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Safety Evaluation Report Related to the License Renewal of Columbia Generating Station Volume 1 Docket Number 50-397 Energy Northwest

Office of Nuclear Reactor Regulation

NUREG-2123 Volume 1

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Safety Evaluation Report Related to the License Renewal of Columbia Generating Station Volume 1

Docket Number 50-397 Energy Northwest Manuscript Completed: February 2012 Date Published: May 2012 Office of Nuclear Reactor Regulation

NUREG-2123 Volume 1

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ABSTRACT

This safety evaluation report (SER) documents the technical review of the Columbia Generating Station (Columbia), license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff (the staff). By letter dated January 19, 2010, Energy Northwest (the applicant) submitted the LRA in accordance with Title 10, Part 54, of the Code of Federal Regulations, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants.” Energy Northwest requests renewal of the operating license (Facility Operating License Number NPF-21) for a period of 20 years beyond the current license period of December 20, 2023. Columbia is located approximately 12 miles north of Richland, WA. The NRC issued the construction permit on March 19, 1973, and the operating license for Columbia on April 13, 1984. The unit is a Mark II boiling-water reactor (BWR) design. General Electric Company supplied the nuclear steam supply system. Burns and Roe, Inc., designed the balance of plant, and Bechtel Power Corporation constructed the plant. The licensed power output of the unit is 3,886 megawatts thermal, with a gross electrical output of approximately 1,230 megawatts electric. This SER presents the status of the staff’s review of information submitted through January 4, 2012. The staff closed six open items previously identified in the SER with open items. SER Section 1.5 summarizes the closure of the open items.

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TABLE OF CONTENTS

ABSTRACT ............................................................................................................................... iii

TABLE OF CONTENTS ............................................................................................................. v

LIST OF TABLES .................................................................................................................... xiv

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 the Open Items .............................................................................1-6 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 Information Sources Used for Scoping and Screening........................2-1 2.1.3 Scoping and Screening Program Review ...........................................2-2

2.1.3.1 Implementing Procedures and Documentation Sources Used for Scoping and Screening ..........................2-3

2.1.3.2 Quality Controls Applied to License Renewal Application Development ......................................................................2-5

2.1.3.3 Training ..............................................................................2-6 2.1.3.4 Scoping and Screening Program Review Conclusion .........2-7

2.1.4 Plant Systems, Structures, and Components Scoping Methodology .......................................................................................2-7 2.1.4.1 Application of the Scoping Criteria in Title 10,

Part 54.4(a)(1) of the Code of Federal Regulations ............2-7 2.1.4.2 Application of the Scoping Criteria in Title 10,

Part 54.4(a)(2) of the Code of Federal Regulations .......... 2-11 2.1.4.3 Application of the Scoping Criteria in Title 10,

Part 54.4(a)(3) of the Code of Federal Regulations .......... 2-19 2.1.4.4 Plant-Level Scoping of Systems and Structures ............... 2-22 2.1.4.5 Mechanical Component Scoping ...................................... 2-23 2.1.4.6 Structural Scoping ............................................................ 2-25 2.1.4.7 Electrical Component Scoping .......................................... 2-25 2.1.4.8 Scoping Methodology Conclusion ..................................... 2-26

2.1.5 Screening Methodology .................................................................... 2-27 2.1.5.1 General Screening Methodology ...................................... 2-27 2.1.5.2 Mechanical Component Screening ................................... 2-28 2.1.5.3 Structural Component Screening ...................................... 2-31 2.1.5.4 Electrical Component Screening....................................... 2-33

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2.1.5.5 Screening Methodology Conclusion ................................. 2-34 2.1.6 Summary of Evaluation Findings ...................................................... 2-34

2.2 Plant-Level Scoping Results ........................................................................... 2-34 2.2.1 Introduction ...................................................................................... 2-34 2.2.2 Summary of Technical Information in the Application ....................... 2-34 2.2.3 Staff Evaluation ................................................................................ 2-35 2.2.4 Conclusion ....................................................................................... 2-36

2.3 Scoping and Screening Results: Mechanical Systems ................................... 2-36 2.3.1 Reactor Vessel, Internals, and Reactor Coolant

Pressure Boundary ........................................................................... 2-37 2.3.1.1 Reactor Pressure Vessel .................................................. 2-37 2.3.1.2 Reactor Vessel Internals .................................................. 2-39 2.3.1.3 Reactor Coolant Pressure Boundary ................................ 2-40

2.3.2 Engineered Safety Features ............................................................. 2-40 2.3.2.1 Residual Heat Removal System ....................................... 2-41 2.3.2.2 Reactor Core Isolation Cooling System ............................ 2-42 2.3.2.3 High-Pressure Core Spray System ................................... 2-42 2.3.2.4 Low-Pressure Core Spray System .................................... 2-43 2.3.2.5 Standby Gas Treatment System ....................................... 2-44

2.3.3 Auxiliary Systems ............................................................................. 2-44 2.3.3.1 Circulating Water System ................................................. 2-47 2.3.3.2 Condensate Processing Radioactive

(Demineralizer) System .................................................... 2-48 2.3.3.3 Containment Atmosphere Control System ........................ 2-48 2.3.3.4 Containment Exhaust Purge and Containment

Supply Purge Systems ..................................................... 2-48 2.3.3.5 Containment Instrument Air System ................................. 2-49 2.3.3.6 Containment Monitoring System ....................................... 2-50 2.3.3.7 Containment Nitrogen System .......................................... 2-51 2.3.3.8 Containment Return Air System ....................................... 2-51 2.3.3.9 Containment Vacuum Breaker System ............................. 2-52 2.3.3.10 Control Air System............................................................ 2-52 2.3.3.11 Control Rod Drive System ................................................ 2-53 2.3.3.12 Control Room Chilled Water System ................................ 2-54 2.3.3.13 Demineralized Water System ........................................... 2-54 2.3.3.14 Diesel Building Heating, Ventilation, and

Air Conditioning Systems.................................................. 2-55 2.3.3.15 Diesel Cooling Water System ........................................... 2-55 2.3.3.16 Diesel (Engine) Exhaust System ...................................... 2-56 2.3.3.17 Diesel Engine Starting Air System .................................... 2-57 2.3.3.18 Diesel Fuel Oil System ..................................................... 2-58 2.3.3.19 Diesel Generator System.................................................. 2-58 2.3.3.20 Diesel Lubricating Oil System ........................................... 2-59 2.3.3.21 Equipment Drains Radioactive System ............................. 2-60 2.3.3.22 Fire Protection System ..................................................... 2-61 2.3.3.23 Floor Drain System ........................................................... 2-69 2.3.3.24 Floor Drain Radioactive System ....................................... 2-70 2.3.3.25 Fuel Pool Cooling System ................................................ 2-71 2.3.3.26 Leak Detection System ..................................................... 2-72 2.3.3.27 Miscellaneous Waste Radioactive System ....................... 2-72 2.3.3.28 Plant Sanitary Drains Systems ......................................... 2-73

