANALYSES .................................................................. 4-1 4.1 Identification of Time-Limited Aging Analyses .............................................. 4-1

4.1.1 Summary of Technical Information in the Application ......................... 4-2 4.1.2 Staff Evaluation ................................................................................. 4-3 4.1.3 UFSAR Supplement ........................................................................ 4-19 4.1.4 Conclusion ....................................................................................... 4-19

4.2 Reactor Vessel Neutron Embrittlement ...................................................... 4-19 4.2.1 Neutron Fluence .............................................................................. 4-20

4.2.1.1 Summary of Technical Information in the Application ........ 4-20 4.2.1.2 Staff Evaluation ................................................................. 4-21 4.2.1.3 UFSAR Supplement .......................................................... 4-23 4.2.1.4 Conclusion ........................................................................ 4-24

4.2.2 Upper-Shelf Energy ......................................................................... 4-24 4.2.2.1 Summary of Technical Information in the Application ........ 4-24 4.2.2.2 Staff Evaluation ................................................................. 4-24 4.2.2.3 UFSAR Supplement .......................................................... 4-25 4.2.2.4 Conclusion ........................................................................ 4-26

4.2.3 Adjusted Reference Temperature .................................................... 4-26 4.2.3.1 Summary of Technical Information in the Application ........ 4-26 4.2.3.2 Staff Evaluation ................................................................. 4-27 4.2.3.3 UFSAR Supplement .......................................................... 4-27 4.2.3.4 Conclusion ........................................................................ 4-28

4.2.4 Pressure – Temperature (P-T) Limits ............................................... 4-28 4.2.4.1 Summary of Technical Information in the Application ........ 4-28 4.2.4.2 Staff Evaluation ................................................................. 4-28 4.2.4.3 UFSAR Supplement .......................................................... 4-29 4.2.4.4 Conclusion ........................................................................ 4-29

4.2.5 Axial Weld Inspection ...................................................................... 4-29 4.2.5.1 Summary of Technical Information in the Application ........ 4-29 4.2.5.2 Staff Evaluation ................................................................. 4-30 4.2.5.3 UFSAR Supplement .......................................................... 4-30 4.2.5.4 Conclusion ........................................................................ 4-30

4.2.6 Circumferential Weld Inspection ...................................................... 4-31 4.2.6.1 Summary of Technical Information in the Application ........ 4-31 4.2.6.2 Staff Evaluation ................................................................. 4-31 4.2.6.3 UFSAR Supplement .......................................................... 4-32 4.2.6.4 Conclusion ........................................................................ 4-33

4.2.7 Reactor Vessel Reflood Thermal Shock .......................................... 4-33 4.2.7.1 Summary of Technical Information in the Application ........ 4-33 4.2.7.2 Staff Evaluation ................................................................. 4-33 4.2.7.3 UFSAR Supplement .......................................................... 4-34 4.2.7.4 Conclusion ........................................................................ 4-34

4.3 Metal Fatigue ............................................................................................. 4-34 4.3.1 ASME Code Section III, Class 1 Fatigue Analysis ............................ 4-34


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4.3.2 ASME Code Section III, Class 2 and 3 and ANSI B31.1 Allowable Stress Calculations ........................................................................ 4-40 4.3.2.1 Summary of Technical Information in the Application ........ 4-40 4.3.2.2 Staff Evaluation ................................................................. 4-40 4.3.2.3 UFSAR Supplement .......................................................... 4-44 4.3.2.4 Conclusion ........................................................................ 4-44

4.3.3 Environmental Fatigue Analyses for RPV and Class 1 Piping .......... 4-45 4.3.3.1 Summary of Technical Information in the Application ........ 4-45 4.3.3.2 Staff Evaluation ................................................................. 4-45 4.3.3.3 UFSAR Supplement .......................................................... 4-57 4.3.3.4 Conclusion ........................................................................ 4-58

4.3.4 Reactor Vessel Internals (RVI) Fatigue Analyses............................. 4-58 4.3.4.1 Summary of Technical Information in the Application ........ 4-58 4.3.4.2 Staff Evaluation ................................................................. 4-58 4.3.4.3 UFSAR Supplement .......................................................... 4-61 4.3.4.4 Conclusion ........................................................................ 4-62

4.3.5 High-Energy Line Break (HELB) Analyses Based Upon Fatigue ...... 4-62 4.3.5.1 Summary of Technical Information in the Application ........ 4-62 4.3.5.2 Staff Evaluation ................................................................. 4-62 4.3.5.3 UFSAR Supplement .......................................................... 4-63 4.3.5.4 Conclusion ........................................................................ 4-63

4.4 Environmental Qualification of Electric Equipment ..................................... 4-64 4.4.1 Summary of Technical Information in the Application ....................... 4-64 4.4.2 Staff Evaluation ............................................................................... 4-64 4.4.3 UFSAR Supplement ........................................................................ 4-65 4.4.4 Conclusion ....................................................................................... 4-65

4.5 Containment Liner Plate and Penetration Fatigue Analyses ...................... 4-66 4.5.1 Summary of Technical Information in the Application ....................... 4-66 4.5.2 Staff Evaluation ............................................................................... 4-66 4.5.3 UFSAR Supplement ........................................................................ 4-68 4.5.4 Conclusion ....................................................................................... 4-68

4.6 Other Plant-Specific TLAAs ....................................................................... 4-68 4.6.1 Reactor Enclosure Crane Cyclic Loading Analysis ........................... 4-68


4.6.2 Emergency Diesel Generator Enclosure Cranes ............................. 4-70 4.6.2.1 Summary of Technical Information in the Application ........ 4-70 4.6.2.2 Staff Evaluation ................................................................. 4-70 4.6.2.3 UFSAR Supplement .......................................................... 4-71 4.6.2.4 Conclusion ........................................................................ 4-71

