ML12340A710

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Official Exhibit - ENT000546-00-BD01 - NUREG-1958, Safety Evaluation Report Related to the License Renewal of Kewaunee Power Station (Jan. 2011) (Excerpted)
ML12340A710
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 01/31/2011
From:
Office of Nuclear Reactor Regulation
To:
Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 23330, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01
Download: ML12340A710 (26)
v • d • e


Text

United States Nuclear Regulatory Commission Official Hearing Exhibit Entergy Nuclear Operations, Inc.

In the Matter of:

(Indian Point Nuclear Generating Units 2 and 3)

ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: ENT000546-00-BD01 Identified: 10/15/2012 Admitted: 10/15/2012 Withdrawn: ENT000546 Rejected: Stricken: Submitted: August 20, 2012 Other:

NUREG-1958 Safety Evaluation Report Related to the License Renewal of Kewaunee Power Station Docket No. 50-305 Dominion Energy Kewaunee, Inc.

Office of Nuclear Reactor Regulation

NUREG-1958 Safety Evaluation Report Related to the License Renewal of Kewaunee Power Station Docket No. 50-305 Dominion Energy Kewaunee, Inc.

Manuscript Completed: January 2011 Date Published: January 2011 Office of Nuclear Reactor Regulation

ABSTRACT This safety evaluation report (SER) documents the technical review of the Kewaunee Power Station (KPS) license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff (the staff). By letter dated August 12, 2008, Dominion Energy Kewaunee, Inc.

(Dominion, DEK, or 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. Dominion requests renewal of the KPS operating license (Facility Operating License Number DPR-43) for a period of 20 years beyond the current expiration at midnight on December 21, 2013.

KPS is located in the Town of Carlton, Wisconsin, in the southeast corner of Kewaunee County, Wisconsin, on the western shore of Lake Michigan. The staff issued the original construction permit for KPS on August 6, 1968, and the operating license on December 21, 1973. The plants nuclear steam supply system consists of a 2-loop pressurized water reactor with a dry, ambient containment (PWR-DRYAMB). The nuclear steam supply system was supplied by Westinghouse. The balance of the plant was originally designed and constructed by Pioneer Service and Engineer Company. KPS operates at a licensed power output of 1,772 megawatt-thermal (MWt), with a gross electrical output of approximately 590 megawatt-electric (MWe).

Unless otherwise indicated, this SER presents the status of the staffs review of information submitted through October 20, 2010, the cutoff date for consideration in the SER. The four open items previously identified by the staff for the SER with open items have been closed (see SER Section 1.5); therefore, no open items remain to be resolved before the final determination is reached by the staff on the LRA.

iii

TABLE OF CONTENTS Abstract .................................................................................................................................. iii Table of Contents .................................................................................................................... v List of Tables ......................................................................................................................... xii Abbreviations and Acronyms ................................................................................................ xiii 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 ................................................................................................. 1-7 1.5 Summary of Open Items .............................................................................................. 1-8 1.6 Summary of Confirmatory Items ................................................................................ 1-10 1.7 Summary of Proposed License Conditions ................................................................ 1-11 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 .............................................................. 2-2 2.1.3.1 Implementing Procedures and Documentation Sources Used for Scoping and Screening .............................................................................................................. 2-3 2.1.3.2 Quality Controls Applied to LRA Development ................................................ 2-6 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 10 CFR 54.4(a)(1) ................................. 2-8 2.1.4.2 Application of the Scoping Criteria in 10 CFR 54.4(a)(2) ............................... 2-10 2.1.4.3 Application of the Scoping Criteria in 10 CFR 54.4(a)(3) ............................... 2-16 2.1.4.4 Plant-Level Scoping of Systems and Structures ........................................... 2-21 2.1.4.5 Mechanical Component Scoping .................................................................. 2-23 2.1.4.6 Structural Scoping ........................................................................................ 2-25 2.1.4.7 Electrical Component Scoping ...................................................................... 2-26 2.1.4.8 Scoping Methodology Conclusion ................................................................. 2-27 2.1.5 Screening Methodology ...................................................................................... 2-28 2.1.5.1 General Screening Methodology .................................................................. 2-28 2.1.5.2 Mechanical Component Screening ............................................................... 2-29 2.1.5.3 Structural Component Screening .................................................................. 2-31 2.1.5.4 Electrical Component Screening ................................................................... 2-32 2.1.5.5 Screening Methodology Conclusion .............................................................. 2-33 2.1.6 Summary of Evaluation Findings ......................................................................... 2-33 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 v

Table of Contents 2.2.3 Staff Evaluation ................................................................................................... 2-34 2.2.4 Conclusion .......................................................................................................... 2-35 2.3 Scoping and Screening Results: Mechanical Systems ............................................... 2-35 2.3.1 Reactor Vessel, Internals, and Reactor Coolant System ..................................... 2-36 2.3.1.1 Reactor Vessel.............................................................................................. 2-37 2.3.1.2 Reactor Vessel Internals ............................................................................... 2-38 2.3.1.3 Reactor Coolant System ............................................................................... 2-39 2.3.1.4 Steam Generators ......................................................................................... 2-41 2.3.2 Engineered Safety Features ................................................................................ 2-42 2.3.2.1 Containment Vessel Internal Spray System .................................................. 2-42 2.3.2.2 Safety Injection System ................................................................................. 2-45 2.3.2.3 Residual Heat Removal System .................................................................... 2-46 2.3.3 Auxiliary Systems ................................................................................................ 2-47 2.3.3.1 New Fuel Storage System............................................................................. 2-48 2.3.3.2 Spent Fuel Storage System .......................................................................... 2-49 2.3.3.3 Spent Fuel Pool Cooling System ................................................................... 2-50 2.3.3.4 Fuel Handling System ................................................................................... 2-51 2.3.3.5 Cranes (Excluding Fuel Handling) System .................................................... 2-51 2.3.3.6 Service Water System ................................................................................... 2-52 2.3.3.7 Component Cooling Water System ............................................................... 2-54 2.3.3.8 Station and Instrument Air System ................................................................ 2-56 2.3.3.9 Chemical and Volume Control System .......................................................... 2-58 2.3.3.10 Control Room Air Conditioning System ....................................................... 2-59 2.3.3.11 Auxiliary Building Air Conditioning System .................................................. 2-61 2.3.3.12 Auxiliary Building Special Ventilation and Steam Exclusion System ............ 2-64 2.3.3.13 Auxiliary Building Ventilation System........................................................... 2-65 2.3.3.14 Reactor Building Ventilation System............................................................ 2-67 2.3.3.15 Turbine Building and Screenhouse Ventilation System ............................... 2-70 2.3.3.16 Shield Building Ventilation System .............................................................. 2-71 2.3.3.17 Technical Support Center Ventilation System ............................................. 2-72 2.3.3.18 Fire Protection System ................................................................................ 2-74 2.3.3.19 Diesel Generator System ............................................................................ 2-85 2.3.3.20 Circulating Water System ............................................................................ 2-86 2.3.3.21 Gaseous Waste Processing and Discharge System.................................... 2-87 2.3.3.22 Liquid Waste Processing and Discharge System ........................................ 2-88 2.3.3.23 Radiation Monitoring System....................................................................... 2-90 2.3.3.24 Makeup and Demineralizer System ............................................................. 2-90 2.3.3.25 Service Water Pretreatment System ........................................................... 2-92 2.3.3.26 Miscellaneous Drains and Sumps System................................................... 2-93 2.3.3.27 Miscellaneous Gas System ......................................................................... 2-95 2.3.3.28 Potable Water System ................................................................................ 2-96 2.3.3.29 Primary Sampling System ........................................................................... 2-96 2.3.4 Steam and Power Conversion Systems............................................................... 2-97 2.3.4.1 Turbine System ............................................................................................. 2-97 2.3.4.2 Main Steam and Steam Dump System.......................................................... 2-99 2.3.4.3 Bleed Steam System................................................................................... 2-100 2.3.4.4 Feedwater System ...................................................................................... 2-101 2.3.4.5 Condensate System .................................................................................... 2-101 2.3.4.6 Steam Generator Blowdown Treatment System .......................................... 2-103 2.3.4.7 Auxiliary Feedwater System ........................................................................ 2-104 2.3.4.8 Air Removal System.................................................................................... 2-106 vi

