ML14339A651

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Updated Safety Analysis Report (Usar), Rev 25 - Chapter 15: Programs and Activities That Manage the Effects of Aging
ML14339A651
Person / Time
Site: Kewaunee  Dominion icon.png
Issue date: 11/24/2014
From:
Dominion Energy Kewaunee
To:
Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML14339A626 List:
References
14-572
Download: ML14339A651 (66)
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Revision 2511/26/14 KPS USAR 15-i

15.1 INTRODUCTION

15-1 15.1.1 Quality Assurance and Administrative Controls.........................

15-1 15.1.2 Operating Experience.............................................

15-3 15.2 AGING MANAGEMENT...........................................

15-3 15.2.1 Aging Management Programs.......................................

15-3 15.2.2 Time Limited Aging Analyses......................................

15-4 15.2.3 Time limited aging analysis Support Programs.........................

15-5 15.3 PROGRAMS THAT MANAGE THE EFFECTS OF AGING...............

15-5 15.3.1 Alloy 600 Inspections.............................................

15-5 15.3.2 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD...

15-6 15.3.3 ASME Section XI, Subsection IWE..................................

15-7 15.3.4 ASME Section XI, Subsection IWF..................................

15-8 15.3.5 Bolting Integrity.................................................

15-8 15.3.6 Boric Acid Corrosion.............................................

15-9 15.3.7 Buried Piping and Tanks Inspection..................................

15-9 15.3.8 Closed-Cycle Cooling Water System.................................

15-11 15.3.9 Compressed Air Monitoring........................................

15-12 15.3.10 External Surfaces Monitoring.......................................

15-12 15.3.11 Fire Protection...................................................

15-13 15.3.12 Flow-Accelerated Corrosion........................................

15-14 15.3.13 Flux Thimble Tube Inspection......................................

15-15 15.3.14 Fuel Oil Chemistry...............................................

15-15 15.3.15 Fuel Oil Tank Inspections..........................................

15-16 15.3.16 Inspection of Overhead Heavy Load and Refueling Handling Systems.......

15-17 15.3.17 Lubricating Oil Analysis...........................................

15-17 15.3.18 Metal Enclosed Bus...............................................

15-17 15.3.19 Non-EQ Electrical Cables and Connections............................

15-18 15.3.20 Non-EQ Electrical Cable Connections................................

15-19 15.3.21 Non-EQ Inaccessible Medium-Voltage Cables.........................

15-20 15.3.22 Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals.........................................................

15-21 15.3.23 Open-Cycle Cooling Water System..................................

15-21 15.3.24 Primary Water Chemistry..........................................

15-22 Chapter 15: Programs and Activities that Manage the Effects of Aging Table of Contents Section Title Page

Revision 2511/26/14 KPS USAR 15-ii Chapter 15: Programs and Activities that Manage the Effects of Aging Table of Contents (continued)

Section Title Page 15.3.25 Reactor Containment Leakage Testing 10 CFR 50, Appendix J............

15-22 15.3.26 Reactor Head Closure Studs........................................

15-23 15.3.27 Reactor Vessel Surveillance........................................

15-23 15.3.28 Secondary Water Chemistry........................................

15-24 15.3.29 Selective Leaching of Materials.....................................

15-24 15.3.30 Steam Generator Tube Integrity.....................................

15-25 15.3.31 Structures Monitoring Program......................................

15-25 15.3.32 Work Control Process.............................................

15-28 15.4 TIME-LIMITED AGING ANALYSES.................................

15-31 15.4.1 Reactor Vessel Neutron Embrittlement................................

15-31 15.4.2 Metal Fatigue....................................................

15-33 15.4.3 Environmental Qualification of Electric Equipment......................

15-36 15.4.4 Containment Fatigue Analysis......................................

15-37 15.4.5 Other Plant-Specific Time-Limited Aging Analyses.....................

15-37 15.5 TLAA SUPPORT PROGRAMS......................................

15-40 15.5.1 Environmental Qualification (EQ) of Electric Components................

15-40 15.5.2 Metal Fatigue of Reactor Coolant Pressure Boundary....................

15-40 15.6 EXEMPTIONS....................................................

15-43 15.7 LICENSE RENEWAL COMMITMENTS..............................

15-43

Revision 2511/26/14 KPS USAR 15-iii Table 15.7-1 License Renewal Commitments...............................