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2.3.3.29 Plant Service Water Systems ........................................... 2-73 2.3.3.30 Potable Cold Water System .............................................. 2-74 2.3.3.31 Potable Hot Water System ............................................... 2-75 2.3.3.32 Primary Containment System ........................................... 2-75 2.3.3.33 Process Sampling System ................................................ 2-75 2.3.3.34 Process Sampling Radioactive System ............................ 2-76 2.3.3.35 Pump House Heating, Ventilation, and

Air Conditioning Systems.................................................. 2-77 2.3.3.36 Radwaste Building Chilled Water System ......................... 2-77 2.3.3.37 Radwaste Building Heating, Ventilation, and

Air Conditioning Systems.................................................. 2-78 2.3.3.38 Reactor Building Heating, Ventilation, and

Air Conditioning Systems.................................................. 2-78 2.3.3.39 Reactor Closed Cooling Water System............................. 2-79 2.3.3.40 Reactor Protection System ............................................... 2-80 2.3.3.41 Reactor Water Cleanup System ....................................... 2-80 2.3.3.42 Service Air System ........................................................... 2-82 2.3.3.43 Standby Liquid Control System......................................... 2-82 2.3.3.44 Standby Service Water System ........................................ 2-83 2.3.3.45 Suppression Pool Temperature Monitoring System .......... 2-84 2.3.3.46 Tower Makeup Water System .......................................... 2-84 2.3.3.47 Traversing Incore Probe System ...................................... 2-85 2.3.3.48 Heating Steam System ..................................................... 2-85 2.3.3.49 Heating Steam Condensate System ................................. 2-86 2.3.3.50 Heating Steam Vent System ............................................. 2-86

2.3.4 Steam and Power Conversion Systems ........................................... 2-87 2.3.4.1 Auxiliary Steam System .................................................... 2-87 2.3.4.2 Condensate (Auxiliary) System......................................... 2-88 2.3.4.3 Condensate (Nuclear) System .......................................... 2-88 2.3.4.4 Main Steam System ......................................................... 2-89 2.3.4.5 Main Steam Leakage Control System............................... 2-90 2.3.4.6 Miscellaneous Drain System ............................................ 2-91 2.3.4.7 Reactor Feedwater System .............................................. 2-91 2.3.4.8 Sealing Steam System ..................................................... 2-91

2.4 Scoping and Screening Results: Structures ................................................... 2-92 2.4.1 Primary Containment ........................................................................ 2-93

2.4.1.1 Summary of Technical Information in the Application ....... 2-93 2.4.1.2 Conclusion ....................................................................... 2-94

2.4.2 Reactor Building ............................................................................... 2-94 2.4.2.1 Summary of Technical Information in the Application ....... 2-94 2.4.2.2 Staff Evaluation ................................................................ 2-95 2.4.2.3 Conclusion ....................................................................... 2-95

2.4.3 Standby Service Water Pump House 1A and 1B and Spray Pond 1A and 1B ..................................................................... 2-96 2.4.3.1 Summary of Technical Information in the Application ....... 2-96 2.4.3.2 Staff Evaluation ................................................................ 2-96 2.4.3.3 Conclusion ....................................................................... 2-97

2.4.4 Circulating Water Pump House ........................................................ 2-97 2.4.4.1 Summary of Technical Information in the Application ....... 2-97 2.4.4.2 Conclusion ....................................................................... 2-97

2.4.5 Diesel Generator Building ................................................................. 2-98

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2.4.5.1 Summary of Technical Information in the Application ....... 2-98 2.4.5.2 Staff Evaluation ................................................................ 2-98 2.4.5.3 Conclusion ....................................................................... 2-98

2.4.6 Fresh Air Intake Structures 1 and 2 .................................................. 2-99 2.4.6.1 Summary of Technical Information in the Application ....... 2-99 2.4.6.2 Conclusion ....................................................................... 2-99

2.4.7 Makeup Water Pump House............................................................. 2-99 2.4.7.1 Summary of Technical Information in the Application ....... 2-99 2.4.7.2 Conclusion ..................................................................... 2-100

2.4.8 Radwaste Control Building ............................................................. 2-100 2.4.8.1 Summary of Technical Information in the Application ..... 2-100 2.4.8.2 Staff Evaluation .............................................................. 2-101 2.4.8.3 Conclusion ..................................................................... 2-101

2.4.9 Service Building .............................................................................. 2-102 2.4.9.1 Summary of Technical Information in the Application ..... 2-102 2.4.9.2 Conclusion ..................................................................... 2-102

2.4.10 Turbine Generator Building ............................................................ 2-102 2.4.10.1 Summary of Technical Information in the Application ..... 2-102 2.4.10.2 Conclusion ..................................................................... 2-103

2.4.11 Water Filtration Building ................................................................. 2-103 2.4.11.1 Summary of Technical Information in the Application ..... 2-103 2.4.11.2 Conclusion ..................................................................... 2-103

2.4.12 Yard Structures .............................................................................. 2-103 2.4.12.1 Summary of Technical Information in the Application ..... 2-103 2.4.12.2 Conclusion ..................................................................... 2-105

2.4.13 Bulk Commodities .......................................................................... 2-105 2.4.13.1 Summary of Technical Information in the Application ..... 2-105 2.4.13.2 Staff Evaluation .............................................................. 2-105 2.4.13.3 Conclusion ..................................................................... 2-107

2.5 Scoping and Screening Results: Electrical and Instrumentation and Controls Systems .......................................................................................... 2-107 2.5.1 Electrical and Instrumentation and Controls Component

Commodity Groups ........................................................................ 2-108 2.5.1.1 Summary of Technical Information in the Application ..... 2-108 2.5.1.2 Staff Evaluation .............................................................. 2-108 2.5.1.3 Conclusion ..................................................................... 2-109

2.6 Conclusion for Scoping and Screening ......................................................... 2-109

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 .................................................................3-4 3.0.2.2 Review of AMR Results ......................................................3-5 3.0.2.3 Updated Final Safety Analysis Report Supplement.............3-6 3.0.2.4 Documentation and Documents Reviewed .........................3-6

3.0.3 Aging Management Programs ............................................................3-6 3.0.3.1 AMPs that are Consistent with the GALL Report .............. 3-10

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3.0.3.2 AMPs that are Consistent with the GALL Report with Exceptions or Enhancements ......................................... 3-118

3.0.3.3 AMPs that are Not Consistent with or Not Addressed in the GALL Report ................................ 3-212

3.0.4 Quality Assurance Program Attributes Integral to Aging Management Programs ........................................................ 3-274 3.0.4.1 Summary of Technical Information in Application ........... 3-274 3.0.4.2 Staff Evaluation .............................................................. 3-274 3.0.4.3 Conclusion ..................................................................... 3-275

3.0.5 Operating Experience for Aging Management Programs ................ 3-276 3.0.5.1 Summary of Technical Information in Application ........... 3-276 3.0.5.2 Staff Evaluation .............................................................. 3-276 3.0.5.3 UFSAR Supplement ....................................................... 3-285 3.0.5.4 Conclusion ..................................................................... 3-287