Table of Contents 2.3.4.9 Heater and Moisture Separator Drains System ........................................... 2-107 2.3.4.10 Heating Steam System ............................................................................. 2-108 2.3.4.11 Main Generator (Mechanical) and Auxiliaries System ............................... 2-109 2.3.4.12 Secondary Sampling System .................................................................... 2-110 2.3.4.13 Turbine Oil Purification System ................................................................. 2-110 2.3.4.14 Turbine Room Traps and Drains System .................................................. 2-111 2.4 Scoping and Screening Results: Structures ............................................................. 2-111 2.4.1 Reactor Containment Vessel............................................................................. 2-112 2.4.1.1 Summary of Technical Information in the Application .................................. 2-112 2.4.1.2 Staff Evaluation .......................................................................................... 2-113 2.4.1.3 Conclusion .................................................................................................. 2-114 2.4.2 Structures and Component Supports ................................................................ 2-114 2.4.2.1 Shield Building ............................................................................................ 2-114 2.4.2.2 Administration Building ............................................................................... 2-116 2.4.2.3 Auxiliary Building ........................................................................................ 2-117 2.4.2.4 Screenhouse Access Tunnel ...................................................................... 2-118 2.4.2.5 Technical Support Center ........................................................................... 2-119 2.4.2.6 Turbine Building .......................................................................................... 2-120 2.4.2.7 Yard Structures........................................................................................... 2-121 2.4.2.8 Discharge Structure .................................................................................... 2-123 2.4.2.9 Discharge Tunnel and Pipe......................................................................... 2-125 2.4.2.10 Intake Structure ........................................................................................ 2-125 2.4.2.11 Screenhouse ............................................................................................ 2-127 2.4.3 Component Supports ........................................................................................ 2-128 2.4.3.1 Summary of Technical Information in the Application .................................. 2-128 2.4.3.2 Conclusion .................................................................................................. 2-128 2.4.4 Miscellaneous Structural Commodities.............................................................. 2-128 2.4.4.1 Summary of Technical Information in the Application .................................. 2-128 2.4.4.2 Staff Evaluation .......................................................................................... 2-129 2.4.4.3 Conclusion .................................................................................................. 2-130 2.4.5 Nuclear Steam Supply System Structural Supports........................................... 2-130 2.4.5.1 Summary of Technical Information in the Application .................................. 2-130 2.4.5.2 Conclusion .................................................................................................. 2-131 2.5 Scoping and Screening Results: Electrical Systems/Commodity Groups ................. 2-131 2.5.1 Electrical and Instrumentation and Controls Systems........................................ 2-132 2.5.1.1 Summary of Technical Information in the Application .................................. 2-132 2.5.1.2 Staff Evaluation .......................................................................................... 2-133 2.5.1.3 Conclusion .................................................................................................. 2-133 2.6 Conclusion for Scoping and Screening .................................................................... 2-134 Section 3 Aging Management Review Results .................................................................. 3-1 3.0 Applicants 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 Staffs Review Process ......................................................................................... 3-4 3.0.2.1 Review of AMPs ............................................................................................. 3-4 3.0.2.2 Review of AMR Results .................................................................................. 3-6 3.0.2.3 USAR Supplement.......................................................................................... 3-6 3.0.2.4 Documentation and Documents Reviewed ..................................................... 3-6 3.0.3 Aging Management Programs ............................................................................... 3-6 vii

Table of Contents 3.0.3.1 AMPs That Are Consistent with the GALL Report ......................................... 3-10 3.0.3.2 AMPS That Are Consistent with the GALL Report with Exceptions or Enhancements........................................................................................................ 3-46 3.0.3.3 AMPs That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................. 3-172 3.0.4 Quality Assurance Program Attributes Integral to Aging Management Programs.......................................................................................................... 3-180 3.0.4.1 Summary of Technical Information in the Application .................................. 3-180 3.0.4.2 Staff Evaluation ........................................................................................... 3-180 3.0.4.3 Conclusion .................................................................................................. 3-181 3.1 Aging Management of Reactor Coolant System ....................................................... 3-182 3.1.1 Summary of Technical Information in the Application ........................................ 3-182 3.1.2 Staff Evaluation ................................................................................................. 3-182 3.1.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-200 3.1.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation is Recommended ........................................................................ 3-214 3.1.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................. 3-233 3.1.3 Conclusion ........................................................................................................ 3-239 3.2 Aging Management of Engineered Safety Features ................................................. 3-240 3.2.1 Summary of Technical Information in the Application ........................................ 3-240 3.2.2 Staff Evaluation ................................................................................................. 3-240 3.2.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-250 3.2.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation Is Recommended ........................................................................ 3-260 3.2.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................. 3-269 3.2.3 Conclusion ........................................................................................................ 3-274 3.3 Aging Management of Auxiliary Systems ................................................................. 3-274 3.3.1 Summary of Technical Information in the Application ........................................ 3-275 3.3.2 Staff Evaluation ................................................................................................. 3-275 3.3.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-292 3.3.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation is Recommended ........................................................................ 3-303 3.3.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................. 3-331 3.3.3 Conclusion ........................................................................................................ 3-370 3.4 Aging Management of Steam and Power Conversion Systems ................................ 3-371 3.4.1 Summary of Technical Information in the Application ........................................ 3-371 3.4.2 Staff Evaluation ................................................................................................. 3-371 3.4.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-378 3.4.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation is Recommended ........................................................................ 3-385 3.4.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................. 3-397 3.4.3 Conclusion ........................................................................................................ 3-408 3.5 Aging Management of Containments, Structures, and Component Supports ........... 3-409 3.5.1 Summary of Technical Information in the Application ........................................ 3-409 3.5.2 Staff Evaluation ................................................................................................. 3-410 3.5.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-423 viii