15-44 Chapter 15: Programs and Activities that Manage the Effects of Aging List of Tables Table Title Page

Revision 2511/26/14 KPS USAR 15-iv Intentionally Blank

Revision 2511/26/14 KPS USAR 15-1 CHAPTER 15 PROGRAMS AND ACTIVITIES THAT MANAGE THE EFFECTS OF AGING

15.1 INTRODUCTION

The application for a Renewed Operating License is required by 10 CFR 54.21(d) to include a USAR Supplement. This appendix comprises the USAR supplement and includes the following sections:

  • Section 15.2 contains a listing of the aging management programs and the status of the program at the time the Renewed Operating License (ROL) was issued.
  • Section 15.3 contains a description of the programs for managing the effects of aging.
  • Section 15.4 contains the evaluation of Time-limited Aging Analyses (TLAAs) for the period of extended operation.
  • Section 15.5 contains a summarized description of the programs that support the TLAAs.
  • Section 15.6 contains a summarized description of the plant-specific exemptions.

The integrated plant assessment for license renewal identified new and existing aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the Current Licensing Basis (CLB) for the period of extended operation. The period of extended operation extends 20 years from the units original operating license expiration date through December 21, 2033.

15.1.1 Quality Assurance and Administrative Controls The Quality Assurance Program is described in Topical Report DOM-QA-1, Dominion Nuclear Facility Quality Assurance Program Description and implements the requirements of 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. The Quality Assurance Program is consistent with the summary in Appendix A.2 of NUREG-1800 (Reference 1). The program includes the elements of corrective action, confirmation process, and administrative controls, which are applicable to the safety-related and non-safety-related systems, structures, and components that are subject to aging management review. In many cases, existing programs were found to be adequate for managing aging effects during the period of extended operation. Generically the three elements are applicable as follows:

Revision 2511/26/14 KPS USAR 15-2 Corrective Actions The Corrective Action Program is implemented in accordance with the requirements of 10 CFR 50, Appendix B and Topical Report DOM-QA-1. A single corrective actions process is applied regardless of the safety classification of the structure or component. Corrective actions are implemented through the initiation of a condition report in accordance with Dominion Fleet and plant procedures for actual or potential problems, including unexpected plant equipment degradation, damage, failure, malfunction or loss. Site documents that implement aging management programs for license renewal direct that a condition report be prepared in accordance with these procedures whenever non-conforming conditions are found, i.e., the acceptance criteria are not met.

Equipment deficiencies are corrected through the initiation of work orders in accordance with plant procedures. Plant procedures also require that condition reports be initiated when equipment deficiencies are discovered or the need for corrective maintenance is identified.

Confirmation Process The focus of the confirmation process is on the follow-up actions that must be taken to verify effective implementation of corrective actions. The measure of effectiveness is in terms of correcting the adverse condition and precluding repetition of significant conditions adverse to quality. Plant procedures include provisions for timely evaluation of adverse conditions and implementation of any corrective actions required, including root cause determinations and prevention of recurrence where appropriate (e.g., significant conditions adverse to quality). These procedures provide for tracking, coordinating, monitoring, reviewing, verifying, validating, and approving corrective actions, to ensure effective corrective actions are taken. The corrective action process is also monitored for potentially adverse trends. The existence of an adverse trend due to recurring or repetitive adverse conditions will result in the initiation of a condition report. The aging management programs required for license renewal would also uncover any unsatisfactory condition due to ineffective corrective action.

Since the same 10 CFR 50, Appendix B corrective actions and confirmation process is applied for nonconforming safety-related and non-safety-related structures and components subject to aging management review for license renewal, the Corrective Action Program is consistent with the NUREG-1801 (Reference 2) elements.

Administrative Controls Administrative controls procedures provide a formal review and approval process on procedures and other forms of administrative control documents, as well as guidance on classifying documents into the proper document type.

Revision 2511/26/14 KPS USAR 15-3 15.1.2 Operating Experience Plant-specific and industry operating experience, including past corrective actions resulting in process enhancements, was considered in development of the aging management programs.

This information provides objective evidence that the effects of aging have been, and will continue to be, adequately managed. The implementing procedures for the review of operating experience provides for incorporating additional plant-specific and industry operating experience into the aging management programs to ensure continued program effectiveness.

15.1 References

1. NUREG-1800, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants, Rev. 1 U.S. Nuclear Regulatory Commission, September 2005.
2. NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.

15.2 AGING MANAGEMENT 15.2.1 Aging Management Programs The aging management programs for Kewaunee Power Station are described in the sections listed below. The list identifies the implementation status of the programs when the ROL was issued.

Existing programs were either fully or partially implemented at the time the ROL was issued. Partially implemented programs require enhancement for full implementation. Programs that are not existing programs need to be developed before being implemented. The implementation status of the listed programs will change as new programs are developed and enhancements to existing programs are completed.