3.1 Aging Management of Reactor Vessel, Internals, and Reactor Coolant Systems ........................................................................................................ 3-287 3.1.1 Summary of Technical Information in the Application ..................... 3-288 3.1.2 Staff Evaluation .............................................................................. 3-288

3.1.2.1 AMR Results that are Consistent with the GALL Report .................................................................. 3-304

3.1.2.2 AMR Results that are Consistent with the GALL Report, for which Further Evaluation is Recommended ............... 3-312

3.1.2.3 AMR Results that are Not Consistent with or Not Addressed in the GALL Report ................................ 3-327

3.1.3 Conclusion ..................................................................................... 3-339 3.2 Aging Management of Engineered Safety Features ...................................... 3-339

3.2.1 Summary of Technical Information in the Application ..................... 3-340 3.2.2 Staff Evaluation .............................................................................. 3-340




3.2.3 Conclusion ..................................................................................... 3-369 3.3 Aging Management of Auxiliary Systems ...................................................... 3-369





3.3.3 Conclusion ..................................................................................... 3-488 3.4 Aging Management of Steam and Power Conversion Systems ..................... 3-488



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3.4.2.2 AMR Results that are Consistent with the GALL Report, for Which Further Evaluation is Recommended .............. 3-502


3.4.3 Conclusion ..................................................................................... 3-520 3.5 Aging Management of Containments, Structures, and

Component Supports .................................................................................... 3-520 3.5.1 Summary of Technical Information in the Application ..................... 3-520 3.5.2 Staff Evaluation .............................................................................. 3-521




3.5.3 Conclusion ..................................................................................... 3-576 3.6 Aging Management of Electrical and Instrumentation and Control ................ 3-576





3.6.3 Conclusion ..................................................................................... 3-591 3.7 Conclusion for AMR Results ......................................................................... 3-591

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 .........................4-1 4.1.2 Staff Evaluation of the Applicant’s Identification of TLAAs ..................4-1

4.1.2.1 Neutron Fluence .................................................................4-3 4.1.2.2 Flow-Induced Vibration Endurance Limit for the Reactor

Vessel Internals ..................................................................4-3 4.1.2.3 Ductility Reduction of Fracture Toughness for the

Reactor Vessel Internals ....................................................4-4 4.1.2.4 Leak-Before-Break Analysis ...............................................4-4 4.1.2.5 Concrete Containment Tendon Pre-stress Analysis............4-4 4.1.2.6 Fatigue Analysis of Containment Liner Plate ......................4-5 4.1.2.7 Intergranular Separation in the Heat-Affect-Zone

(HAZ) of Reactor Vessel Low-Alloy Steel under Austenitic Stainless Steel Cladding ....................................4-5

4.1.2.8 Low-Temperature Overpressure Protection Analyses .........4-6 4.1.2.9 Fatigue Analysis of Reactor Coolant Pump Flywheel ..........4-6 4.1.2.10 Fatigue Analysis of Polar Crane .........................................4-6 4.1.2.11 Metal Corrosion Allowances ...............................................4-7 4.1.2.12 Inservice Local Metal Containment Corrosion Analyses .....4-7 4.1.2.13 TLAAs related to BWRVIP Report

Applicant Action Items (AAIs) .............................................4-8

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4.1.3 Staff Evaluation of the Applicant’s Identification of Those Exemptions in the CLB That Are Based on TLAAs ........................... 4-13

4.1.4 Conclusion ....................................................................................... 4-15 4.2 Reactor Vessel Neutron Embrittlement ........................................................... 4-15

4.2.1 Neutron Fluence Values ................................................................... 4-17 4.2.1.1 Summary of Technical Information in the Application ....... 4-17 4.2.1.2 Staff Evaluation ................................................................ 4-18 4.2.1.3 UFSAR Supplement ......................................................... 4-22 4.2.1.4 Conclusion ....................................................................... 4-22

4.2.2 Upper-Shelf Energy .......................................................................... 4-23 4.2.2.1 Summary of Technical Information in the Application ....... 4-23 4.2.2.2 Staff Evaluation ................................................................ 4-24 4.2.2.3 UFSAR Supplement ......................................................... 4-33 4.2.2.4 Conclusion ....................................................................... 4-33

4.2.3 Adjusted Reference Temperature ..................................................... 4-33 4.2.3.1 Summary of Technical Information in the Application ....... 4-33 4.2.3.2 Staff Evaluation ................................................................ 4-34 4.2.3.3 UFSAR Supplement ......................................................... 4-39 4.2.3.4 Conclusion ....................................................................... 4-39

4.2.4 Pressure-Temperature Limits ........................................................... 4-39 4.2.4.1 Summary of Technical Information in the Application ....... 4-39 4.2.4.2 Staff Evaluation ................................................................ 4-40 4.2.4.3 UFSAR Supplement ......................................................... 4-42 4.2.4.4 Conclusion ....................................................................... 4-43

4.2.5 Reactor Vessel Circumferential Weld Examination Relief ................. 4-43 4.2.5.1 Summary of Technical Information in the Application ....... 4-43 4.2.5.2 Staff Evaluation ................................................................ 4-44 4.2.5.3 UFSAR Supplement ......................................................... 4-47 4.2.5.4 Conclusion ....................................................................... 4-47

4.2.6 Reactor Vessel Axial Weld Failure Probability .................................. 4-47 4.2.6.1 Summary of Technical Information in the Application ....... 4-47 4.2.6.2 Staff Evaluation ................................................................ 4-47 4.2.6.3 UFSAR Supplement ......................................................... 4-49 4.2.6.4 Conclusion ....................................................................... 4-49

4.3 Metal Fatigue .................................................................................................. 4-49 4.3.1 Reactor Pressure Vessel Fatigue Analyses ...................................... 4-50

4.3.1.1 Summary of Technical Information in the Application ....... 4-50 4.3.1.2 Staff Evaluation ................................................................ 4-50 4.3.1.3 UFSAR Supplement ......................................................... 4-53 4.3.1.4 Conclusion ....................................................................... 4-53

4.3.2 Reactor Pressure Vessel Internals ................................................... 4-54 4.3.2.1 Summary of Technical Information in the Application ....... 4-54 4.3.2.2 Staff Evaluation ................................................................ 4-55 4.3.2.3 UFSAR Supplement ......................................................... 4-56 4.3.2.4 Conclusion ....................................................................... 4-57

4.3.3 Reactor Coolant Pressure Boundary Piping and Component Fatigue Analyses ........................................................... 4-57 4.3.3.1 Summary of Technical Information in the Application ....... 4-57 4.3.3.2 Staff Evaluation ................................................................ 4-58 4.3.3.3 UFSAR Supplement ......................................................... 4-59 4.3.3.4 Conclusion ....................................................................... 4-59