4.6.3 RPV Core Plate Rim Hold-down Bolt Loss of Preload ..................... 4-71 4.6.3.1 Summary of Technical Information in the Application ........ 4-71 4.6.3.2 Staff Evaluation ................................................................. 4-72 4.6.3.3 UFSAR Supplement .......................................................... 4-72 4.6.3.4 Conclusion ........................................................................ 4-72

4.6.4 Main Steam Line Flow Restrictors Erosion ...................................... 4-73 4.6.4.1 Summary of Technical Information in the Application ........ 4-73 4.6.4.2 Staff Evaluation ................................................................. 4-73

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4.6.4.3 UFSAR Supplement .......................................................... 4-74 4.6.4.4 Conclusion ........................................................ 4-74

4.6.5 Jet Pump Auxiliary Spring Wedge Assembly .................................. 4-74 4.6.5.1 Summary of Technical Information in the Application ........ 4-74 4.6.5.2 Staff Evaluation ................................................................. 4-75 4.6.5.3 UFSAR Supplement .......................................................... 4-77 4.6.5.4 Conclusion ........................................................................ 4-78

4.6.6 Jet Pump Restrainer Bracket Pad Repair Clamps ........................... 4-78 4.6.6.1 Summary of Technical Information in the Application ........ 4-78 4.6.6.2 Staff Evaluation ................................................................. 4-78 4.6.6.3 UFSAR Supplement .......................................................... 4-80 4.6.6.4 Conclusion ........................................................................ 4-80

4.6.7 Refueling Bellows and Support Cyclic Loading Analysis .................. 4-80 4.6.7.1 Summary of Technical Information in the Application ........ 4-80 4.6.7.2 Staff Evaluation ................................................................. 4-81 4.6.7.3 UFSAR Supplement .......................................................... 4-82 4.6.7.4 Conclusion ........................................................................ 4-82

4.6.8 Downcomers and MSRV Discharge Piping ...................................... 4-82 4.6.8.1 Summary of Technical Information in the Application ........ 4-82 4.6.8.2 Staff Evaluation ................................................................. 4-83 4.6.8.3 UFSAR Supplement .......................................................... 4-87 4.6.8.4 Conclusion ........................................................................ 4-87

4.6.9 Jet Pump Slip Joint Repair Clamps .................................................. 4-87 4.6.9.1 Summary of Technical Information in the Application ........ 4-87 4.6.9.2 Staff Evaluation ................................................................. 4-88 4.6.9.3 UFSAR Supplement .......................................................... 4-89 4.6.9.4 Conclusion ........................................................................ 4-89

4.6.10 Fuel Pool Girder Loss of Prestress ................................................ 4-89 4.6.10.1 Summary of Technical Information in the Application ...... 4-89 4.6.10.2 Staff Evaluation ............................................................... 4-90 4.6.10.3 UFSAR Supplement ........................................................ 4-91 4.6.10.4 Conclusion ...................................................................... 4-91

4.6.11 RHR and Core Spray Suction Strainer Fatigue Analysis ................ 4-91 4.6.11.1 Summary of Technical Information in the Application ...... 4-91 4.6.11.2 Staff Evaluation ............................................................... 4-91 4.6.11.3 UFSAR Supplement ........................................................ 4-93 4.6.11.4 Conclusion ...................................................................... 4-93

4.7 Conclusion for TLAAs ................................................................................ 4-93 SECTION 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ... 5-1 SECTION 6 CONCLUSION ................................................................................................... 6-1

Appendices

APPENDIX A: Limerick Generating Station, Units 1 and 2 License Renewal Commitments ............................................................................................... A-1 APPENDIX B: Chronology ................................................................................................... B-1 APPENDIX C: Principal Contributors.................................................................................. C-1 APPENDIX D: References .................................................................................................... D-1

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LIST OF TABLES Table 1.4-1 Current Interim Staff Guidance ......................................................................... 1-7 Table 3.0.3-1 Aging Management Programs .......................................................................... 3-8 Table 3.1-1 Staff Evaluation for Reactor Vessel, Reactor Vessel Internals and Reactor Coolant System Components in the SRP-LR .................................. 3-181 Table 3.2-1 Staff Evaluation for Engineered Safety Features Systems Components in the GALL Report ..................................................................................... 3-2311 Table 3.3-1 Staff Evaluation for Auxiliary Systems Components in the GALL Report ...... 3-257 Table 3.4-1 Staff Evaluation for Steam and Power Conversion Systems Components in the GALL Report ....................................................................................... 3-321 Table 3.5-1 Staff Evaluation for Structures and Component Supports Components in the GALL Report ....................................................................................... 3-343 Table 3.6-1 Staff Evaluation for Electrical and Instrumentation and Controls Components in the GALL Report .................................................................. 3-395

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ABBREVIATIONS AAI applicant action item ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards ADAMS Agencywide Documents Access and Management System AERM aging effect requiring management AFW auxiliary feedwater AMP aging management program AMR aging management review ANSI American National Standards Institute ART adjusted reference temperature ASCE American Society of Civil Engineers ASME American Society of Mechanical Engineers AST alternate source term ASTM American Society for Testing and Materials ATWS anticipated transient without scram B&W Babcock and Wilcox BTP branch technical position BWR boiling-water reactor BWRVIP Boiling Water Reactor Vessel Integrity Project CAP corrective action program CASS cast austenitic stainless steel CFR Code of Federal Regulations CLB current licensing basis CMAA Crane Manufacturers Association of America CMRT certified material test record CRD control rod drive CRL component record list CSC containment spray cooling CS core spray CST condensate storage tank Cu copper CUF cumulative usage factor CuFen environmentally adjusted fatigue usage factor CW circulating water DBA design-basis accident DBE design-basis event EAF environmentally assisted fatigue ECCS emergency core cooling system EDG emergency diesel generator EFPY effective full-power year EMA equivalent margins analysis EPRI Electric Power Research Institute EQ environmental qualification