1. Alloy 600 Inspections [Section 15.3.1] [Existing]
2. ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD [Section 15.3.2]

[Existing - Requires Enhancement]

3. ASME Section XI, Subsection IWE [Section 15.3.3] [Existing]
4. ASME Section XI, Subsection IWF [Section 15.3.4] [Existing]
5. Bolting Integrity [Section 15.3.5] [Existing - Requires Enhancement]
6. Boric Acid Corrosion [Section 15.3.6] [Existing]
7. Buried Piping and Tanks Inspection [Section 15.3.7][Existing - Requires Enhancement]
8. Closed-Cycle Cooling Water System [Section 15.3.8] [Existing - Requires Enhancement]

Revision 2511/26/14 KPS USAR 15-4

9. Compressed Air Monitoring [Section 15.3.9] [Existing - Requires Enhancement]
10. External Surfaces Monitoring [Section 15.3.10] [Existing - Requires Enhancement]
11. Fire Protection [Section 15.3.11] [Existing - Requires Enhancement]
12. Flow-Accelerated Corrosion [Section 15.3.12] [Existing]
13. Flux Thimble Tube Inspection [Section 15.3.13] [Existing]
14. Fuel Oil Chemistry [Section 15.3.14] [Existing - Requires Enhancement]
15. Fuel Oil Tank Inspections [Section 15.3.15] [Existing - Requires Enhancement]
16. Inspection of Overhead Heavy Load and Refueling Handling Systems [Section 15.3.16]

[Existing - Requires Enhancement]

17. Lubricating Oil Analysis [Section 15.3.17] [Existing]
18. Metal Enclosed Bus [Section 15.3.18] [Existing - Requires Enhancement]
19. Non-EQ Electrical Cables and Connections [Section 15.3.19] [To Be Developed]
20. Non-EQ Electrical Cable Connections [Section 15.3.20] [To Be Developed]
21. Non-EQ Inaccessible Medium-Voltage Cables [Section 15.3.21] [To Be Developed]
22. Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals

[Section 15.3.22] [To Be Developed]

23. Open-Cycle Cooling Water System [Section 15.3.23] [Existing - Requires Enhancement]
24. Primary Water Chemistry [Section 15.3.24] [Existing]
25. Reactor Containment Leakage Testing 10 CFR 50, Appendix J [Section 15.3.25] [Existing]
26. Reactor Head Closure Studs [Section 15.3.26] [Existing]
27. Reactor Vessel Surveillance [Section 15.3.27] [Existing - Requires Enhancement]
28. Secondary Water Chemistry [Section 15.3.28] [Existing]
29. Selective Leaching of Materials [Section 15.3.29] [To Be Developed]
30. Steam Generator Tube Integrity [Section 15.3.30] [Existing]
31. Structures Monitoring Program [Section 15.3.31] [Existing - Requires Enhancement]
32. Work Control Process [Section 15.3.32] [To Be Developed]

15.2.2 Time Limited Aging Analyses

1. Reactor Vessel Neutron Embrittlement [Section 15.4.1]
2. Metal Fatigue [Section 15.4.2]

Revision 2511/26/14 KPS USAR 15-5

3. Environmental Qualification of Electric Equipment [Section 15.4.3]
4. Containment Fatigue Analysis [Section 15.4.4]
5. Other Plant-Specific Time-Limited Aging Analyses [Section 15.4.5]

15.2.3 Time limited aging analysis Support Programs

1. Environmental Qualification (EQ) of Electric Components [Section 15.5.1] [Existing]
2. Metal Fatigue of Reactor Coolant Pressure Boundary [Section 15.5.2] [Existing - Requires Enhancements]

15.3 PROGRAMS THAT MANAGE THE EFFECTS OF AGING This section provides summaries of the programs credited for managing the effects of aging.

The Quality Assurance Program implements the requirements of 10 CFR 50, Appendix B, and is consistent with the summary in NUREG-1800 (Reference 1), Section A.2. The Quality Assurance program includes the elements of corrective action, confirmation process, and administrative controls and is applicable to the safety-related and non-safety-related systems, structures, and components that are within the scope of license renewal.

15.3.1 Alloy 600 Inspections Program Description The Alloy 600 Inspections program is a plant-specific program that consists of the applicable ten elements as described in Appendix A of NUREG-1800. The program meets the NUREG-1801 (Reference 2) expectation to have a plant-specific program for managing nickel alloy materials to comply with the applicable NRC publications and industry guidelines.

The Alloy 600 Inspections program manages the aging effects of primary water stress corrosion cracking in Alloy 600 base metal and Alloy 82/182 dissimilar metal welds and Alloy 690 base metal and Alloy 52/152 dissimilar metal welds. The program performs visual/bare metal, liquid penetrant, eddy current, and ultrasonic examinations to detect cracking of the in-scope components in accordance with the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program, which is consistent with the regulatory requirements of 10 CFR 50.55a.