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4.3.4 Non-Class 1 Component Fatigue Analyses ...................................... 4-59 4.3.4.1 Summary of Technical Information in the Application ....... 4-59 4.3.4.2 Staff Evaluation ................................................................ 4-60 4.3.4.3 UFSAR Supplement ......................................................... 4-60 4.3.4.4 Conclusion ....................................................................... 4-61

4.3.5 Effects of Reactor Coolant Environment on Fatigue Life of Components and Piping ................................................................... 4-61 4.3.5.1 Summary of Technical Information in the Application ....... 4-61 4.3.5.2 Staff Evaluation ................................................................ 4-62 4.3.5.3 UFSAR Supplement ......................................................... 4-75 4.3.5.4 Conclusion ....................................................................... 4-75

4.4 Environmental Qualification (EQ) of Electrical Equipment ............................... 4-75 4.4.1 Summary of Technical Information in the Application ....................... 4-76 4.4.2 Staff Evaluation ................................................................................ 4-76 4.4.3 UFSAR Supplement ......................................................................... 4-77 4.4.4 Conclusion ....................................................................................... 4-77

4.5 Loss of Prestress in Concrete Containment Tendons ..................................... 4-77 4.5.1 Summary of Technical Information in the Application ....................... 4-77 4.5.2 Staff Evaluation ................................................................................ 4-77 4.5.3 UFSAR Supplement ......................................................................... 4-77 4.5.4 Conclusion ....................................................................................... 4-77

4.6 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analyses ......................................................................................................... 4-78 4.6.1 ASME Class MC Components .......................................................... 4-79

4.6.1.1 Summary of Technical Information in the Application ....... 4-79 4.6.1.2 Staff Evaluation ................................................................ 4-80 4.6.1.3 UFSAR Supplement ......................................................... 4-80 4.6.1.4 Conclusion ....................................................................... 4-80

4.6.2 Downcomers .................................................................................... 4-81 4.6.2.1 Summary of Technical Information in the Application ....... 4-81 4.6.2.2 Staff Evaluation ................................................................ 4-81 4.6.2.3 UFSAR Supplement ......................................................... 4-81 4.6.2.4 Conclusion ....................................................................... 4-82

4.6.3 SRV Discharge Piping ...................................................................... 4-82 4.6.3.1 Summary of Technical Information in the Application ....... 4-82 4.6.3.2 Staff Evaluation ................................................................ 4-82 4.6.3.3 UFSAR Supplement ......................................................... 4-83 4.6.3.4 Conclusion ....................................................................... 4-83

4.6.4 Diaphragm Floor Seal....................................................................... 4-83 4.6.4.1 Summary of Technical Information in the Application ....... 4-83 4.6.4.2 Staff Evaluation ................................................................ 4-83 4.6.4.3 UFSAR Supplement ......................................................... 4-84 4.6.4.4 Conclusion ....................................................................... 4-84

4.6.5 Emergency Core Cooling System Suction Strainers ......................... 4-84 4.6.5.1 Summary of Technical Information in the Application ....... 4-84 4.6.5.2 Staff Evaluation ................................................................ 4-84 4.6.5.3 UFSAR Supplement ......................................................... 4-85 4.6.5.4 Conclusion ....................................................................... 4-85

4.7 Other Plant-Specific Time Limited Aging Analyses .......................................... 4-85 4.7.1 Reactor Vessel Shell Indication ........................................................ 4-85

4.7.1.1 Summary of Technical Information in the Application ....... 4-85

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4.7.1.2 Staff Evaluation ................................................................ 4-86 4.7.1.3 UFSAR Supplement ......................................................... 4-92 4.7.1.4 Conclusion ....................................................................... 4-92

4.7.2 Sacrificial Shield Wall ....................................................................... 4-93 4.7.2.1 Summary of Technical Information in the Application ....... 4-93 4.7.2.2 Staff Evaluation ................................................................ 4-93 4.7.2.3 UFSAR Supplement ......................................................... 4-93 4.7.2.4 Conclusion ....................................................................... 4-93

4.7.3 Main Steam Line Flow Restrictor Erosion Analyses .......................... 4-94 4.7.3.1 Summary of Technical Information in the Application ....... 4-94 4.7.3.2 Staff Evaluation ................................................................ 4-94 4.7.3.3 UFSAR Supplement ......................................................... 4-95 4.7.3.4 Conclusion ....................................................................... 4-95

4.7.4 Core Plate Rim Hold-Down Bolts ...................................................... 4-96 4.7.4.1 Summary of Technical Information in the Application ....... 4-96 4.7.4.2 Staff Evaluation ................................................................ 4-96 4.7.4.3 UFSAR Supplement ......................................................... 4-98 4.7.4.4 Conclusion ....................................................................... 4-98

4.7.5 Crane Load Cycle Limit .................................................................... 4-99 4.7.5.1 Summary of Technical Information in the Application ....... 4-99 4.7.5.2 Staff Evaluation ................................................................ 4-99 4.7.5.3 UFSAR Supplement ....................................................... 4-102 4.7.5.4 Conclusion ..................................................................... 4-102

4.8 Conclusion for Time Limited Aging Analyses ................................................ 4-102

SECTION 5 REVIEW BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ....................................................................................................5-1

SECTION 6 CONCLUSION .....................................................................................................6-1

APPENDIX A COLUMBIA GENERATING STATION LICENSE RENEWAL COMMITMENTS .............................................................................................. A-1

APPENDIX B CHRONOLOGY ............................................................................................... B-1

APPENDIX C PRINCIPAL CONTRIBUTORS......................................................................... C-1

APPENDIX D REFERENCES ................................................................................................. D-1

xiv

LIST OF TABLES

Table 2.2-1 UFSAR Systems ................................................................................................ 2-35 Table 2.3-1 Continuation Issue for License Renewal Drawings ............................................. 2-46 Table 3.0-1 Columbia AMPs ...................................................................................................3-6 Table 3.1-1 Staff evaluation for RV, RVI, and RCS components in the GALL Report .......... 3-289 Table 3.2-1 Staff evaluation for ESF system components in the GALL Report .................... 3-341 Table 3.3-1 Staff evaluation for auxiliary system components in the GALL Report .............. 3-371 Table 3.4-1 Staff evaluation for steam and power conversion system components

in the GALL Report ......................................................................................... 3-490 Table 3.5-1 Staff evaluation for containments, structures, and component supports

in the GALL Report ......................................................................................... 3-522 Table 3.6-1 Staff evaluation for electrical and I&Cs in the GALL Report .............................. 3-577

xv

ABBREVIATIONS

AAI applicant action item AC alternating current ACI American Concrete Institute ACRS Advisory Committee on Reactor Safeguards ACSR aluminum conductor steel reinforced ADAMS Agencywide Documents Access and Management System ADS automatic depressurization system AERM aging effect requiring management AMP aging management program AMR aging management review ANSI American National Standards Institute AQ augmented quality AR Action Request ART adjusted reference temperature ASME American Society of Mechanical Engineers ASTM American Standards for Testing and Materials ATWS anticipated transient without scram