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ESF engineered safety features ESW emergency service water Fen environmental fatigue life correction factor FERC Federal Energy Regulatory Commission FIV flow induced vibration FR Federal Register ft-lb foot-pound GALL Generic Aging Lessons Learned Report GDC general design criteria or general design criterion GE General Electric GEIS generic environmental impact statement GL generic letter GSI generic safety issue HELB high-energy line break HPCI high-pressure coolant injection HPSI high-pressure safety injection HVAC heating, ventilation, and air conditioning I&C instrumentation and controls IASCC irradiation assisted stress corrosion cracking ICMH in-core monitoring housing ID inside diameter IGSCC intergranular stress corrosion cracking ILRT integrated leak rate test IN information notice INPO Institute of Nuclear Power Operations IPA integrated plant assessment ISG interim staff guidance ISI inservice inspection ISP Integrated Surveillance Program kV kilovolt LBB leak-before-break LER licensee event report LGS Limerick Generating Station LLRT local leak rate test LOCA loss-of-coolant accident LPCI low-pressure coolant injection LPRM low-power range monitor LRA license renewal application LTOP low-temperature overpressure protection MC metal containment MEB metal enclosed bus MIC microbiologically influenced corrosion MoS2 molybdenum disulfide

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MSIP mechanical stress improvement process MSIV main steam isolation valve MSRV main steam relief valve MUR measurement uncertainty recapture n/cm2 neutrons per square centimeter NDE nondestructive examination NEI Nuclear Energy Institute NFPA National Fire Protection Association Ni nickel NPS nominal pipe size NRC U.S. Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation OBE operational basis earthquake OE operating experience OI open item PCIG primary containment instrument gas PDL Plastics Design Library pH potential of hydrogen P&ID plant piping and instrumentation drawing PoF probability of failure P-T pressure-temperature PTS pressurized thermal shock PVC polyvinyl chloride PWR pressurized-water reactor PWSCC primary water stress corrosion cracking QA quality assurance QAP quality assurance program RAI request for additional information RAMA Radiation Analysis Modeling Application RCIC reactor core isolation cooling RCP reactor coolant pump RCPB reactor coolant pressure boundary RCS reactor coolant system RCSC Research Council for Structural Connections RG regulatory guide RHR residual heat removal RHRSW residual heat removal service water RI-ISI risk-informed inservice inspection RPV reactor pressure vessel RTNDT reference temperature nil ductility transition RVI reactor vessel internals RWCU reactor water cleanup

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SBO station blackout SC structure and component SCC stress corrosion cracking SDC shutdown cooling SDV scram discharge volume SER safety evaluation report SGTS standby gas treatment system SLC standby liquid control SOER significant operating experience reports SPC suppression pool cooling SPU stretch power uprate SRP standard review plan SRP-LR “Standard Review Plan for Review of License Renewal Applications for Nuclear

Power Plants” SRV safety relief valve SSC system, structure, and component SSE safe-shutdown earthquake SW service water TIP traversing in-core probe TLAA time-limited aging analysis TS technical specifications TTA tolyltriazole UFSAR updated final safety analysis report USE upper-shelf energy UT ultrasonic examination UV ultraviolet VFLD vessel flange leak detector WLI water level instrumentation WTD water treatment and distribution yr year Zn zinc

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SECTION 1

INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction This document is a safety evaluation report (SER) on the license renewal application (LRA) for Limerick Generating Station (LGS), Units 1 and 2, as filed by Exelon Generation Company, LLC (Exelon or the applicant). By letter dated June 22, 2011, Exelon submitted its application to the United States (U.S.) Nuclear Regulatory Commission (NRC) for renewal of the LGS operating licenses for an additional 20 years. The staff prepared this report to summarize the results of its safety review of the LRA for compliance with Title 10 of the Code of Federal Regulations (10 CFR) Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants.” The NRC project manager for the license renewal review is Robert Kuntz. Mr. Kuntz may be contacted by telephone at 301-415-3733 or by email at [email protected]. Alternatively, written correspondence may be sent to the following address: Division of License Renewal U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attention: Robert Kuntz, Mail Stop 011-F1 In its June 22, 2011, submission letter, the applicant requested renewal of the operating licenses issued under Section 103 (Operating Licenses No. NPF-39 and NPF-85) of the Atomic Energy Act of 1954, as amended, for LGS Units 1 and 2 for a period of 20 years beyond the current expiration at midnight October 26, 2024, and June 22, 2029, respectively. LGS is located approximately 21 miles northwest of Philadelphia, PA. The NRC issued the LGS Units 1 and 2 construction permits on June 19, 1974, and the operating license for LGS Unit 1 on August 8, 1985, and LGS Unit 2 on August 25, 1989. LGS Units 1 and 2 are of a boiling-water reactor design. General Electric supplied the nuclear steam supply system and Bechtel originally designed and constructed the balance of the plant. LGS Units 1 and 2 both have a licensed power output of 3,515 megawatts thermal. The updated final safety analysis report (UFSAR) shows details of the plant and the site. The license renewal process consists of two concurrent reviews, a technical review of safety issues and an environmental review. The NRC regulations in 10 CFR Part 54 and 10 CFR Part 51, “Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions,” respectively, set forth requirements for these reviews. The safety review for the LGS license renewal is based on the applicant’s LRA and responses to the staff’s requests for additional information (RAIs). The applicant supplemented the LRA and provided clarifications through its responses to the staff’s RAIs in audits, meetings, and docketed correspondence. Unless otherwise noted, the staff reviewed and considered information submitted through July 11, 2012. The staff reviewed information received after this date depending on the stage of the safety review and the volume and complexity of the information. The public may view the LRA and all pertinent information and materials, including the UFSAR, at the NRC Public Document Room located on the first floor of One White Flint North, 11555 Rockville Pike, Rockville, MD 20852-2738 (301-415-4737 or 800-397-4209), and at Pottstown Regional Public Library, 500 East High Street, Pottstown, PA 19464-5656. In