Revision 2511/26/14 KPS USAR 15-6 15.3.2 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program Description The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program is an existing program that corresponds to NUREG-1801,Section XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program manages the aging effects of change in dimensions, cracking, loss of fracture toughness, loss of material, and loss of preload for the ASME Class 1, 2, and 3 piping, including piping less than four inches nominal pipe size, and components fabricated of nickel alloys, stainless steel, and steel. In addition, the program manages the aging effect of cracking for the steel reactor coolant pump motor flywheels.

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program performs visual, surface, ultrasonic, and eddy current examinations based on the inspection extent, schedule, and techniques specified in Tables IWB-2500-1, IWC-2500-1, and IWD-2500-1, respectively, for Class 1, 2, and 3 components.

Commitments

  • Aging Management of Reactor Vessel Internals The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program will be enhanced 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 to augment the current inspections.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 1.

  • Aging Management of Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program will be enhanced to include identification of the limiting susceptible cast austenitic stainless steel reactor vessel internals components from the standpoint of thermal aging susceptibility, neutron fluence, and cracking. For each identified component, a plan will be developed, which accomplishes aging management through either a supplemental examination or a component-specific evaluation. The plan will be submitted for NRC review and approval not less than 24 months before entering the period of extended operation.

Revision 2511/26/14 KPS USAR 15-7 This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 2.

  • Aging Management of Small Bore Circumferential Welds The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program will be enhanced to ensure that for Examination Category B-J, Item No. B9.21, eight ASME Class 1 small-bore circumferential welds will receive volumetric and surface examinations during each 10-year lSI inspection interval during the period of extended operation.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 42.

  • Aging Management of Small Bore Socket Welds The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program will be enhanced to ensure ten volumetric examinations of ASME Class 1 small-bore socket welds will be performed using a demonstrated, nuclear-industry endorsed, inspection methodology that can detect cracking within the specified examination volume, if a methodology becomes available. In the event that a demonstrated, nuclear-industry endorsed, inspection methodology is not available, destructive examinations of socket welds will be substituted for volumetric non-destructive examinations. Each destructive weld examination will be considered equivalent to performing two volumetric weld examinations, such that a maximum of five destructive examinations will be performed.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 43.

15.3.3 ASME Section XI, Subsection IWE Program Description The ASME Section XI, Subsection IWE program is an existing program that corresponds to NUREG-1801,Section XI.S1, ASME Section XI, Subsection IWE.

The ASME Section XI, Subsection IWE program manages aging effects in the Class MC metal Reactor Containment Vessel, including loss of material, cracking and loss of sealing for steel, stainless steel and elastomers.

The ASME Section XI, Subsection IWE program consists of condition monitoring examinations of metal pressure boundary surfaces and welds, penetrations, integral attachments and their welds, moisture barriers, and pressure-retaining bolted connections. The program requirements include scope, schedule, examination methods, and acceptance standards for components. The ASME Section XI, Subsection IWE program requires periodic visual examination (general visual and VT-3) of all pressure-retaining components and augmented examinations of surfaces likely to experience accelerated degradation and aging. Augmented examinations include a VT-1 visual exam and possible ultrasonic thickness measurements.

Revision 2511/26/14 KPS USAR 15-8 The ASME Section XI, Subsection IWE program inspections have been effective in maintaining the integrity of the Reactor Containment Vessel pressure boundary and structural integrity and ensuring that aging effects are discovered and repaired before the loss of structure or component intended functions.

15.3.4 ASME Section XI, Subsection IWF Program Description The ASME Section XI, Subsection IWF program is an existing program that corresponds to NUREG-1801,Section XI.S3, ASME Section XI, Subsection IWF.

The ASME Section XI, Subsection IWF program manages the aging effects of loss of material and loss of mechanical function for the in-scope steel supports and hangers.

The ASME Section XI, Subsection IWF program performs visual examinations of Class 1, Class 2, and Class 3 component supports. The program support and hanger inspections fulfill the requirements specified by 10 CFR 50.55a(g). Removal, repair, monitoring, or analytical evaluation are identified as acceptable corrective action options.

15.3.5 Bolting Integrity Program Description The Bolting Integrity program is an existing program that corresponds to NUREG-1801,Section XI.M18, Bolting Integrity.

The Bolting Integrity program manages the aging effects of cracking, loss of material, and loss of preload for bolting/fasteners.