Table of Contents 3.5.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation is Recommended ........................................................................ 3-429 3.5.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................ 3-447 3.5.3 Conclusion ........................................................................................................ 3-457 3.6 Aging Management of Electrical Commodity Group ................................................. 3-458 3.6.1 Summary of Technical Information in the Application ........................................ 3-458 3.6.2 Staff Evaluation ................................................................................................. 3-458 3.6.2.1 AMR Results That Are Consistent with the GALL Report ............................ 3-461 3.6.2.2 AMR Results That Are Consistent with the GALL Report, for Which Further Evaluation is Recommended ........................................................................ 3-462 3.6.2.3 AMR Results That Are Not Consistent with or Not Addressed in the GALL Report ............................................................................................................ 3-467 3.6.3 Conclusion ........................................................................................................ 3-469 3.7 Conclusion for Aging Management Review Results ................................................. 3-470 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 ..................................................................................................... 4-2 4.1.3 Conclusion ............................................................................................................ 4-4 4.2 Reactor Vessel Neutron Embrittlement ........................................................................ 4-4 4.2.1 Neutron Fluence ................................................................................................... 4-4 4.2.1.1 Summary of Technical Information in the Application ...................................... 4-4 4.2.1.2 Staff Evaluation .............................................................................................. 4-5 4.2.1.3 USAR Supplement.......................................................................................... 4-6 4.2.1.4 Conclusion ...................................................................................................... 4-7 4.2.2 Upper-Shelf Energy Evaluation ............................................................................. 4-7 4.2.2.1 Summary of Technical Information in the Application ...................................... 4-7 4.2.2.2 Staff Evaluation .............................................................................................. 4-8 4.2.2.3 USAR Supplement.......................................................................................... 4-9 4.2.2.4 Conclusion ...................................................................................................... 4-9 4.2.3 Pressurized Thermal Shock Limits for Reactor Vessel Materials Due to Neutron Embrittlement ......................................................................................... 4-9 4.2.3.1 Summary of Technical Information in the Application ...................................... 4-9 4.2.3.2 Staff Evaluation ............................................................................................ 4-10 4.2.3.3 USAR Supplement........................................................................................ 4-11 4.2.3.4 Conclusion .................................................................................................... 4-11 4.2.4 Pressure-Temperature Limits .............................................................................. 4-12 4.2.4.1 Summary of Technical Information in the Application .................................... 4-12 4.2.4.2 Staff Evaluation ............................................................................................ 4-12 4.2.4.3 USAR Supplement........................................................................................ 4-13 4.2.4.4 Conclusion .................................................................................................... 4-13 4.3 Metal Fatigue ............................................................................................................. 4-13 4.3.1 Fatigue of ASME Class 1 Components ............................................................... 4-14 4.3.1.1 Component Design Transient Cycles ............................................................ 4-15 4.3.1.2 ASME Class 1 Vessels and Surge Line Piping ............................................. 4-17 4.3.1.3 Reactor Coolant Loop Piping ........................................................................ 4-19 4.3.1.4 Pressurizer Lower Head and Surge Line ...................................................... 4-20 4.3.1.5 Effects of Reactor Coolant Environment on Fatigue Life of ASME Code Class 1 Piping and Components................................................................................. 4-22 ix

Table of Contents 4.3.2 Fatigue of Non-ASME Code Class 1 Components............................................... 4-29 4.3.2.1 Non-Class 1 Piping ....................................................................................... 4-29 4.3.2.2 Auxiliary Heat Exchangers ............................................................................ 4-31 4.4 Environmental Qualification of Electrical Equipment................................................... 4-32 4.4.1 Summary of Technical Information in the Application .......................................... 4-33 4.4.2 Staff Evaluation ................................................................................................... 4-33 4.4.3 USAR Supplement .............................................................................................. 4-33 4.4.4 Conclusion .......................................................................................................... 4-34 4.5 Concrete Containment Tendon Prestress .................................................................. 4-34 4.5.1 Summary of Technical Information in the Application .......................................... 4-34 4.5.2 Staff Evaluation ................................................................................................... 4-34 4.5.3 USAR Supplement .............................................................................................. 4-34 4.5.4 Conclusion .......................................................................................................... 4-34 4.6 Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis .... 4-35 4.6.1 Reactor Containment Vessel Fatigue .................................................................. 4-35 4.6.1.1 Summary of Technical Information in the Application .................................... 4-35 4.6.1.2 Staff Evaluation ............................................................................................. 4-35 4.6.1.3 USAR Supplement ........................................................................................ 4-35 4.6.1.4 Conclusion .................................................................................................... 4-36 4.6.2 Containment Penetration Fatigue ........................................................................ 4-36 4.6.2.1 Summary of Technical Information in the Application .................................... 4-36 4.6.2.2 Staff Evaluation ............................................................................................. 4-36 4.6.2.3 USAR Supplement ........................................................................................ 4-36 4.6.2.4 Conclusion .................................................................................................... 4-36 4.7 Other Plant-Specific Time-Limited Aging Analyses..................................................... 4-37 4.7.1 Crane Load Cycle Limit ....................................................................................... 4-37 4.7.1.1 Summary of Technical Information in the Application .................................... 4-37 4.7.1.2 Staff Evaluation ............................................................................................. 4-37 4.7.1.3 USAR Supplement ........................................................................................ 4-38 4.7.1.4 Conclusion .................................................................................................... 4-38 4.7.2 Reactor Coolant Pump Flywheel ......................................................................... 4-38 4.7.2.1 Summary of Technical Information in the Application .................................... 4-38 4.7.2.2 Staff Evaluation ............................................................................................. 4-38 4.7.2.3 USAR Supplement ........................................................................................ 4-39 4.7.2.4 Conclusion .................................................................................................... 4-39 4.7.3 Leak-Before-Break .............................................................................................. 4-39 4.7.3.1 Summary of Technical Information in the Application .................................... 4-39 4.7.3.2 Staff Evaluation ............................................................................................. 4-40 4.7.3.3 USAR Supplement ........................................................................................ 4-47 4.7.3.4 Conclusion .................................................................................................... 4-47 4.7.4 Reactor Vessel Underclad Cracking .................................................................... 4-47 4.7.4.1 Summary of Technical Information in the Application .................................... 4-47 4.7.4.2 Staff Evaluation ............................................................................................. 4-47 4.7.4.3 USAR Supplement ........................................................................................ 4-48 4.7.4.4 Conclusion .................................................................................................... 4-49 4.7.5 Reactor Coolant Loop Piping Flaw Tolerance Evaluation .................................... 4-49 4.7.5.1 Summary of Technical Information in the Application .................................... 4-49 4.7.5.2 Staff Evaluation ............................................................................................. 4-49 4.7.5.3 USAR Supplement ........................................................................................ 4-52 4.7.5.4 Conclusion .................................................................................................... 4-53 x