B&PV boiler and pressure vessel B4C boron carbide BADGER Boron-10 Areal Density Gage for Evaluating Racks BTP Branch Technical Position BWR boiling-water reactor BWRVIP Boiling-Water Reactor Vessel and Internals Project

C Celsius CAS control air system CASS cast austenitic stainless steel CB&I Chicago Bridge and Iron CCH control room chilled water CEA control element assembly CEP containment exhaust purge CF chemistry factor CFR Code of Federal Regulations CI confirmatory item CIA containment instrument air CLB current licensing basis CMS containment monitoring system CN containment nitrogen

Abbreviations

xvi

CO condensate (auxiliary) CO2 carbon dioxide Columbia Columbia Generating Station CPR condensate processing radioactive CR condition report CRA containment return air CRD control rod drive CRDRL control rod drive return line CSP containment supply purge CSR cable spreading room CST condensate storage tank CUF cumulative usage factor CVB containment vacuum breaker CVN Charpy-V notch CW circulating water

DBA design basis accident DBE design basis event DCW diesel cooling water DE diesel (engine) exhaust DEH digital electro-hydraulic control system DG diesel generator DLO diesel lubricating oil DO dissolved oxygen DOT Department of Transportation ∆RTNDT reference nil-ductility temperature caused by irradiation DSA diesel starting air DW demineralized water

EAF environmentally-assisted fatigue ECCS emergency core cooling system EDR equipment drain radioactive EFPY effective full power years EMA equivalent margin analysis EPRI Electrical Power Research Institute EQ environmental qualification ESF engineered safety feature

F Fahrenheit FD floor drain FDR floor drain radioactive

Abbreviations

xvii

Fen environmental life correction factors FERC Federal Energy Regulatory Commission FIV flow-induced vibrators FPC fuel pool cooling FR Federal Register FSAR final safety analysis report ft foot ft2 square foot FW feedwater

GALL generic aging lessons learned GE General Electric GEIS generic environmental impact statement GL Generic Letter

HAZ heat-affected zone HCO heating steam condensate HELB high-energy line break HPCS high-pressure core spray HS heating steam HSV heating steam vent HVAC heating, ventilation, and air conditioning HWC hydrogen water chemistry

I&C instrumentation and control IASCC irradiation-assisted stress-corrosion cracking ID inside diameter IGA intergranular attack IGSCC intergranular stress-corrosion cracking IN Information Notice in. inch INPO Institute of Nuclear Plant Operation IPA integrated plant assessment IR infrared ISG interim staff guidance ISI inservice inspection ISP Integrated Surveillance Program

Ki applied stress intensity Kic fracture toughness based on crack initiation kV kilovolt

Abbreviations

xviii

lb pound LBB leak-before-break LD leak detection LER licensee event report LOCA loss of coolant accident LPCI low pressure coolant injection LPCS low-pressure core spray LRA license renewal application LRIC License Renewal Implementation Coordinator LRP leak reduction program LTR licensing topical report LWR light-water reactor

M margin term MEB metal-enclosed bus MEL master equipment list MIC microbiologically-influenced corrosion MRSM maintenance rule scoping matrix MS main steam MSIV main steam isolation valve MSLC main steam leakage control MWR miscellaneous waste radioactive MWt megawatt thermal

NACE National Association of Corrosion Engineers NDE non-destructive examination NEI Nuclear Energy Institute NESC National Electrical Safety Code NFPA National Fire Protection Association NMCA noble metal chemical application NPS nominal pipe size NRC U.S. Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NSAS nonsafety affecting safety NSSS nuclear steam supply system nvt measure of fluence in n/cm2 NWC normal water chemistry

OBE operating-basis earthquake OD outside diameter OEM original equipment manufacturer

Abbreviations

xix

OI open item OQAPD operational quality assurance program description OTSG once-through steam generator

P&ID piping and instrumentation drawing PDI performance demonstrative initiative PER problem evaluation request PFSS post-fire safe shutdown PGCC Power Generation Control Cabinet PM preventive maintenance ppb parts per billion ppm parts per million PS process sampling PSD plant sanitary drain PSR process sampling radioactive P-T pressure-temperature PVC polyvinyl chloride PWC potable cold water PWH potable hot water PWR pressurized-water reactor PWSCC primary water stress-corrosion cracking

QA quality assurance QIDs qualification information documents

RAI request for additional information RCC reactor closed cooling water RCIC reactor core isolation cooling RCPB reactor coolant pressure boundary RCS reactor coolant system RFO refueling outage RFW reactor feedwater RG regulatory guide RH relative humidity RHR residual heat removal RI-ISI risk-informed inservice inspection RP regulatory position RPS reactor protection system RPV reactor pressure vessel RRC reactor recirculation RTNDT reference nil-ductility temperature

Abbreviations

xx

RV reactor vessel RVI reactor vessel internals RVID Reactor Vessel Integrity Database RWCU reactor water cleanup

SA service air SBO station blackout SC structure and component SCC stress-corrosion cracking SDV scram discharge volume SER safety evaluation report SFA steam/feedwater application SFP spent fuel pool SGT standby gas treatment SLC standby liquid control SPTM suppression pool temperature monitoring

SRP-LR Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants

SRRF stress range reduction factor SRV safety relief valve SSC structure, system, and component SSE safe shutdown earthquake SSP Supplemental Surveillance Program SSW standby service water

TIP traversing incore probe TLAA time-limited aging analysis TMU tower makeup water TS technical specification TSW plant service water

UFSAR updated final safety analysis report USE upper-shelf energy UT ultrasonic testing UV ultraviolet

V volt VAC volts alternating current VIP Vessel Internals Program WCH radwaste building chilled water Zn zinc

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SECTION 1

INTRODUCTION AND GENERAL DISCUSSION

1.1

This document is a safety evaluation report (SER) on the license renewal application (LRA) for Columbia Generating Station (Columbia), as filed by Energy Northwest (the applicant). By letter dated January 19, 2010, Energy Northwest submitted its application to the U.S. Nuclear Regulatory Commission (NRC) for renewal of the Columbia operating license for an additional 20 years. The NRC staff (the staff) prepared this report to summarize the results of its safety review of the LRA for compliance with Title 10, Part 54, “Requirements for Renewal of Operating Licenses for Nuclear Power Plants,” of the Code of Federal Regulations (10 CFR Part 54). The NRC project manager for the license renewal review is Arthur D. Cunanan. Mr. Cunanan may be contacted by telephone at 301-415-3897 or by electronic mail at

Introduction

[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: Arthur D. Cunanan, Mail Stop 011-F1