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addition, the public may find the LRA, as well as materials related to the license renewal review, on the NRC website at http://www.nrc.gov. This SER summarizes the results of the staff’s safety review of the LRA and describes the technical details considered in evaluating the safety aspects of proposed operation of Units 1 and 2 for an additional 20 years beyond the term of the current operating licenses. The staff reviewed the LRA in accordance with NRC regulations and the guidance in NUREG-1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants” (SRP-LR), issued December 2010. SER Sections 2 through 4 address the staff’s evaluation of license renewal issues considered during the review of the application. SER Section 5 is reserved for the report of the Advisory Committee on Reactor Safeguards (ACRS). The SER conclusions are in Section 6. SER Appendix A is a table showing the applicant’s commitments for renewal of the operating licenses. SER Appendix B is a chronology of the principal correspondence between the staff and the applicant regarding the LRA review. SER Appendix C is a list of principal contributors to the SER, and Appendix D is a bibliography of the references in support of the staff’s review. In accordance with 10 CFR Part 51, the staff prepared a draft plant-specific supplement to NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS).” This supplement discusses the environmental considerations for license renewal for LGS Units 1 and 2. The staff plans to issue a draft, plant-specific GEIS supplement. The final, plant-specific GEIS supplement will then be issued after consideration of public comment on the draft plant-specific GEIS. 1.2 License Renewal Background Pursuant to the Atomic Energy Act of 1954, as amended, and NRC regulations, operating licenses for commercial power reactors are issued for 40 years and can be renewed for up to 20 additional years. The original 40-year license term was selected based on economic and antitrust considerations rather than on technical limitations; however, some individual plant and equipment designs may have been engineered for an expected 40-year service life. In 1982, the staff anticipated interest in license renewal and held a workshop on nuclear power plant aging. This workshop led the NRC to establish a comprehensive program plan for nuclear plant aging research. From the results of that research, a technical review group concluded that many aging phenomena are readily manageable and pose no technical issues precluding life extension for nuclear power plants. In 1986, the staff published a request for comment on a policy statement that would address major policy, technical, and procedural issues related to license renewal for nuclear power plants. In 1991, the staff published 10 CFR Part 54, the License Renewal Rule (Volume 56, page 64943, of the Federal Register (FR) (56 FR 64943), dated December 13, 1991). The staff participated in an industry-sponsored demonstration program to apply 10 CFR Part 54 to a pilot plant and to gain the experience necessary to develop implementation guidance. To establish a scope of review for license renewal, 10 CFR Part 54 defined age-related degradation unique to license renewal. However, during the demonstration program, the staff found that adverse aging effects on plant systems and components are managed during the initial license period and that the scope of the review did not allow sufficient credit for management programs,

http://www.nrc.gov/

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particularly the implementation of 10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,” which regulates management of plant-aging phenomena. As a result of this finding, the staff amended 10 CFR Part 54 in 1995. As published May 8, 1995, (60 FR 22461), amended 10 CFR Part 54 establishes a simpler, more stable, and more predictable regulatory process than the previous 10 CFR Part 54. In particular, as amended, 10 CFR Part 54 focuses on the management of adverse aging effects rather than on the identification of age-related degradation unique to license renewal. The staff made these rule changes to ensure that important systems, structures, and components (SSCs) will continue to perform their intended functions during the period of extended operation. In addition, the amended 10 CFR Part 54 clarifies and simplifies the integrated plant assessment (IPA) process to be consistent with the revised focus on passive, long-lived structures and components (SCs). Concurrent with these initiatives, the staff pursued a separate rulemaking effort (61 FR 28467, June 5, 1996) and amended 10 CFR Part 51 to focus the scope of the review of environmental impacts of license renewal to fulfill NRC responsibilities under the National Environmental Policy Act of 1969. 1.2.1 Safety Review License renewal requirements for power reactors are based on two key principles: (1) The regulatory process is adequate to ensure that the licensing bases of all currently

operating plants maintain an acceptable level of safety, with the possible exceptions of the detrimental aging effects on the functions of certain SSCs, as well as a few other safety-related issues, during the period of extended operation.

(2) The plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term.

In implementing these two principles, 10 CFR 54.4, “Scope,” defines the scope of license renewal as including those SSCs that (1) are safety-related, (2) whose failure could affect safety-related functions, or (3) are relied on to demonstrate compliance with the NRC’s regulations for fire protection, environmental qualification (EQ), pressurized thermal shock (PTS), anticipated transient without scram (ATWS), and station blackout (SBO). Pursuant to 10 CFR 54.21(a), a license renewal applicant must review all SSCs within the scope of 10 CFR Part 54 to identify SCs subject to an aging management review (AMR). Those SCs subject to an AMR perform an intended function without moving parts or without change in configuration or properties and are not subject to replacement based on a qualified life or specified time period. Pursuant to 10 CFR 54.21(a), a license renewal applicant must demonstrate that the aging effects will be managed such that the intended function(s) of those SCs will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. However, active equipment is considered to be adequately monitored and maintained by existing programs. In other words, detrimental aging effects that may affect active equipment can be readily identified and corrected through routine surveillance, performance monitoring, and maintenance. Surveillance and maintenance programs for active equipment, as well as other maintenance aspects of plant design and licensing basis, are required throughout the period of extended operation.