The Bolting Integrity program relies on recommendations for a comprehensive bolting integrity program as delineated in NUREG-1339, (Reference 3), and industry recommendations as delineated in the Electric Power Research Institute (EPRI) NP-5769, (Reference 4), with the exceptions noted in NUREG-1339. The Bolting Integrity program addresses three subject areas:

proper assembly of bolted joints through instructions/procedures; the procurement, receipt and storage of bolting materials; and the training of plant personnel with respect to bolting issues. The program addresses bolting associated with pressure boundary, mechanical, and high strength bolting for component supports. Maintenance procedures provide detailed instructions for removal and installation of bolted pressure boundary closures, and provide generic guidance on proper bolting practices.

Commitments

  • Bolting Program Improvements

Revision 2511/26/14 KPS USAR 15-9 The Bolting Integrity program will be enhanced to further incorporate applicable EPRI and industry bolting guidance. Topic enhancements will include proper joint assembly, torque values, gasket types, use of lubricants, and other bolting fundamentals.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 3.

15.3.6 Boric Acid Corrosion Program Description The Boric Acid Corrosion program is an existing program that corresponds to NUREG-1801,Section XI.M10, Boric Acid Corrosion.

The Boric Acid Corrosion program manages the aging effect of loss of material for the aluminum, copper alloys, electrical conductor material, and steel for the in-scope systems, structures, and components that are subject to borated water leakage. The program performs visual inspections to identify boric acid leakage. The scope of the program includes those systems and components, which are potential sources of borated water leakage and potential targets of borated water leakage.

Generic Letter 88-05, (Reference 5) and industry guidance are used as reference documents for providing guidance for evaluating the severity of boric acid leakage and for determining the appropriate corrective actions.

The Boric Acid Corrosion program is supported by the inspection opportunities afforded by other programs, including inspections performed during plant operator rounds, system engineer walkdowns, inservice inspection pressure tests and inspections, and Reactor Containment Vessel inspections performed during power operation and immediately following unit shutdown.

15.3.7 Buried Piping and Tanks Inspection Program Description The Buried Piping and Tanks Inspection program is an existing program that corresponds to NUREG-1801,Section XI.M34, Buried Piping and Tanks Inspection.

The Buried Piping and Tanks Inspection program manages the aging effect of loss of material for the buried steel (including cast iron) and stainless steel components such as piping, valves, and tanks in the in-scope buried portions of the Circulating Water System, Emergency Diesel Generators fuel oil system, Technical Support Center Diesel Generator fuel oil system, and Fire Protection System.

Revision 2511/26/14 KPS USAR 15-10 The following materials are utilized in buried applications with the associated protective measures:

  • Steel (including cast iron)/coated,
  • Steel/coated and wrapped,
  • Steel/uncoated, and
  • Stainless steel/coated and wrapped The program includes the use of preventive measures, such as coatings and wrappings and performs opportunistic and deliberate visual inspections of the external surface of a representative sample of the material/protective measures combinations of the in-scope buried piping and components. The program inspections inspect for evidence of damaged wrapping; coating defects, such as coating perforation, holidays, or other damage; and evidence of loss of material on the external surface of the piping or component.

Commitments

  • Program Inspection Implementation The Buried Piping and Tanks Inspection program will be enhanced to perform visual inspections of a representative sample of material/protective measure combinations for in-scope buried piping and tanks.

Visual inspections of the external surface of the components will be performed to identify damaged wrapping (if present), degraded or damaged coating (if present), and evidence of loss of material. Each piping inspection will include a minimum of ten linear feet of piping.

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 ten years and the second ten years of the period of extended operation.
  • 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 ten years and the second ten years of the period of extended operation.

Revision 2511/26/14 KPS USAR 15-11

  • 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 ten years and the second ten years of the period of extended operation. The inspections will be performed in the non-catholically 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 ten 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 ten 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 ten years of the period of extended operation.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 4.

15.3.8 Closed-Cycle Cooling Water System Program Description The Closed-Cycle Cooling Water System program is an existing program that corresponds to NUREG-1801,Section XI.M21, Closed-Cycle Cooling Water System.

The Closed-Cycle Cooling Water System program manages the aging effects of cracking, loss of material, and reduction of heat transfer for the steel, stainless steel, and copper alloys in the piping, heat exchangers, and other components in the Component Cooling System, Emergency Diesel Generator cooling water subsystems, and Control Room Air Conditioning System. The Component Cooling System provides cooling water to a number of heat exchangers and other equipment in other systems that are included in the scope of the program. The Closed-Cycle Cooling Water System program manages the in-scope systems with corrosion control strategies and chemistry specifications, including the use of inhibitors; and performance monitoring, including system operation monitoring, system testing, heat exchanger thermal performance testing, heat exchanger tube eddy current testing, and pump performance testing monitoring.