Table of Contents 4.8 Conclusion for Time-Limited Aging Analyses ............................................................. 4-53 Section 5 Review by the Advisory Committee on Reactor Safeguards .............................. 5-1 Section 6 Conclusion ......................................................................................................... 6-1 Appendix A Commitments for License Renewal of Kewaunee Power Station .................... A-1 Appendix B Chronology ..................................................................................................... B-1 Appendix C Principal Contributors ..................................................................................... C-1 Appendix D References ..................................................................................................... D-1 xi

Aging Management Review Results 3.1.2.2.15 Changes in Dimensions Due to Void Swelling The staff reviewed LRA Section 3.1.2.2.15 and Table 3.1.1, item 3.1.1-33 against the criteria in SRP-LR Section 3.1.2.2.15. LRA Section 3.1.2.2.15 addresses changes in dimensions due to void swelling that could occur in stainless steel and Ni-alloy PWR RVI components exposed to reactor coolant as an aging effect that the applicant will manage, consistent with the SRP-LR, by the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program. This AMP is enhanced with Commitment No. 1, which is also identified in the USAR supplement description of the program.

SRP-LR Section 3.1.2.2.15 states that:

[c]hanges in dimensions due to void swelling could occur in stainless steel and nickel alloy PWR reactor internal components exposed to reactor coolant. The GALL Report recommends no further [AMR] if the applicant provides a commitment in the FSAR Supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

As described in LRA Sections 3.1.2.2.15, A.2.1.2, and B.2.1.2, the applicant made Commitment No. 1 to enhance its ASME Section XI ISI, Subsections IWB, IWC, and IWD Program to incorporate all three GALL Report requirements stated above regarding managing aging effects on RVIs. Therefore, the staff concludes that the applicants program meets the SRP-LR Section 3.1.2.2.15 criteria because using the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program with Commitment No. 1 to manage the aging effects due to SCC and IASCC is consistent with the SRP-LR guidance. The staff also confirmed that LRA Table 3.1.2-2 identified all GALL AMR Table IV.B2 items under this aging mechanism (IV.B2-1, IV.B2-4, IV.B2-7, IV.B2-11, IV.B2-15, IV.B2-19, IV.B2-23, IV.B2-27, IV.B2-29, IV.B2-35, IV.B2-39, and IV.B2-41).

The staff concludes that the LRA is consistent with the GALL Report and that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the CLB during the period of extended operation, as required by 10 CFR 54.21(a)(3).

3.1.2.2.16 Cracking Due to Stress-Corrosion Cracking and Primary Water Stress-Corrosion Cracking The staff reviewed LRA Section 3.1.2.2.16 against the criteria in SRP-LR Section 3.1.2.2.16.

Item 1. The staff reviewed LRA Section 3.1.2.2.16.1 against the criteria in SRP-LR Section 3.1.2.2.16. LRA Table 3.1.1, item 3.1.1-34 describes the cracking due to SCC and PWSCC of austenitic stainless steel reactor vessel components that were exposed to reactor coolant. The AMR items corresponding to item 3.1.1-34 include the CRDM pressure housing and the stainless steel portion of the closure head instrument tubes and spare CRDM penetrations, bottom head instrument tube penetrations, and closure head CRDM penetrations (Table 3.1.2-1). The applicant stated that cracking due to SCC of these components is managed by the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program and Primary Water Chemistry Program. The applicant further stated that the programs are consistent with the GALL Report.

3-229

Aging Management Review Results The staff reviewed LRA item 3.1.1-34 in comparison with the GALL Report, Volume 1, Table 1, ID 34. In its review, the staff noted that for these components or portion of the components constructed of austenitic stainless steel, the GALL Report recommends a combination of ASME Section XI ISI and control of primary water chemistry to manage the effect of cracking due to SCC. The staffs reviews of the applicants ASME Section XI ISI, Subsections IWB, IWC, and IWD Program and the Primary Water Chemistry Program are discussed in SER Sections 3.0.3.2.1 and 3.0.3.1.9, respectively. In its review, the staff found that the applicants programs are consistent with the GALL Report and are, therefore, acceptable.

On the basis of its review, the staff determines that the applicants proposed program is acceptable for managing the cracking due to SCC in austenitic stainless steel reactor vessel components corresponding to item 3.1.1-34. The staff concludes that the applicant has demonstrated that the effects of aging for these components will be adequately managed so that the intended functions will be maintained consistent with the CLB for the period of extended operation, as required by 10 CFR 54.21(a)(3).

Steam generator components associated with LRA Section 3.1.2.2.16.1. The components covered in GALL Report Table 3.1.1, item 3.1.1-35 are applicable to B&W model OTSGs. KPS has Westinghouse recirculating steam generators, so this item is not applicable to KPS, except for the case discussed in the following paragraphs.

SRP-LR Section 3.1.2.2.16.1 identifies that cracking due to PWSCC could occur on the primary coolant side of PWR steel steam generator tube-to-tubesheet welds made or clad with Ni-alloy.

The GALL Report recommends controls of the ASME Code Section XI ISI, Subsections IWB, IWC, and IWD and Water Chemistry programs to manage this aging, and recommends no further AMR for PWSCC of Ni-alloy if the applicant complies with applicable NRC orders and provides a commitment in its USAR supplement to implement applicable NRC bulletins, GLs, and staff-accepted industry guidelines. GALL Report Revision 1, Volume 2 addresses this aging in item IV.D2-4, stating the item is applicable to OTSGs, but not to recirculating steam generators.

USAR Section 4.2.2.6 states that the applicants steam generator tubes are fabricated from Alloy 690TT (Thermally Treated), that the side of the tubesheet in contact with the reactor coolant is clad with Inconel (Alloy 600 in USAR Table 4.2-1), and that the tube-to-tubesheet joints are welded.

The staff noted that the ASME Code Section XI does not require inspection of the tube-to-tubesheet welds. In addition, no specific NRC orders or bulletins address inspection requirements for these welds. The staffs concern is that, if the tubesheet cladding is Alloy 600, autogenous tube-to-tubesheet welds may not have sufficient chromium content to prevent initiation of PWSCC, even when the steam generator tubes are made from Alloy 690TT, which is the configuration of the applicants steam generator tubes. Consequently, such a PWSCC crack initiated in this region, close to a tube, could propagate into or through the weld, causing a failure of the weld and of the RCPB, even for recirculating steam generators such as those of the applicant. Therefore, because the NRC has not approved a redefinition of the pressure boundary for these steam generators in which the autogenous tube-to-tubesheet weld is no longer included, the staff considers that the effectiveness of the primary water chemistry program should be verified to ensure PWSCC cracking is not occurring.