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In accordance with 10 CFR 54.21(d), the LRA is required to include a UFSAR supplement with a summary description of the applicant’s programs and activities for managing aging effects and an evaluation of time-limited aging analyses (TLAAs) for the period of extended operation. License renewal also requires TLAA identification and updating. During the plant design phase, certain assumptions about the length of time the plant can operate are incorporated into design calculations for several plant SSCs. In accordance with 10 CFR 54.21(c)(1), the applicant must either show that these calculations will remain valid for the period of extended operation, project the analyses to the end of the period of extended operation, or demonstrate that the aging effects on these SSCs will be adequately managed for the period of extended operation. In 2005, the NRC revised Regulatory Guide (RG) 1.188, “Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses.” This RG endorses Nuclear Energy Institute (NEI) 95-10, Revision 6, “Industry Guideline for Implementing the Requirements of 10 CFR Part 54 – the License Renewal Rule,” issued in June 2005. NEI 95-10 details an acceptable method of implementing 10 CFR Part 54. The staff also used the SRP-LR to review the LRA. In the LRA, the applicant fully used the process defined in NUREG-1801, Revision 2, “Generic Aging Lessons Learned (GALL) Report,” issued December 2010. The GALL Report summarizes staff-approved aging management programs (AMPs) for many SCs subject to an AMR. If an applicant commits to implementing these staff-approved AMPs, the time, effort, and resources for LRA review can be greatly reduced, improving the efficiency and effectiveness of the license renewal review process. The GALL Report summarizes the aging management evaluations, programs, and activities credited for managing aging for most of the SCs used throughout the industry. The report also is a quick reference for both applicants and staff reviewers to AMPs and activities that can manage aging adequately during the period of extended operation. 1.2.2 Environmental Review Regulations on environmental protection are contained in 10 CFR Part 51. In December 1996, the staff revised the environmental protection regulations to facilitate the environmental review for license renewal. The staff prepared the GEIS to document its evaluation of possible environmental impacts associated with nuclear power plant license renewals. For certain types of environmental impacts, the GEIS contains generic findings that apply to all nuclear power plants and are codified in Appendix B, “Environmental Effect of Renewing the Operating License of a Nuclear Power Plant,” to Subpart A, “National Environmental Policy Act – Regulations Implementing Section 102(2),” of 10 CFR Part 51. Pursuant to 10 CFR CFR 51.53(c)(3)(i), a license renewal applicant may incorporate these generic findings in its environmental report. In accordance with 10 CFR 51.53(c)(3)(ii), an environmental report also must include analyses of environmental impacts that must be evaluated on a plant-specific basis (i.e., Category 2 issues). In accordance with the National Environmental Policy Act of 1969 and 10 CFR Part 51, the staff is reviewing the plant-specific environmental impacts of license renewal, including whether there is new and significant information not considered in the GEIS. As part of its scoping process, the staff held a public meeting on September 22, 2011, at the Sunnybrook Ballroom, 50 North Sunnybrook Road, Pottstown, PA 19464, to identify plant-specific environmental issues. The draft plant-specific GEIS supplement will document the results of the environmental review and will make a preliminary recommendation as to the license renewal action. The staff will hold

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another public meeting to discuss the draft plant-specific GEIS supplement. After considering comments on the draft, the staff will publish the final plant-specific Supplement to the GEIS separately from this report. 1.3 Principal Review Matters Requirements for renewal of operating licenses for nuclear power plants are described in 10 CFR Part 54. The staff’s technical review of the LRA was in accordance with NRC guidance and 10 CFR Part 54 requirements. The license renewal standards are set forth in 10 CFR 54.29, “Standards for Issuance of a Renewed License.” This SER describes the results of the staff’s safety review. In accordance with 10 CFR 54.19(a), the NRC requires a license renewal applicant to submit general information, which the applicant provided in LRA Section 1. The staff reviewed LRA Section 1 and finds that the applicant has submitted the required information. In accordance with 10 CFR 54.19(b), the NRC requires that the LRA include “conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license.” On this issue, the applicant stated in the LRA:

10 CFR 54.19(b) requires that “each application must include conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license.” The current indemnity agreement (B-101) for LGS states in Article VII that the agreement shall terminate at the time of expiration of that license specified in Item 3 of the Attachment to the agreement, which is the last to expire; provided that, except as may otherwise be provided in applicable regulations or orders of the Commission, the term of this agreement shall not terminate until all the radioactive material has been removed from the location and transportation of the radioactive material from the location has ended as defined in subparagraph 5(b), Article I. Item 3 of the Attachment to the indemnity agreement includes license number SNM-1926. Applicant requests that any necessary conforming changes be made to Article VII and Item 3 of the Attachment, and any other sections of the indemnity agreement as appropriate to ensure that the indemnity agreement continues to apply during both the terms of the current licenses and the terms of the renewed licenses. Applicant understands that no changes may be necessary for this purpose if the current license numbers are retained.

The staff intends to maintain the original license numbers upon issuance of the renewed licenses, if approved. Therefore, conforming changes to the indemnity agreement need not be made and the 10 CFR 54.19(b) requirements have been met. In accordance with 10 CFR 54.21, “Contents of Application – Technical Information,” the NRC requires that the LRA contain (a) an IPA, (b) a description of any CLB changes during the staff’s review of the LRA, (c) an evaluation of TLAAs, and (d) an UFSAR supplement. LRA Sections 3 and 4 and Appendix B address the license renewal requirements of 10 CFR 54.21(a), (b), and (c). LRA Appendix A satisfies the license renewal requirements of 10 CFR 54.21(d).