Revision 2511/26/14 KPS USAR 15-12 Commitments

  • Nitrate Monitoring Implement nitrate monitoring for the Component Cooling System on a frequency consistent with the existing monitoring for ammonia.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 40.

15.3.9 Compressed Air Monitoring Program Description The Compressed Air Monitoring program is an existing program that corresponds to NUREG-1801,Section XI.M24, Compressed Air Monitoring.

The Compressed Air Monitoring program manages the aging effect of loss of material for the steel, stainless steel, and copper alloy components in the Station and Instrument Air System and the air start subsystems for the Emergency Diesel Generators.

The Compressed Air Monitoring program performs air quality sampling, visual inspections, and periodic testing to verify the adequacy of the air quality and to detect air leakage. The program addresses the requirements of NRC Generic Letter 88-14 (Reference 6).

Commitments

  • Implementation of Industry Guidelines The Compressed Air Monitoring program will be enhanced to incorporate the compressed air system testing and maintenance recommendations from ASME OM-S/G-1998, Part 17 (Reference 7) and EPRI TR-108147 (Reference 8) and to identify these documents as part of the program basis.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 5.

15.3.10 External Surfaces Monitoring Program Description The External Surfaces Monitoring program is an existing program that corresponds to NUREG-1801,Section XI.M36, External Surfaces Monitoring.

The External Surfaces Monitoring program manages the aging effects of change in material properties, cracking, delamination, loss of material, and hardening and loss of strength by visually inspecting the external surfaces of in-scope components, piping, supports, structural members, and structural commodities, whether they are constructed of metal or elastomers.

Revision 2511/26/14 KPS USAR 15-13 The program credits the activities of Operations, Engineering and Health Physics to perform the external surface visual inspections. Nuclear Auxiliary Operators perform rounds each shift in accessible plant areas and perform general inspections, which include specific inspection details related to monitoring equipment aging. System Engineers perform comprehensive visual inspections during walkdowns of plant systems and components during both normal operation and refueling outages. The guidance for System Engineer walkdowns provides a walkdown checklist of attributes to be observed, which includes inspection criteria related to aging management. Health Physics technicians routinely perform radiological surveys in the radiologically controlled areas of the plant and look for any evidence of boron precipitation and active radioactive system leaks observed while performing these surveys.

The External Surfaces Monitoring program includes the inspection of areas of the plant containing in-scope equipment or structural commodities requiring aging management that are infrequently accessed because there is no operational need for plant personnel to access the area or the stay times in the area are limited.

Commitments

  • Infrequently Accessed Areas Inspections The External Surfaces Monitoring program will be enhanced to inspect the accessible external surfaces of in-scope components, piping, supports, structural members, and structural commodities, in the infrequently accessed areas, consistent with the criteria used in other plant areas.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 6.

  • Inspections and Walkdowns Training The External Surfaces Monitoring program will be enhanced to provide training for Operations, Engineering, and Health Physics personnel performing the program inspections and walkdowns. The training will address the requirements of the External Surfaces Monitoring program for license renewal, the need to document the identified conditions with sufficient detail to support monitoring and trending the aging effects, and the aging effects monitored by the program and how to identify them.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 7.

15.3.11 Fire Protection Program Description The Fire Protection program is an existing program that corresponds to NUREG-1801, Sections XI.M26, Fire Protection and XI.M27, Fire Water System.

Revision 2511/26/14 KPS USAR 15-14 The Fire Protection program manages the aging effects of change in material properties, cracking, delamination, increased hardness, loss of material, loss of sealing, loss of strength, shrinkage, and spalling for the fire protection components and features.

The Fire Protection program performs chemical treatment and periodic flushing of the water-based fire suppression system and periodic inspection and testing of the water-based, CO2, and halon fire suppression systems. The program also performs visual inspections of fire barriers, fire barrier penetrations and seals, fire barrier expansion joints, doors, fire wraps, and the reactor coolant pump oil collection system.

Commitments

  • Inspect or Replace Fire Sprinklers The Fire Protection program will be enhanced to test a representative sample of sprinkler heads or to replace all affected sprinkler heads in accordance with the requirements of NFPA 25 (Reference 9).

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 8.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 9.

  • Reactor Coolant Pump Oil Collection System Inspections The Fire Protection program inspections of the reactor coolant pump oil collection system will be revised to include additional inspection criteria for the visual inspection of the system and to perform a one-time inspection of the internal surfaces of the reactor coolant pump oil collection tank.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 10.

15.3.12 Flow-Accelerated Corrosion Program Description The Flow-Accelerated Corrosion program is an existing program that corresponds to NUREG-1801,Section XI.M17, Flow-Accelerated Corrosion.