In its January 19, 2010, submission letter, the applicant requested renewal of the operating license issued under Section 103 (Operating License No. NPF-21) of the Atomic Energy Act of 1954, as amended, for a period of 20 years beyond the current license date of December 20, 2023. Columbia is located approximately 12 miles north of Richland, WA. The NRC issued the construction permit on March 19, 1973. The NRC issued the operating license for Columbia on April 13, 1984. The unit is a Mark II boiling-water reactor (BWR) design. General Electric Company supplied the nuclear steam supply system; Burns and Roe, Inc., designed the balance of plant; and Bechtel Power Corporation constructed the plant. The licensed power output of the unit is 3,886 megawatt thermal with a gross electrical output of approximately 1,230 megawatt electric. 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 Columbia license renewal is based on the applicant’s LRA and on its 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 January 4, 2012. The staff may consider information received after that date depending on the progress 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, (301-415-4737/800-397-4209). The LRA may also be viewed at the Richland Public Library, 955 Northgate Drive, Richland, WA 99352 and at the Kennewick Branch of Mid-Columbia

mailto:[email protected]�

Introduction and General Discussion

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Libraries, 1620 South Union Street, Kennewick, WA 99338. In addition, the public may find the LRA, as well as materials related to the license renewal review, on the NRC Web site 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 the unit’s proposed operation for an additional 20 years beyond the term of the current operating license. The staff reviewed the LRA in accordance with NRC regulations and the guidance in NUREG-1800, Revision 1, “Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants” (SRP-LR), dated September 2005.

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 conclusions of this SER are in Section 6.

SER Appendix A is a table showing the applicant’s commitments for renewal of the operating license. 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, and as part of the environmental review, the staff is also preparing a draft plant-specific supplement to NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear Plants (GEIS).” Issued separately from this SER, this supplement will discuss the environmental considerations for the license renewal of Columbia Generating Station.

1.2

Under 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 based on an expected 40-year service life.

License Renewal Background

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 and implement a comprehensive program plan for nuclear plant aging research. From the results of the nuclear plant aging 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 (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 period of the initial license and that the scope of the review did not allow sufficient credit for management programs, 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. Published on May 8, 1995, in Volume 60, page 22461, of the Federal Register (60 FR 22461),

http://www.nrc.gov/�


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the amended 10 CFR Part 54 establishes a regulatory process that is more stable and predictable 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 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 (Volume 61, page 28467, of the Federal Register (61 FR 28467)), dated June 5, 1996, and amended 10 CFR Part 51 to focus the scope of the review of environmental impacts of license renewal in order to fulfill NRC responsibilities under the National Environmental Policy Act of 1969 (NEPA).

1.2.1 Safety Review

License renewal requirements for power reactors are based on two key principles:

• 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 exception of the detrimental aging effects on the function of certain SSCs as well as a few other safety-related issues, during the period of extended operation.

• 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 defines the scope of license renewal as including SSCs that are safety-related, whose failure could affect safety-related functions, or that are relied on to demonstrate compliance with NRC 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 are those that perform an intended function without moving parts or without a change in configuration or properties (i.e., are “passive”), and are not subject to replacement based on a qualified life or specified time period (i.e., are “long lived”). As required by 10 CFR 54.21(a), an applicant for a renewed license must demonstrate that aging effects will be managed in such a way that the intended function(s) of those SSCs will be maintained, consistent with the current licensing basis (CLB), for the period of extended operation; however, active equipment is considered adequately monitored and maintained by existing programs. In other words, detrimental aging effects that may affect active equipment are readily detectable and can be 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.

Pursuant to 10 CFR 54.21(d), each LRA is required to include a UFSAR supplement that must have a summary description of the applicant’s programs and activities for managing aging effects and the evaluation of time-limited aging analyses (TLAAs) for the period of extended operation.


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License renewal also requires TLAA identification and updating. During the plant design phase, certain assumptions are made about the length of time the plant can operate. These assumptions are incorporated into design calculations for several plant SSCs. In accordance with 10 CFR 54.21(c)(1), the applicant must 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 effects of aging on these SSCs can be adequately managed for the period of extended operation.

In 2005, the staff developed and issued 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 by NEI. NEI 95-10 details an acceptable method of implementing the Rule. The staff also used the SRP-LR to review this application.

In its LRA, the applicant stated that it used the process defined in NUREG-1801, “Generic Aging Lessons Learned (GALL) Report,” dated September 2005. The GALL Report provides a summary of staff-approved aging management programs (AMPs) for the aging of many SCs subject to an AMR. An applicant’s willingness to commit to carrying out these staff-approved AMPs could potentially reduce the time, effort, and resources in reviewing an applicant’s LRA and, thereby, improve the efficiency and effectiveness of the license renewal review process. The report is also a reference for both applicants and staff reviewers to quickly identify AMPs and activities that can provide adequate aging management during the period of extended operation. It is incumbent on the applicant to ensure that the conditions and operating experience at the plant are bounded by the conditions and operating experience for which the GALL Report was evaluated. If these bounding conditions are not met, the applicant should address the additional effects of aging and augment its aging management program (AMP) as appropriate.

1.2.2 Environmental Review

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 the possible environmental impacts associated with renewing licenses of nuclear power plants. For certain types of environmental impacts, the GEIS establishes generic findings applicable to all nuclear power plants. These generic findings are codified in Appendix B to Subpart A of 10 CFR Part 51. Pursuant to 10 CFR 51.53(c)(3)(i), an applicant for license renewal may incorporate these generic findings in its environmental report. In accordance with 10 CFR 51.53(c)(3)(ii), an environmental report must also include analyses of environmental impacts that must be evaluated on a plant-specific basis (i.e., Category 2 issues).

In accordance with NEPA and the requirements of 10 CFR Part 51, the staff performed a plant-specific review of the environmental impacts of license renewal, including whether or not the GEIS had considered new and significant information. As part of its scoping process, the staff held two public meetings on April 6, 2010, at the Richland Public Library in Richland, WA, to identify plant-specific environmental issues that might impact Columbia. The staff issued the draft site-specific GEIS supplement on August 23, 2011. After considering comments on the draft, the staff will prepare and publish a final plant-specific GEIS supplement separately.


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1.3

Part 54 of 10 CFR describes the requirements for renewing operating licenses for nuclear power plants. The staff performed its technical review of the LRA in accordance with NRC guidance and 10 CFR Part 54 requirements. Section 54.29 of 10 CFR sets forth the standards for renewing a license. This SER describes the results of the staff’s safety review.

Principal Review Matters

Pursuant to 10 CFR 54.19(a), the NRC requires a license renewal applicant to submit general information. The applicant provided this general information in LRA Section 1, which it submitted by letter, dated January 19, 2010. The staff reviewed LRA Section 1 and found that the applicant had submitted the information required by 10 CFR 54.19(a).

Pursuant to 10 CFR 54.19(b), the staff requires that each 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.” The applicant stated the following in LRA Section 1.1.8 on this issue:

Per 10 CFR 54.19(b), license renewal applications are required to 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 (No. B-94) for Columbia states, in Article VII, that the agreement shall terminate at the time of expiration of the license specified in Item 3 of the Attachment (to the agreement). Item 3 of the Attachment to the indemnity agreement, as revised by Amendment No. 1, lists Columbia operating license NPF-21. Energy Northwest requests that conforming changes be made to Article VII of the indemnity agreement, and Item 3 of the Attachment to that agreement, specifying the extension of agreement to the expiration date of the renewed Columbia facility operating license sought in this application. In addition, should the license number be changed upon issuance of the renewal license, Energy Northwest requests that conforming changes be made to Item 3 of the Attachment to the indemnity agreement and to other sections of the agreement as deemed appropriate.