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In accordance with 10 CFR 54.21(b), the NRC requires that each year following submission of the LRA and at least 3 months before the scheduled completion of the staff’s review, the applicant submit an LRA amendment identifying any CLB changes to the facility that affect the contents of the LRA, including the UFSAR supplement. By letter dated June 14, 2012, the applicant submitted an LRA update, which summarizes the CLB changes that have occurred during the staff’s review of the LRA. This submission satisfies 10 CFR 54.21(b) requirements. Pursuant to 10 CFR 54.22, “Contents of Application – Technical Specifications,” the NRC requires that the LRA include changes or additions to the technical specifications (TS) necessary to manage aging effects during the period of extended operation. In LRA Appendix D, the applicant stated that it had not identified any TS changes necessary for issuance of the renewed LGS Units 1 and 2 operating licenses. This statement adequately addresses the 10 CFR 54.22 requirements. The staff evaluated the technical information required by 10 CFR 54.21 and 10 CFR 54.22 in accordance with NRC regulations and SRP-LR guidance. SER Sections 2, 3, and 4 document the staff’s evaluation of the LRA technical information. As required by 10 CFR 54.25, “Report of the Advisory Committee on Reactor Safeguards,” ACRS will issue a report documenting its evaluation of the staff’s LRA review and SER. SER Section 5 is reserved for the ACRS report when it is issued. SER Section 6 documents the findings required by 10 CFR 54.29. 1.4 Interim Staff Guidance License renewal is a living program. The staff, industry, and other interested stakeholders gain experience and develop lessons learned with each renewed license. The lessons learned address the staff’s performance goals of maintaining safety, improving effectiveness and efficiency, reducing regulatory burden, and increasing public confidence. Interim staff guidance (ISG) is documented for use by the staff, industry, and other interested stakeholders until incorporated into such license renewal guidance documents as the SRP-LR and the GALL Report. Table 1.4-1 shows the current set of ISGs, as well as the SER sections in which the staff addresses them.

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Table 1.4-1 Current Interim Staff Guidance

ISG Issue (Approved ISG Number)

Purpose SER Section

“Staff Guidance for Preparing Severe Accident Mitigation Alternatives Analyses”

(LR-ISG-2006-03)

This ISG is related to severe accident management alternatives for environmental impact statements

N/A for the SER

“Ongoing Review of Operating Experience”

(LR-ISG-2011-05)

This LR-ISG clarifies the staff’s existing position in the SRP-LR that acceptable license renewal AMPs should be informed and enhanced when necessary, based on the ongoing review of both plant-specific and industry operating experience.

SER Section 3.0.5

1.5 Summary of Open Items As a result of its review of the LRA, including additional information submitted through July 11, 2012, the staff had identified the following open items (OI) when it issued the SER with Open Items on July 30, 2012. An item is considered open if, in the staff’s judgment, it does not meet all applicable regulatory requirements at the time of the issuance of this SER. The staff has assigned a unique identifying number to each OI. Open Item 3.0.3.2.13-1 ASME Code Section XI, Subsection IWE LGS Units 1 and 2 have seen corrosion in the suppression pool liner and downcomers. The applicant’s proposed aging management of the suppression pool liner and downcomers is within the scope of the American Society of Mechanical Engineers (ASME) Code Section XI, Subsection IWE program. As described in SER Section 3.0.3.2.13, the staff had an open item for aging management of the suppression pool liner and downcomers. Specifically, the open item was related to the following concerns: • The applicant has developed an acceptance criterion for the degradation of the

downcomers; however, this criterion is not identified in the AMP or the associated procedures.

• The criteria used for selecting locations for recoating (i.e., criteria for coating degradation, general corrosion, and pitting corrosion) may not be adequate. In addition, it is not clear how the coating degradation can be effectively identified for each liner plate underwater in the suppression pool. Also, the applicant’s proposed criteria for augmented inspection is not consistent with the ASME Code, Section XI, Subsection IWE requirement that detailed visual and ultrasonic thickness measurement be completed on 100 percent of surface areas subjected to accelerated corrosion or areas where the absence or repeated loss of coatings has resulted in substantial corrosion or pitting.

Based on its audit and review of the application, and review of the applicant’s response to the open item, the staff finds that the program elements for which the applicant claimed consistency

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with the GALL Report are consistent with the corresponding program elements of GALL Report AMP. Open Item 3.0.3.2.13-1 is closed.

Open Item 3.0.5.1 Operating Experience for Aging Management Programs LR-ISG-2011-05 states that enhancements to the existing programmatic activities for the ongoing review of operating experience that are necessary for license renewal should be put in place no later than the date the renewed operating licenses are issued. The applicant described several enhancements; however, it planned to implement them after issuance of the renewed licenses. As discussed in SER Section 3.0.5, the staff could not determine whether operating experience related to aging management and age-related degradation will be considered in the period between issuance of the renewed licenses and implementation of the enhancements. In response, the applicant stated that the enhancements to the Operating Experience program will be implemented no later than the date when the renewed operating licenses are issued and conducted on an ongoing basis throughout the terms of the renewed licenses. The staff finds this implementation schedule acceptable because it is consistent with the guidance in LR-ISG-2011-05. Implementation of these enhancements will ensure that the applicant fully considers all available information to inform the aging management activities on an ongoing basis throughout the terms of the renewed licenses. Open Item 3.0.5-1 is closed.