The Flow-Accelerated Corrosion program manages the aging effect of wall thinning, thus assuring that the structural integrity of all steel (carbon or low-alloy) piping and components

Revision 2511/26/14 KPS USAR 15-15 containing high-energy fluids (two phase as well as single phase) is maintained. The program applies to both safety-related and non-safety-related components.

The Flow-Accelerated Corrosion program is based on EPRI 1011838, (Reference 10), and predicts, detects, and monitors FAC in plant piping and other pressure retaining components. The program (a) conducts an analysis to determine critical locations using CHECWORKS software, (b) performs limited baseline inspections to determine the extent of wall thinning at those locations, and (c) performs follow-up inspections to confirm the predictions, or repairs or replacements of piping and components as necessary. CHECWORKS is a predictive computer program that uses past inspection data to predict wear rates.

15.3.13 Flux Thimble Tube Inspection Program Description The Flux Thimble Tube Inspection program is an existing program that corresponds to NUREG-1801,Section XI.M37, Flux Thimble Tube Inspection.

The Flux Thimble Tube Inspection program manages the aging effect of loss of material of the flux thimble tube wall.

The flux thimble tubes provide a path for the incore neutron flux monitoring system detectors and form part of the RCS pressure boundary. Flux thimble tubes are subject to loss of material (primarily at the fuel assembly lower nozzle) where flow-induced fretting causes wear at discontinuities in the path from the reactor vessel instrument nozzle to the fuel assembly instrument guide tube. The eddy current testing inspection method is used to monitor for loss of material primarily due to wear of the flux thimble tubes. Program requirements have been established, including inspection methodology, tube wear acceptance criterion, prediction of future wall loss rates, inspection frequency, corrective actions, and maintenance of program documents and test results. The program implements the recommendations of NRC Bulletin 88-09, (Reference 11), as identified in Wisconsin Public Service Corporation (WPSC) letter NRC-88-2 dated January 6, 1989 (Reference 12).

15.3.14 Fuel Oil Chemistry Program Description The Fuel Oil Chemistry program is an existing program that corresponds to NUREG-1801,Section XI.M30, Fuel Oil Chemistry.

The Fuel Oil Chemistry program manages the aging effect of loss of material on piping and components in the systems that supply fuel oil from the storage tanks to the Emergency Diesel Generators and the Technical Support Center Diesel Generator by providing reasonable assurance that potentially harmful contaminants are maintained at low concentrations.

Revision 2511/26/14 KPS USAR 15-16 The Fuel Oil Chemistry program samples the fuel oil for the existence of contaminants such as water and microbiological organisms, and verifies the quality of new oil before its introduction into the diesel generator fuel oil storage tanks. The program defines specific acceptance criteria for contaminant concentrations, which reflect ASTM guidelines for parameters that maintain contaminant concentrations below unacceptable levels. Should unacceptable indications be observed, the condition is documented and evaluated using the Corrective Action Program.

Commitments

  • Testing Criteria Quarterly laboratory testing of fuel oil samples for water, sediment and particulates will be performed on the Emergency Diesel Generators and Technical Support Center Diesel Generator day tank. The testing acceptance criteria will be consistent with the requirements specified in ASTM D975-06b (Reference 13) for water and sediment and ASTM D6217 (Reference 14) for particulates.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 30 15.3.15 Fuel Oil Tank Inspections Program Description The Fuel Oil Tank Inspections program is an existing program that corresponds to NUREG-1801,Section XI.M30, Fuel Oil Chemistry.

The Fuel Oil Tank Inspections program manages the aging effect of loss of material internal to the underground diesel generator fuel oil storage tanks. The program periodically drains, cleans, and inspects (both visual inspections and nondestructive examinations) the internal surfaces of the fuel oil storage tanks to ensure that there is no loss of intended function. The program's schedule for cleaning and inspection is aligned with the recommendations of Regulatory Guide 1.137, Revision 1 (Reference 15).

Commitments

  • Fuel Oil Storage Tanks Inspection and Cleaning The Fuel Oil Tank Inspections program will be enhanced to provide guidance for the periodic draining, cleaning and inspection activities.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 11.

Revision 2511/26/14 KPS USAR 15-17 15.3.16 Inspection of Overhead Heavy Load and Refueling Handling Systems Program Description The Inspection of Overhead Heavy Load and Refueling Handling Systems program is an existing program that corresponds to NUREG-1801,Section XI.M23, Inspection of Overhead Heavy Load and Light Load (Related to Refueling) Handling Systems.