The staff intends to maintain the original license number upon issuance of the renewed license, if approved. Therefore, conforming changes to the indemnity agreement need not be made, and the 10 CFR 54.19(b) requirements have been met. Pursuant to 10 CFR 54.21, the staff requires that each LRA contain the following:

• an integrated plant assessment (IPA) • a description of any CLB changes during the staff’s review of the LRA • an evaluation of TLAAs • a UFSAR supplement

LRA Sections 3 and 4 and Appendix B address the license renewal requirements of 10 CFR 54.21(a), 10 CFR 54.21(b), and 10 CFR 54.21(c). LRA Appendix A satisfies the license renewal requirements of 10 CFR 54.21(d).

Pursuant to 10 CFR 54.21(b), the staff 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 of the facility that materially affect the contents of the LRA, including the UFSAR supplement.

Pursuant to 10 CFR 54.22, the staff requires that an applicant’s LRA include changes or additions to the technical specifications necessary to manage aging effects during the period of


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extended operation. In LRA Appendix D, the applicant stated that, “no Technical Specification Changes are necessary to manage the effects of aging during the period of extended operation.”

The staff evaluated the technical information required by 10 CFR 54.21 and 10 CFR 54.22 in accordance with NRC regulations and the guidance of the SRP-LR. SER Sections 2, 3, and 4 document the staff’s evaluation of the technical information in the LRA.

As required by 10 CFR 54.25, the ACRS will issue a report to document its evaluation of the staff’s LRA review and associated SER. SER Section 5 will incorporate the ACRS report once it is issued. SER Section 6 will document the findings required by 10 CFR 54.29.

1.4

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 NRC’s safety goal of ensuring adequate protection of public health and safety and the environment. 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.

Interim Staff Guidance

The GALL Report, Revision 2, dated December 2010, have incorporated all current and proposed ISGs up to that date.

1.5

As a result of its review of the LRA, including additional information submitted through January 4, 2012, the staff closed the six open items (OIs), previously identified in the "Safety Evaluation Report with Open Items Related to the License Renewal of Columbia Generating Station" (ADAMS Accession No. ML11349A022). .

Summary of the Open Items

OI 3.0.3.3.7

The staff noted that the applicant did not include the high-voltage station post insulators at the 230 kV ASHE A809 Breaker located in the Ashe Substation even though this breaker provides an alternate path of power during a station blackout. The applicant stated that it did not include these station post insulators because it concluded the spray drift phenomenon would not occur due to the significant distance from the circulating water system cooling towers.

SER Section 3.0.3.3.7— High-Voltage Porcelain Insulators

During discussions with the inspection team, the applicant provided a cooling tower drift study that identified that hard water deposits on the 500 kV breaker station post insulators had allowed arcing to occur. The study did not demonstrate that the 230 kV station post insulators would remain unaffected. Further, the applicant could not provide any other information to support its conclusions. Consequently, the applicant issued Action Requests 228661 and 228673 to resolve this concern. The applicant indicated that it would either establish appropriate coating or cleaning tasks or develop information that would demonstrate why the phenomenon would not affect the 230 kV switchyard station post insulators. In its August 11, 2011, response, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Section 3.0.3.3.7. Open item OI 3.0.3.3.7 is closed.


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OI B.1.4-1

The applicant did not fully describe how it will use future operating experience to ensure that the AMPs will remain effective for managing the aging effects during the period of extended operation. While the majority of the program descriptions contain statements indicating that future operating experience will be used to adjust the programs as appropriate, the details of this process are not fully described. In addition, some program descriptions contain no such statements and, for these AMPs, it is not clear whether the applicant intends to implement actions to monitor operating experience on an ongoing basis and use it to ensure the continued effectiveness of these AMPs. Further, the LRA does not state whether new AMPs will be developed, as necessary. In its December 16, 2011, response, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Section 3.0.5.2. Open item OI B.1.4-1 is closed.

SER Section 3.0.2.1 – Operating Experience

OI 4.2-1

The applicant provided the projected upper shelf energy (USE) to 54 effective full power years (EFPY) for the N12 nozzle forgings in its LRA supplement stating that the unirradiated (initial) transverse USE of 62 ft-lbs and copper content of 0.27 percent used in the calculation of projected USE for the N12 nozzles are based on the results of a statistical analysis of data by the original equipment manufacturer for SA-508 Class 1 forging material. The applicant states that the requirement for 50 ft-lbs minimum USE at the end of vessel life is met for the current license period and for the period of extended operation for the N12 nozzle forgings. Therefore, the applicant states that the USE will not drop below 50 ft-lbs prior to the period of extended operation.

SER Section 4.2.2 – Upper-Shelf Energy

The staff reviewed the applicant's response and had concerns that the applicant did not provide a technical basis on the unirradiated (initial) transverse USE of 62 ft-lbs and copper content of 0.27 percent used in the calculation of projected USE for the N12 nozzles. The applicant should provide this basis and demonstrate that the USE value for the N12 nozzle forgings will not pass below 50 ft-lbs prior to the period of extended operation for verification by the staff. In its November 1, 2011, response, as supplemented by information provided on December 6, 2011, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Section 4.2.2.2. Open item OI 4.2-1 is closed.

OI 4.3-1

The applicant stated that it addresses the effects of the coolant environment on component fatigue life by assessing the impact of the reactor coolant environment on a sample of critical components identified in NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," March 1995.

SER Section 4.3 – Metal Fatigue

However, the staff noted that the applicant's plant-specific configuration may contain additional locations (including, but not limited to, those provided in LRA Tables 4.3-3 and 4.3-5) that may need to be analyzed for the effects of the reactor coolant environment other than those identified in NUREG/CR-6260. This may include locations that are limiting or bounding for the applicant's particular plant-specific configuration or that have calculated environmentally-adjusted fatigue (EAF) values that are greater than those calculated EAF


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values by the applicant for locations that correspond to those identified in NUREG/CR-6260. In its December 16, 2011, response, as supplemented by information provided on January 4, 2012, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Section 4.3.5.2. Open item OI 4.3-1 is closed.

OI 4.7.4-1

The staff noted that the applicant had submitted a TLAA for the core plate rim hold-down bolts but had not selected one of the three options of 10 CFR 54.21(c)(1) to demonstrate its evaluation of the TLAA. Also, the applicant did not provide an AMR line item for the core plate rim hold-down bolts with the aging effect of loss of preload due to stress relaxation. Further, the applicant stated that it intended to deviate from BWRVIP-25 inspection guidelines, which could result in inadequate management of the aging effect. In its November 4, 2011, response, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Sections 3.0.3.1.6 and 4.7.4. Open item OI 4.7.4-1 is closed.