1.6 Summary of Confirmatory Items As a result of its review of the LRA, including additional information submitted through July 11, 2012 the staff determines that no confirmatory items exist that would require a formal response from the applicant. 1.7 Summary of Proposed License Conditions Following the staff’s review of the LRA, including subsequent information and clarifications from the applicant, the staff identified two proposed license conditions. The first license condition requires the applicant to include the UFSAR supplement required by 10 CFR 54.21(d) in the next UFSAR update, required by 10 CFR 50.71(e), following the issuance of the renewed licenses. The applicant may make changes to the programs and activities described in the UFSAR supplement, provided the applicant evaluates such changes in accordance with the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section. The second license condition requires future activities described in the UFSAR supplement to be completed before the period of extended operation. In its SER with Open Items issued on July 30, 2012, the staff proposed that the applicant shall complete these activities no later than six months before the period of extended operation, and shall notify the NRC in writing when implementation of these activities is complete. In particular, the NRC is directing the applicant to complete certain license renewal activities no later than 6 months prior to PEO in order to ensure the completion of its inspection requirements under NRC Inspection Procedure (IP) 71003, “Post-Approval Site Inspection for License Renewal.” Through this IP, the staff verifies that the license renewal commitments and selected AMPs are satisfactorily implemented, the description of the AMPs and related activities are, or will be, contained in the UFSAR, and the description of the programs is consistent with the programs implemented by the licensee. Notwithstanding the “Enhancement or Implementation Schedule” detailed in

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Appendix A, “Limerick Generating Station, Units 1 and 2, License Renewal Commitments,” to this SER and the NRC staff’s findings presented in various sections of the SER, the scheduler requirements proposed in the second license condition shall take precedence. In a letter dated October 12, 2012, the applicant provided its comments on this license condition. The applicant stated that this proposed license condition would require completion of most activities described in the license renewal commitment list six months earlier than it had committed to perform these activities. The applicant further stated that the proposed license condition creates consequences that the staff may not have intended or appreciated. Specifically, the current operating licenses for Units 1 and 2 expire on October 26, 2024, and June 22, 2029, respectively, and the applicant performs its refueling outages in the spring. A license condition requiring that the activities be completed at least six months prior to entering the PEO would mean that the applicant would not have the opportunity to perform inspections during the last scheduled refueling outage prior to PEO for Units 1 or 2. Thus, the applicant concluded that by not allowing aging management activities to be performed in the last refueling outage prior to the PEO, there are additional undesirable consequences. For example, certain aging management programs specifically require that inspections be done close to the PEO to allow more time for aging effects to develop and be detected by inspection. The staff reviewed the applicant’s comments and supporting basis and found that certain aspects of the proposed license condition could preclude scheduling actions to both obtain better performance of the specific aging management program activities and make more use of outage work periods. On this basis, the proposed second license condition was revised to state that:

The applicant’s UFSAR supplement submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, describes certain programs to be implemented and activities to be completed prior to the period of extended operation. a. The applicant shall implement those new programs and enhancements to existing

programs no later than 6 months prior to PEO.

b. The applicant shall complete those activities as noted in Commitment Nos. 18, 19, 20, 22, 23, 24, 28, 29, 30, 38, 39, 40, 41, 42, 43, and 47 of Appendix A of NUREG-XXXX, “Limerick Safety Evaluation Report for License Renewal,” by the 6-month date prior to PEO or the end of the last refueling outage prior to the PEO, whichever occurs later.

The applicant shall notify the NRC in writing within 30 days after having accomplished item (a) above and include the status of those activities that have been or remain to be completed in item (b) above.

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SECTION 2

STRUCTURES AND COMPONENTS SUBJECT TO AGING MANAGEMENT REVIEW

2.1 Scoping and Screening Methodology 2.1.1 Introduction Title 10 of the Code of Federal Regulations (10 CFR) 54.21, “Contents of Application – Technical Information,” requires Exelon Generation Company, LLC (Exelon or the applicant) to identify the structures, systems, and components (SSCs) within the scope of license renewal in accordance with 10 CFR 54.4(a). In addition, the license renewal application (LRA) must contain an integrated plant assessment (IPA) that identifies and lists those structures and components (SCs), contained in the SSCs identified to be within the scope of license renewal, that are subject to an aging management review (AMR). 2.1.2 Summary of Technical Information in the Application LRA Section 2, “Scoping and Screening Methodology for Identifying Structures and Components Subject to Aging Management Review, and Implementation Results,” provides the technical information required by 10 CFR 54.21(a). LRA Section 2.1, “Scoping and Screening Methodology,” describes the methodology used by the applicant to identify the SSCs at the Limerick Generating Station (LGS), Units 1 and 2, within the scope of license renewal (scoping) and the SCs subject to an AMR (screening). LRA Section 2.1.1, “Introduction,” states, in part, that the applicant had considered the following in developing the scoping and screening methodology described in LRA Section 2: • 10 CFR Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power

Plants,” (the rule) • Nuclear Energy Institute (NEI) 95-10, Revision 6, “Industry Guideline for Implementing the

Requirements of 10 CFR Part 54 – the License Renewal Rule,” issued June 2005 (NEI 95-10)

2.1.3 Scoping and Screening Program Review The staff of the U.S. Nuclear Regulatory Commission (NRC) (the staff) evaluated the applicant’s scoping and screening methodology in accordance with the guidance contained in NUREG-1800, Revision 2, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,” (SRP-LR), Section 2.1, “Scoping and Screening Methodology.” The following regulations provide the basis for the acceptance criteria the staff used to assess the