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Aging Management Review Results In a conference call on October 13, 2010, between the staff and the applicant, the staff questioned how cracking in the applicants steam generator tube-to-tubesheet welds will be managed if that material is susceptible to PWSCC. The applicant agreed to provide information on its management of this issue.

In its response dated October 20, 2010, the applicant stated that it will commit to developing a plan to address potential failure of the steam generator primary-to-secondary pressure boundary due to PWSCC cracking of tube-to-tubesheet welds. The applicant further stated that the plan will consist of two resolution options:

In the first option, the applicant stated that it would perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to establish a technical basis for concluding that the structural integrity of the steam generator tube-to-tubesheet interface is adequately maintained even with the presence of tube-to-tubesheet weld cracking, and that the steam generator tube-to-tubesheet weld is not required for the RCPB.

In the second option, the applicant stated that it would perform a one-time inspection of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. The applicant also stated that if weld cracking is identified, the condition will be resolved through repair or engineering evaluation for continued service, as appropriate, and that an ongoing monitoring program will be established to perform routine inspections of tube-to-tubesheet welds for the remaining life of the steam generators.

Moreover, the applicant stated that it will develop its plan prior to the period of extended operation. As described in its response to RAI 3.1.2.2.13-1a dated September 23, 2010, the applicant explained that the lower portions of its steam generators, including the tubes and tubesheets, have accumulated less than 10 years of service time since having been replaced in 2001. Considering this limited service time of the replaced portions of the steam generators, the applicant further stated that the implementation of its plan, including weld inspections for the presence of PWSCC cracking if necessary, will be completed prior to 50 years of plant operation (i.e., prior to 2023). Finally, the applicant stated that Commitment No. 53, covering the above plan to manage the aging effect due to PWSCC of steam generator tube-to-tubesheet welds, will be added to LRA Appendix A, USAR Table A6.0-1.

Based on its review, the staff finds the applicants plan and associated Commitment No. 53 acceptable because the applicant stated that it will manage the aging effect of cracking due to PWSCC in the steam generator tube-to-tubesheet welds either by demonstrating that those welds do not have a structural integrity or pressure boundary function, or by implementing a one-time inspection capable of detecting PWSCC cracking on a representative number of tube-to-tubesheet welds of each steam generator, in a time period consistent with the detection of potential PWSCC cracks and the period of extended operation. The staff finds that the timing of this inspection prior to 50 years of plant operation is acceptable because at that time, the replaced lower portion of the steam generator will have been in operation for less than 22 years, and it is unlikely that significant PWSCC cracking will have initiated. The staff also notes that, in case the aging effect is revealed, this one-time inspection program is accompanied by an appropriate corrective action process, including an evaluation of the degradation and the implementation of routine inspections for the remaining life of the steam generators. The staff concludes that the applicant has demonstrated that the effects of aging for these components 3-231

Aging Management Review Results will be adequately managed so that their intended functions will be maintained consistent with the CLB during the period of extended operation, as required by 10 CFR 54.21(a)(3).

Item 2. The components covered by Table 3.1.1, item 3.1.1-36 are not applicable to KPS. See SER Section 3.1.2.1.1.

3.1.2.2.17 Cracking Due to Stress-Corrosion Cracking, Primary Water Stress-Corrosion Cracking, and Irradiated-Assisted Stress-Corrosion Cracking The staff reviewed LRA Section 3.1.2.2.17 and Table 3.1.1, item 3.1.1-37 against the criteria in SRP-LR Section 3.1.2.2.17. LRA Section 3.1.2.2.17 addresses cracking due to SCC, PWSCC, and IASCC that could occur in stainless steel and Ni-alloy PWR reactor internal components exposed to reactor coolant as an aging effect that the applicant will manage, consistent with the SRP-LR, with the Primary Water Chemistry Program and the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program. The ASME Section XI ISI, Subsections IWB, IWC, and IWD Program is enhanced with Commitment No. 1, which is also identified in the USAR supplement description of the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program.

SRP-LR Section 3.1.2.2.17 states that:

[c]racking due to [SCC, PWSCC, and IASCC] could occur in PWR stainless steel and nickel alloy reactor vessel internals components. The existing program relies on control of water chemistry to mitigate these effects. However, the existing program should be augmented to manage these aging effects for reactor vessel internals components. The GALL Report recommends no further AMR if the applicant provides a commitment in the USAR Supplement to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

As indicated in SER Section 3.0.3.1.9, the staff accepts the Primary Water Chemistry Control Program for mitigating the aging effects due to SCC, PWSCC, and IASCC, meeting one of the requirements mentioned in SRP-LR Section 3.1.2.2.17. Furthermore, the applicant made Commitment No. 1 in LRA Sections 3.1.2.2.17, A.2.1.2, and B.2.1.2 to enhance its ASME Section XI ISI, Subsections IWB, IWC, and IWD Program to incorporate all three GALL Report requirements stated above regarding managing aging effects on reactor internals. Therefore, the staff concludes that the applicants program meets the SRP-LR Section 3.1.2.2.17 criteria because, in addition to the required Water Chemistry Control Program, using the ASME Section XI ISI, Subsections IWB, IWC, and IWD Program with Commitment No. 1 to manage the aging effects due to SCC, PWSCC, and IASCC is consistent with the SRP-LR guidance.

The staff also confirmed that LRA Table 3.1.2-2 identified all GALL AMR Table IV.B2 items under this aging mechanism (IV.B2-16, IV.B2-20, IV.B2-28, and IV.B2-40). The staff concludes that the LRA is consistent with the GALL Report and that the applicant has demonstrated that the effects of aging will be adequately managed so that the intended functions will be maintained consistent with the CLB during the period of extended operation, as required by 10 CFR 54.21(a)(3).

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APPENDIX A COMMITMENTS FOR LICENSE RENEWAL OF KEWAUNEE POWER STATION During the review of the Kewaunee Power Station (KPS) license renewal application (LRA) by the staff of the U.S. Nuclear Regulatory Commission (the staff), Dominion Energy Kewaunee, Inc. (Dominion, DEK, or the applicant), made commitments related to aging management programs (AMPs) to manage aging effects of structures and components (SCs) prior to the period of extended operation. The following table lists these commitments, along with the implementation schedules and the sources of the commitment.

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule 1 The ASME Code Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program At least 2 years prior to ASME Code Section XI will be enhanced to: (1) participate in the industry programs for investigating and managing entering the period of Inservice Inspection, aging effects on reactor internals; (2) evaluate and implement the results of the industry extended operation. Subsections IWB, IWC, programs as applicable to the reactor internals; and (3) upon completion of these programs, and IWD but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the staff for review and approval to augment the current inspections.

2 The ASME Code Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program At least 2 years prior to ASME Code Section XI will be enhanced to include identification of the limiting susceptible cast austenitic stainless entering the period of Inservice Inspection, (CASS) steel reactor vessel internal components from the standpoint of thermal aging extended operation. Subsections IWB, IWC, susceptibility, neutron fluence, and cracking. For each identified component, a plan will be and IWD developed that accomplishes aging management through either a supplemental examination or a component-specific evaluation. The plan will be submitted for staff review and approval, not less than 24 months before entering the period of extended operation.