SER Section 4.7.4 – Core Plate Rim Hold-Down Bolts

OI 4.7.5-1

The applicant states that the analysis of the polar crane does not meet the definition of a TLAA. However, the staff believes the analysis of the polar crane does meet the definition of a TLAA because the polar crane has a design limit of cycles in the Crane Manufacturers Association of America (CMAA) 70 specification of 20,000 to 100,000 lifts, and an "assumed design assessment" of the number of lifts compared to the CMAA 70 specification. In its October 5, 2011, response, as supplemented by information provided on November 16, 2011, the applicant provided the additional information to address the staff's concern. The staff reviewed and accepted the applicant's response, as documented in SER Section 4.7.5. Open item OI 4.7.5-1 is closed.

SER Section 4.1.2.9 – Fatigue Analysis of Polar Crane

1.6

As a result of its review of the LRA, including additional information submitted through January 4, 2012, the staff identified no confirmatory items (CI).

Summary of Confirmatory Items

1.7

Following the staff’s review of the LRA, including subsequent information and clarifications provided by the applicant, the staff identified three additional proposed license conditions related to license renewal.

Summary of Proposed License Conditions

The first license condition states that the information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, is henceforth part of the UFSAR which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs and activities described in the supplement provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

The second license condition states that Appendix A of the Safety Evaluation Report Related to the License Renewal of Columbia and the licensee's UFSAR supplement submitted pursuant to


1-9

10 CFR 54.21(d), describes certain future programs and activities to be completed before the period of extended operation. Energy Northwest shall complete these activities no later than December 20, 2023, and shall notify the NRC in writing when implementation of these activities is complete.

The third license condition requires the applicant to install wedges on or before December 20, 2021, and to submit a report to NRC staff summarizing the results of the installation of wedges and, if applicable, corrective action.

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SECTION 2

STRUCTURES AND COMPONENTS SUBJECT TO AGING MANAGEMENT REVIEW

2.1

2.1.1 Introduction

Scoping and Screening Methodology

Title 10, Section 54.21, “Contents of Application—Technical Information,” of the Code of Federal Regulations (10 CFR 54.21), requires that each license renewal application (LRA) must contain an integrated plant assessment (IPA). The IPA must list and identify all of the structures, systems, and components (SSCs) within the scope of license renewal and all structures and components (SCs) subject to an aging management review (AMR), in accordance with 10 CFR 54.4.

LRA Section 2.1, “Scoping and Screening Methodology,” describes the scoping and screening methods used to identify the SSCs at the Columbia Generating Station (Columbia) that are within the scope of license renewal and the SCs that are subject to an AMR. The staff reviewed the scoping and screening methods applied by Energy Northwest (the applicant) to determine whether it meets the scoping requirements of 10 CFR 54.4(a) and the screening requirements of 10 CFR 54.21.

In developing the scoping and screening methods for the LRA, the applicant stated that it considered the following documents:

• 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,” (NEI 95-10)

• correspondence between the U.S. Nuclear Regulatory Commission (NRC), other applicants, and the NEI

2.1.2 Information Sources Used for Scoping and Screening

In LRA Sections 2 and 3 the applicant provided the technical information required by 10 CFR 54.4, “Scope,” and 10 CFR 54.21(a). This safety evaluation report (SER) with open items, contains sections entitled “Summary of Technical Information in the Application,” which provides informational summaries of technical information provided in the LRA.

In LRA Section 2.1, the applicant described the process used to identify the SSCs that meet the license renewal scoping criteria under 10 CFR 54.4(a) and the process used to identify the SCs that are subject to an AMR, as required by 10 CFR 54.21(a)(1). The applicant provided the results of the process used for identifying the SCs subject to an AMR in the following LRA Sections:

• LRA Section 2.2, “Plant-Level Scoping Results”

Structures and Components Subject to Aging Management Review

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• LRA Section 2.3, “Scoping and Screening Results: Mechanical Systems”

• LRA Section 2.4, “Scoping and Screening Results: Structures”

• LRA Section 2.5, “Scoping and Screening Results: Electrical and Instrumentation and Control Systems”

In LRA Section 3.0, “Aging Management Review Results,” the applicant described its aging management results as follows:

• LRA Section 3.1, “Aging Management of Reactor Vessel, Internals, and Reactor Coolant System”

• LRA Section 3.2, “Aging Management of Engineered Safety Features”

• LRA Section 3.3, “Aging Management of Auxiliary Systems”

• LRA Section 3.4, “Aging Management of Steam and Power Conversion Systems”

• LRA Section 3.5, “Aging Management of Containments, Structures, and Component Supports”

• LRA Section 3.6, “Aging Management of Electrical and Instrumentation and Control Systems”

In LRA Section 4.0, “Time-Limited Aging Analyses,” the applicant identified and described the evaluation of time-limited aging analyses (TLAAs).

2.1.3 Scoping and Screening Program Review

The staff evaluated the LRA scoping and screening methodology in accordance with the guidance contained in NUREG-1800, Revision 1, “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 form the basis for the acceptance criteria of the scoping and screening methodology review:

• 10 CFR 54.4(a), as it relates to the identification of plant 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)(1) and (a)(2), as they relate to the methods used by the applicant to identify plant SCs subject to an AMR

As part of the review of the applicant’s scoping and screening methodology, the staff used guidance contained in the SRP-LR and reviewed the activities described in the following sections of the 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)

• Section 2.2, to ensure that the applicant described a process for determining the SCs that are subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1) and (a)(2)

Structures and Components Subject to Aging Management Review

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In addition, the staff conducted a scoping and screening methodology audit at Columbia, during the week of May 10–13, 2010. The scoping and screening methodology audit focused on ensuring that the applicant had developed and implemented adequate guidance to carry out the scoping and screening of SSCs in accordance with the methods described in the LRA and the requirements of the Rule. First, the staff reviewed implementation of the project-level guidelines and topical reports describing the applicant’s scoping and screening methodology. Second, the staff conducted detailed discussions with the applicant on the implementation and control of the license renewal program. Third, the staff reviewed the administrative control documentation used by the applicant during the scoping and screening process. Fourth, the staff reviewed the quality practices used by the applicant to develop the LRA. Finally, the staff reviewed the training and qualification of the LRA development team.

The staff evaluated the quality attributes of the applicant’s aging management program (AMP) activities described in Appendix A, “Final Safety Analysis Report Supplement,” and Appendix B, “Aging Management Programs,” of the LRA. On a sampling basis, the staff performed a system review of the service water; emergency diesel generators (DGs) and support systems; and the turbine building, including a review of the scoping and screening results reports and supporting design documentation used to develop the reports. The purpose of the staff’s review was to ensure that the applicant had appropriately implemented the methodology outlined in the administrative controls and to verify that the results are consistent with the current licensing basis (CLB) documentation.

2.1.3.1 Implementing Procedures and Documentation Sources Used 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, dated August 19, 2010 (ADAMS Accession No. ML102160357), to verify that the