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adequacy of the scoping and screening methodology that the applicant used to develop the LRA: • 10 CFR 54.4(a), as it relates to the identification of SSCs within the scope of the rule • 10 CFR 54.4(b), as it relates to the identification of the intended functions of SSCs within the

scope of the rule • 10 CFR 54.21(a), as it relates to the methods used by the applicant to identify plant SCs

subject to an AMR The staff reviewed the information in LRA Section 2.1 to ensure that the applicant described a process for identifying SSCs that are within the scope of license renewal in accordance with the requirements of 10 CFR 54.4(a) and the SCs that are subject to an AMR in accordance with the requirements of 10 CFR 54.21(a). In addition, the staff conducted a scoping and screening methodology audit at the LGS site during the week of September 19–23, 2011. The audit focused on ensuring that the applicant had developed and implemented adequate guidance to conduct the scoping and screening of SSCs in accordance with the methodology described in the LRA and the requirements of the rule. The staff reviewed the project-level guidelines, topical reports, and implementing procedures that described the applicant’s scoping and screening methodology. The staff conducted detailed discussions with the applicant on the development of the license renewal application, the quality practices the applicant used during LRA development, and the training of the applicant’s staff that participated in LRA development. On a sampling basis, the staff performed a review of scoping and screening results reports and supporting current licensing basis (CLB) information for the safety-related service water (SW) system and the turbine building. In addition, the staff performed walkdowns of selected portions of the essential SW system, fuel pool cooling and cleanup system, emergency diesel generator (EDG) fuel oil transfer subsystem, EDG air start subsystem, and the turbine building, as a part of the sampling review of the implementation of the applicant’s 10 CFR 54.4(a)(2) scoping methodology. 2.1.3.1 Implementation Procedures and Documentation Sources for Scoping and Screening The staff reviewed the applicant’s scoping and screening implementing procedures, as documented in the “Scoping and Screening Methodology Audit Report Regarding the Limerick Generating Station, Units 1 and 2,” dated December 9, 2011, to verify that the process used to identify SSCs within the scope of license renewal and SCs subject to an AMR was consistent with the SRP-LR. Additionally, the staff reviewed the scope of CLB documentation and the process the applicant used, relative to the requirements of 10 CFR 54.4, “Scope,” and 10 CFR 54.21, and it confirmed that the applicant adequately implemented its procedural guidance during the scoping and screening process. 2.1.3.1.1 Summary of Technical Information in the Application In LRA Section 2.1, the applicant addressed the following information sources for the license renewal scoping and screening process: • updated final safety analysis report (UFSAR)

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• fire protection evaluation report • environmental qualification (EQ) master list • maintenance rule database • design baseline documents • component record list (CRL) • other CLB references, such as NRC safety evaluation reports (SERs), licensing

correspondence, engineering drawings, and engineering evaluations and calculations 2.1.3.1.2 Staff Evaluation Scoping and Screening Implementation Procedures. The staff reviewed the applicant’s scoping and screening methodology implementing procedures, including license renewal guidelines, documents and reports, as documented in the staff’s audit report, to ensure that the guidance is consistent with the requirements of the rule, and with the guidance in the SRP-LR and Regulatory Guide (RG) 1.188, “Standard Format and Content for Applications To Renew Nuclear Plant Operating Licenses,” which endorses the use of NEI 95-10. The staff finds the overall process used to implement the 10 CFR Part 54 requirements described in the implementing procedures and AMRs is consistent with the rule, the SRP-LR, and the NRC-endorsed industry guidance. The applicant’s implementing procedures contain guidance for determining plant SSCs within the scope of the rule and SCs contained in systems within the scope of license renewal that are subject to an AMR. During the review of the implementing procedures, the staff focused on the consistency of the detailed procedural guidance with information contained in the LRA, including the implementation of staff positions documented in the SRP-LR, and the information in the applicant’s responses dated January 27, 2012, to the staff’s requests for additional information (RAIs), dated January 5, 2012. After reviewing the LRA and supporting documentation, the staff determined that the scoping and screening methodology instructions are consistent with the methodology description provided in LRA Section 2.1. The applicant’s methodology is sufficiently detailed in the implementing procedures to provide concise guidance on the scoping and screening process to be followed during the LRA activities. Sources of CLB Information. Regulations in 10 CFR 54.21(a)(3) require for each structure and component determined to be subject to an AMR to demonstrate that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the CLB for the period of extended operation. The CLB is defined in 10 CFR 54.3(a), in part, as the set of NRC requirements applicable to a specific plant and an applicant’s written commitments for ensuring compliance with, and operation within, applicable NRC requirements and the plant-specific design bases that are docketed and in effect. The CLB includes applicable NRC regulations, orders, license conditions, exemptions, technical specifications, and design-basis information (documented in the most recent UFSAR). The CLB also includes licensee commitments remaining in effect that were made in docketed licensing correspondence, such as licensee responses to NRC bulletins, generic letters, and enforcement actions, and licensee commitments documented in NRC safety evaluations or licensee event reports.

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During the scoping and screening methodology audit, the staff confirmed that the applicant’s detailed license renewal program guidelines specified the use of the CLB source information in developing scoping evaluations. The staff reviewed pertinent information sources that the applicant used, including the UFSAR, design-basis information, and plant piping and instrumentation drawings (P&IDs). In addition, the staff determined that the applicant had used additional sources of plant information pertinent to the scoping and screening process, including the CRL, analyses, and reports. The staff determined that the applicant’s primary repository for system identification and component safety classification information was the CRL, UFSAR, and P&IDs. During the audit, the staff discussed the applicant’s administrative controls for the CRL and the other information sources used to verify system information. These controls are described and implemented by plant procedures. Based on a review of the administrative controls, and a sample of the system classification information contained in the applicable documentation, the staff concludes that the applicant has established adequate measures to control the integrity and reliability of system identification and safety classification data; therefore, the staff determined that the information sources the applicant used during the scoping and screening process provided a controlled source of system and component data to support scoping and screening evaluations. In addition, the staff reviewed the implementing procedures and results reports used to support identification of SSCs that the applicant relied on to demonstrate compliance with the requirements of 10 CFR 54.4(a). The applicant’s license renewal program guidelines provided a list of documents used to support scoping evaluations. The staff determined that the design documentation sources, required to be used by the applicant’s implementing procedures, provided sufficient information to ensure that the applicant identified SSCs to be included within the scope of license renewal that were consistent with the plan

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