3 The Bolting Integrity Program will be enhanced to further incorporate applicable Electric Power Prior to the Period of Bolting Integrity Research Institute (EPRI) and industry bolting guidance. Topic enhancements will include Extended Operation proper joint assembly, torque values, gasket types, use of lubricants, and other bolting fundamentals.

4 The Buried Piping and Tanks Inspection program will be enhanced to perform visual Prior to the Period of Letter 10-548, Response inspections of a representative sample of material/protective measure combinations for Extended Operation to RAI B2.1.7-3a.

in-scope buried piping and tanks.

The following materials are utilized in buried applications with the associated protective And measures:

  • Steel (including cast iron)/coated, During the first 10 years
  • Steel/coated and wrapped, of the period of extended
  • Steel/uncoated, and operation
  • Stainless steel/coated and wrapped And Visual inspections of the external surfaces of the components will be performed to identify damaged wrapping (if present), degraded or damaged coating (if present), and evidence of During the second loss of material. Each piping inspection will include a minimum of 10 linear feet of piping. 10 years of the period of extended operation The following inspections will be performed:

The circulating water system 30 inch diameter recirculation line, which is coated and wrapped carbon steel, will receive one inspection prior to the period of extended operation, and additional inspections within the first 10 years and second 10 years of the period of extended operation. (Continued next page)

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule The circulating water system recirculation line vent piping, which is coated and wrapped stainless steel, will receive one inspection prior to the period of extended operation and additional inspections within the first 10 years and second 10 years of the period of extended operation.

The diesel generator system fuel oil piping, which includes coated and wrapped carbon steel fuel oil supply and return piping, storage tank vent piping, and day tank vent piping, will receive one inspection prior to the period of extended operation and additional inspections within the first 10 years and second 10 years of the period of extended operation. The inspections will be performed in the non-cathodically protected portion of the piping.

The diesel generator system fuel oil storage tanks, which are coated carbon steel, will receive one inspection of one tank prior to the period of extended operation. An additional tank inspection will be performed within each of the first and second 10 years of the period of extended operation.

The diesel generator system fuel oil storage tanks hold down straps, which are uncoated carbon steel, will be inspected in conjunction with the associated fuel oil storage tank inspection. One set will be inspected prior to the period of extended operation, and one set will be inspected within each of the first and second 10 years of the period of extended operation.

The fire protection system piping, which is coated ductile iron, will receive three inspections prior to the period of extended operation, and three additional inspections within each of the first and second 10 years of the period of extended operation.

5 The Compressed Air Monitoring Program will be enhanced to incorporate the compressed air Prior to the Period of Compressed Air system testing and maintenance recommendations from the ASME OM-S/G-1998, Part 17 and Extended Operation Monitoring the EPRI TR-108147 and to identify these documents as part of the program basis.

6 The External Surfaces Monitoring Program will be enhanced to inspect the accessible external Prior to the Period of External Surfaces surfaces of in-scope components, piping, supports, structural members, and structural Extended Operation Monitoring commodities, in the infrequently accessed areas, consistent with the criteria used in other plant areas.

7 The External Surfaces Monitoring Program will be enhanced to provide training for operations, Prior to the Period of External Surfaces engineering, and health physics personnel performing the program inspections and walkdowns. Extended Operation Monitoring The training will address: (1) the requirements of the External Surfaces Monitoring Program for license renewal, (2) the need to document the identified conditions with sufficient detail to support monitoring and trending the aging effects, and (3) the aging effects monitored by the program and how to identify them.

8 The Fire Protection Program will be enhanced to test a representative sample of sprinkler Prior to the sprinkler Fire Protection heads or to replace all affected sprinkler heads in accordance with the requirements of National heads achieving Fire Protection Association (NFPA) 25. 50 years of service life.

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule 9 The Fire Protection Program fire barrier penetration seal inspections will be revised to include Prior to the Period of Fire Protection the elastomer shield building fire boots. Extended Operation

10. The Fire Protection Program inspections of the reactor coolant pump oil collection system will Prior to the Period of Fire Protection be revised to include additional inspection criteria for the visual inspection of the system and to Extended Operation perform a one-time inspection of the internal surfaces of the reactor coolant pump oil collection tank.
11. The Fuel Oil Tank Inspections Program will be enhanced to provide guidance for the periodic Prior to the Period of Fuel Oil Tanks Inspection draining, cleaning, and inspection activities. Extended Operation
12. The Inspection of Overhead Heavy Load and Refueling Handling Systems Program will be Prior to the Period of Inspection of Overhead enhanced to clarify the requirements of visual inspection of structural members, including Extended Operation Heavy Load and structural bolting, of the in-scope heavy load and refueling handling cranes and associated Refueling Handling equipment. Systems
13. The Metal-Enclosed Bus (MEB) Program will be enhanced to include augmented periodical Prior to the Period of Letter 09-469, Response visual inspections of the MEB internal surfaces, bus supports, bus insulation, taped joints, and Extended Operation. to RAI B2.1.18-1 boots for signs of degradation or aging. Thereafter, the inspection of all MEB will not exceed a 10-year interval and the inspection of the sample of bolted connections will not exceed a 5-year interval.
14. The Non-EQ Electrical Cables and Connections Program will be established. The program will Prior to the Period of Non-EQ Electrical Cables periodically visually inspect for accessible electrical cables and connections installed in an Extended Operation. and Connections adverse localized equipment environment. Should an adverse localized environment be Thereafter, the observed, a representative sample of electrical cables and connections installed within that inspections will not environment will be visually inspected for jacket surface anomalies. exceed a 10-year interval.
15. The Non-EQ Electrical Cable Connections Program will be established. The program will Prior to the Period of Non-EQ Electrical Cables perform a one-time inspection, on a sampling basis, to confirm the absence of loosening of Extended Operation and Connections bolted connections.

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule

16. The Non-EQ Inaccessible Medium-Voltage Cables Program will be established. The program Prior to the Period of Non-EQ Inaccessible will periodically inspect the in-scope manholes and pulling pit for water collection and will Extended Operation. Medium-Voltage Cables, remove water, if required. The program will periodically perform a test on the in-scope cables to Thereafter, the manholes and Letter 10-548, RAI provide an indication of the condition of the conductor insulation. and pulling pit response to inspections will be RAI B2.1.21-1a.

performed at least every 2 years.

And Thereafter, the cable testing will be performed at least every 10 years.

17. The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals Prior to the Period of Non-EQ Instrumentation Program will be established. The program will periodically perform a proven cable system test Extended Operation. Circuits Subject to for detecting deterioration of the insulation system for those electrical cables and connections Thereafter, the cable Sensitive, High-Voltage, disconnected during calibration, or will periodically review the results and findings of testing and calibration Low-Level Signals calibrations for those electrical cables that remain connected during the calibration process. reviews will not exceed a 10-year interval.
18. The Open-Cycle Cooling Water System Program will be enhanced to add the applicable aging Prior to the Period of Open-Cycle Cooling effects as inspection criteria for the circulating water system underwater visual inspections. Extended Operation Water System
19. The Reactor Vessel Surveillance Program will be enhanced to include the applicable limitations Prior to the Period of Reactor Vessel on operating conditions to which the surveillance capsules were exposed (e.g., neutron flux, Extended Operation Surveillance spectrum, irradiation temperature, etc.).
20. The Reactor Vessel Surveillance Program will be enhanced to include requirements for storing, Prior to the Period of Reactor Vessel and possible recovery, of tested and untested capsules (removed from the reactor vessel after Extended Operation Surveillance August 31, 2000).
21. The Selective Leaching of Materials Program will be established. The program will perform a Prior to the Period of Selective Leaching of one-time visual inspection and hardness measurement or qualitative examination of selected Extended Operation Materials components, within the scope of license renewal for selective leaching.
22. The Structures Monitoring Program will be enhanced to clearly define structures, structural Prior to the Period of Letter 09-469, Response elements, and miscellaneous structural commodities that are in-scope. Defined scope to Extended Operation to RAI B2.1.18-2 include the MEB enclosure assemblies, structural supports, and enclosure seals.
23. The Structures Monitoring Program will be enhanced to monitor groundwater quality and verify Prior to the Period of Structures Monitoring that it remains non-aggressive to below-grade concrete. Extended Operation Program
24. The Structures Monitoring Program will be enhanced to improve criteria for the detection of Prior to the Period of Structures Monitoring aging effects for the underwater visual inspections of the in-scope structures. Extended Operation Program A-5

Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule

25. The Work Control Process Program will be established. The program will perform one-time Prior to the Period of Letter 09-597, Changes to inspections as a verification of the effectiveness of chemistry control programs. The program Extended Operation the WCP Program will also perform visual inspections of component internal surfaces and external surfaces of selected components to manage the effects of aging when the surfaces are made available for examination through surveillance and maintenance activities.
26. Deleted N/A Letter 09-597
27. Deleted N/A Letter 09-597
28. The Metal Fatigue of Reactor Coolant Pressure Boundary Program will be enhanced to include Prior to the Period of Metal Fatigue of Reactor a routine assessment of the transient cycle count totals and fatigue usage status for monitored Extended Operation Coolant Pressure locations, including an action limit for the initiation of corrective action. Boundary
29. The following will be further evaluated as part of the applicants ongoing performance Prior to the Period of Environmental Report -

improvement programs: Extended Operation SAMA Analysis, Letters SAMA 160: Install Emergency Diesel Generator (EDG) exhaust duct insulation.09-028 and 09-291 Concurrent implementation of SAMAs 81,160,166, and 167.

Implementation of temporary screenhouse ventilation.

30. Quarterly laboratory testing of fuel oil samples for water, sediment, and particulates will be Prior to the Period of Letter 09-680, performed on the EDG day tanks and on the technical support center diesel generator (TSC Extended Operation RAI response to DG) day tank. The testing acceptance criteria will be consistent with the requirements specified B2.1.14-3 in American Society for Testing and Materials (ASTM) D975-06b for water and sediment and ASTM D6217 for particulates.
31. The Work Control Process Program will be enhanced to provide for a one-time-inspection of Prior to the Period of Letter 09-469, Response the EDG day tanks and the TSC DG day tank. An exterior surfaces ultrasonic test (UT) Extended Operation to RAI B2.1.15-1 inspection will be performed to verify wall thickness of the bottom of each day tank. Based upon the UT inspections, the most limiting EDG day tank will also be drained, cleaned, and visually inspected as a leading indicator for the remaining tanks.
32. The 14 potentially cost beneficial SAMAs identified in LRA Appendix E, Attachment F, will be Prior to the Period of Environmental Report -

further evaluated as part of the applicants ongoing performance improvement programs. Extended Operation SAMA Analysis

33. Develop a plan for identification and remediation of reactor refueling cavity liner leakage to be Prior to the Period of Letter 09-760, Response implemented during the period of extended operation. Extended Operation to RAI B2.1.31-4a
34. At least one core bore sample will be taken from the waste drumming room reinforced concrete Prior to the end of 2011 Letter 10-093, Response ceiling below the spent fuel pool. The core sample location and depth will be sufficient to to RAI B2.1.31-5a validate the strength of the concrete and the extent of any degradation. The core sample will be tested for compressive strength and will be subject to petrographic examination. Reinforcing steel in the core sample area will be exposed and inspected for material condition.

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule

35. Develop an action plan for identification and remediation of spent fuel pool (SFP) liner leakage Prior to the Period of Letter 09-760, Response to be implemented during the period of extended operation. Extended Operation to RAI B2.1.31-5a
36. If SFP liner leakage persists during the period of extended operation, an additional concrete Prior to the end of the Letter 09-760, Response core sample will be taken from the waste drumming room reinforced concrete ceiling below the first 10 years of the to RAI B2.1.31-5a spent fuel pool. The core sample location and depth will be sufficient to validate the strength of Period of Extended the concrete and the extent of any degradation. The core sample will be tested for compressive Operation strength and will be subject to petrographic examination. Reinforcing steel in the core sample area will be exposed and inspected for material condition.
37. Perform a VT-1 visual examination of the stainless steel cladding of a safety injection pump for Prior to the Period of Letter 09-777, Response indications of cracking or corrosion due to cladding breach. Extended Operation to RAI 3.2.2.2.2
38. The Boron Carbide Surveillance Program, which includes neutron attenuation testing, will During the Period of Letter 09-777, continue to be performed during the period of extended operation every 3 years. Extended Operation Supplemental Response to RAI 3.3.2.2.6-1
39. A surveillance program will be implemented to perform verification that the Boral spent fuel Prior to 2017. Letter 09-777, storage rack neutron absorber B-10 areal density is maintained within the bounds of the spent Surveillance program will Supplemental Response fuel pool criticality analysis. Alternatively, the criticality analysis for the spent fuel pool will be be performed every 10 to RAI 3.3.2.2.6-2 revised to eliminate credit for the Boral neutron absorber material. years thereafter
40. Implement nitrate monitoring for the component cooling system on a frequency consistent with Prior to the Period of Letter 10-008, Response the existing monitoring for ammonia. Extended Operation to RAI B2.1.8-3a
41. Perform a fatigue analysis of the surge line hot leg nozzle and the charging line nozzle in Completed Letter 10-033, Final accordance with ASME Boiler and Pressure Vessel (B&PV) Code Section III, Subsection Response to RAI B3.2-2, NB-3200 guidance and determine the cumulative usage factor (CUF), considering the effects of Letter 10-324, Completion the reactor coolant environment. Confirm that CUF is less than 1.0 at the end of 60 years of of Kewaunee Power plant operation. Station License Renewal Commitment 41.
42. For Examination Category B-J, item No. B9.21, eight ASME Class 1 small-bore circumferential During each 10-year lSI Letter 10-033, welds will receive volumetric and surface examinations during each 10-year lSI inspection inspection interval during Supplemental Response interval during the period of extended operation. the period of extended to RAI B2.1.2-1 operation A-7

Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule

43. Ten volumetric examinations of ASME Class 1 small-bore socket welds will be performed using Four volumetric Letter 10-665, a demonstrated, nuclear-industry endorsed, inspection methodology that can detect cracking examinations or two Supplemental Response within the specified examination volume, if a methodology becomes available. In the event that destructive examinations to RAI B2.1.2-2 a demonstrated, nuclear-industry endorsed, inspection methodology is not available, (or an equivalent destructive examinations of socket welds will be substituted for volumetric nondestructive combination of examinations. Each destructive weld examination will be considered equivalent to performing examinations) prior to two volumetric weld examinations, such that a maximum of five destructive examinations will the period of extended be performed. operation.

Remaining examinations within three years of entering the period of extended operation.

44. Core samples will be obtained from the inside surface of a concrete wall (below the Prior to the Period of Letter 10-093, Response groundwater table elevation) or from the foundation basemat in the vicinity of the groundwater Extended Operation to RAI B2.1.31-3a wells for which average sampling results have exceeded the chloride concentration limit of 500 ppm. The concrete core samples will be tested to determine if the chloride content within the concrete could cause degradation due to corrosion of reinforcing steel.
45. In the event that the chloride content in the groundwater does not decrease to below 500 ppm Prior to the end of the Letter 10-093, Response within the first ten years of the period of extended operation, core samples will be obtained first 10 years of to RAI B2.1.31-3a from the inside surface of a concrete wall (below the groundwater table elevation) or from the extended operation.

foundation basemat in the vicinity of a groundwater well for which average sampling results have exceeded the chloride concentration limit of 500 ppm. The concrete core samples will be tested to determine if the chloride content within the concrete could cause degradation due to corrosion of reinforcing steel.

46. If the results of the core sample testing of the waste drumming room reinforced concrete ceiling Prior to the Period of Letter 10-093, Response leakage site (related to potential SFP liner leakage - Commitment 34) indicate degradation of Extended Operation to RAI B2.1.31-4a the structural integrity of the concrete, at least one core bore sample will be taken near at least one of the refueling cavity liner leakage indication sites. The core sample location and depth will be sufficient to validate the strength of the concrete and the extent of any degradation. The core sample will be tested for compressive strength and will be subject to petrographic examination. Reinforcing steel in the core sample area will be exposed and inspected for material condition.
47. Submit three examples of operating experience associated with the Work Control Process - Within 2 years following Letter 10-286; Response Internal Surfaces Monitoring Program for NRC staff review in determining the effectiveness of implementation of the to RAI B2.1.32-5 the program to detect and correct the effects of aging prior to the loss of function. WCP aging management program A-8

Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule

48. The cathodic protection system associated with the diesel generator fuel oil storage tanks and During the Period of Letter 10-548 Response protected portions of the fuel oil lines, and the circulating water system recirculation piping, will Extended Operation to RAI-B2.1.7-3a each be maintained available a minimum of 90% of the time during the period of extended operation. In addition, NACE cathodic protection system surveys will be performed at least annually during the period of extended operation.

49 Recognizing that the EPRI Steam Generator Maintenance Program (SGMP) resolution is still Prior to 2023 Letter 10-548 Response under development, Kewaunee will perform an inspection of each steam generator to assess to RAI-3.1.2.2.13-1a the condition of the divider plate assembly. The examination technique(s) will be capable of detecting PWSCC in the divider plate assembly and associated welds. The steam generator divider plate inspections will be completed prior to exceeding 10 years into the period of extended operation. In addition, Dominion Energy Kewaunee, Inc., (Dominion, DEK, or the applicant) will continue to actively participate in the EPRI SGMP studies.

50 Perform an audit of the Internal Surfaces Monitoring portion of the Work Control Process Prior to the Period of Letter 10-595 Program inspections to confirm that the components representing the leading indicators of Extended Operation and (Supplemental Response aging for each of the material/environment combinations have been inspected at least once every 10 years to RAI B2.1.32-5a) during the audit period. thereafter.

If any scheduled surveillance and maintenance activities which were intended to encompass Deliberate focused components as leading indicators of aging in each of the material/environment combinations inspections will be have not been performed, then perform deliberate focused inspections of these components. performed within 5 years of completion of the audits.

51 DEK will perform a fatigue evaluation of the pressurizer lower head and surge line that is Prior to the Period of Letter 10-595 consistent with the requirements of ASME B&PV Code,Section III, NB-3200 and will determine Extended Operation Supplemental Response the cumulative fatigue usage through the period of extended operation. to RAI B3.2-2a 52 DEK will perform a review of design basis ASME Code Class 1 component fatigue evaluations Prior to the Period of Letter 10-595 to determine whether the NUREG/CR-6260-based components that have been evaluated for Extended Operation Supplemental Response the effects of the reactor coolant environment on fatigue usage are the limiting components for to RAI B3.2-2a the Kewaunee plant configuration. If more limiting components are identified, the most limiting component will be evaluated for the effects of the reactor coolant environment on fatigue usage.

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Appendix A APPENDIX A: LONG TERM COMMITMENTS FOR LICENSE RENEWAL OF KPS No. Commitment Implementation Source Schedule 53 DEK will develop a plan to address the potential for failure of the primary-to-secondary Develop a plan prior to Letter 10-595 pressure boundary due to PWSCC cracking of tube-to-tubesheet welds. the Period of Extended The plan will consist of two resolution options: Operation

1. Perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order Implement the to: requirements of the plan a) Establish a technical basis which concludes that the structural integrity of the steam prior to 2023 generator tube-to-tubesheet interface is adequately maintained with the presence of tube-to-tubesheet weld cracking, and b) Establish a technical basis which concludes that the steam generator tube-to-tubesheet welds are not required to perform a reactor coolant pressure boundary function.

-or-

2. Perform a one-time inspection of a representative number of tube-to-tubesheet welds in each steam generator to determine if PWSCC cracking is present. If weld cracking is identified:

a) The condition will be resolved through repair or engineering evaluation to justify continued service, as appropriate, and b) An ongoing monitoring program will be established to perform routine tube-to-tubesheet inspections for the remaining life of the steam generators.

54 The Structures Monitoring Program will be revised to include the evaluation criteria of Prior to the Period of Letter 10-707, Response ACI 349.3R-96, Chapter 5, as the criteria to be used when evaluating conditions or findings Extended Operation to RAI B2.1.31-9.

identified during concrete structure inspections. This will be done prior to the performance of the next scheduled inspection, which will occur prior to the period of extended operation.

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