The Inspection of Overhead Heavy Load and Refueling Handling Systems program manages the aging effect of loss of material due to general corrosion and rail wear for the in-scope steel cranes, trolleys, bridges and rails. The program is implemented through periodic visual inspections of the crane, trolley, bridge and rail structural members.

Commitments

  • Inspection Criteria The Inspection of Overhead Heavy Load and Refueling Handling Systems program will be enhanced to clarify the requirements of visual inspection of structural members, including structural bolting, of the in-scope heavy load and refueling handling cranes and associated equipment.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 12.

15.3.17 Lubricating Oil Analysis Program Description The Lubricating Oil Analysis program is an existing program that corresponds to NUREG-1801,Section XI.M39, Lubricating Oil Analysis Program.

The Lubricating Oil Analysis program manages the aging effects of loss of material and reduction of heat transfer for aluminum, copper alloys, stainless steel, and steel mechanical system components within the scope of license renewal.

The Lubricating Oil Analysis program maintains oil systems contaminants (primarily water and particulates) within acceptable limits, thereby preserving an environment that is not conducive to loss of material or reduction of heat transfer. Lubricating oil testing activities include sampling and analysis of lubricating oil for detrimental contaminants, such as water, particulates, and metals.

15.3.18 Metal Enclosed Bus Program Description The Metal Enclosed Bus program is an existing program that corresponds to NUREG-1800,Section XI.E4, Metal Enclosed Bus.

Revision 2511/26/14 KPS USAR 15-18 The Metal Enclosed Bus program manages the aging effects of reduced insulation resistance, electrical failure and loosening of bolted connections for non-segregated metal enclosed bus (MEB) and internal components within the scope of license renewal.

The program performs visual inspections of the in-scope MEB for cracks, corrosion, foreign debris, excessive dust buildup, and evidence of water intrusion, and performs visual inspections of component insulation for surface anomalies, such as discoloration, cracking, chipping or surface contamination.

The program performs visual inspections of a sample of accessible MEB bolted connections that are covered with heat shrink tape, sleeving, insulated boots, etc., for surface anomalies, such as discoloration, cracking, chipping or surface contamination.

The inspection of all MEB will be completed prior to the period of extended operation and will be repeated every ten years thereafter.

The inspection of the sample of bolted connections will be completed prior to the period of extended operation and will be repeated every five years thereafter.

Commitments

  • Additional Visual Inspections and Corrective Actions The Metal Enclosed Bus program will be enhanced to include augmented periodical visual inspections of the MEB internal surfaces, bus supports, bus insulation, taped joints and boots for signs of degradation or aging.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 13.

15.3.19 Non-EQ Electrical Cables and Connections Program Description The Non-EQ Electrical Cables and Connections program is a new program that will correspond to NUREG-1801,Section XI.E1, Electrical Cable and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.

The Non-EQ Electrical Cables and Connections program will manage the aging effects of reduced insulation resistance and electrical failure of accessible non-EQ electrical cables and connections within the scope of license renewal that are subject to an adverse localized environment.

The program will perform a plant walkdown to visually inspect for accessible electrical cables and connections installed in an adverse localized environment. Should an adverse localized environment be observed, a representative sample of electrical cables and connections installed within that environment will be visually inspected for the aging mechanisms associated with

Revision 2511/26/14 KPS USAR 15-19 jacket surface anomalies, such as embrittlement, discoloration, cracking, or surface contamination.

The first inspection will be completed prior to the period of extended operation, and will be repeated every ten years thereafter.

Commitments

  • Program Implementation The Non-EQ Electrical Cables and Connections program will be established.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 14.

15.3.20 Non-EQ Electrical Cable Connections Program Description The Non-EQ Electrical Cable Connections program is a new program that will correspond to NUREG-1801,Section XI. E6, Electrical Cable Connections Not Subject To 10 CFR 50.49 Environmental Qualification Requirements (Revised).

The Non-EQ Electrical Cable Connections program will manage the aging effect of loosening of bolted connections for non-EQ electrical cable connections within the scope of license renewal.

The program will perform a one-time inspection, on a sampling basis, to confirm the absence of loosening of bolted connections due to thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion and oxidation.

A representative sample of non-EQ electrical cable connections (metallic parts) associated with cables within the scope of license renewal will be tested at least once prior to the period of extended operation to provide an indication of the integrity of the cables connections.

Commitments

  • Program Implementation The Non-EQ Electrical Cable Connections program will be established.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 15.

Revision 2511/26/14 KPS USAR 15-20 15.3.21 Non-EQ Inaccessible Medium-Voltage Cables Program Description The Non-EQ Inaccessible Medium-Voltage Cables program is a new program that will correspond to NUREG-1801,Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.

The Non-EQ Inaccessible Medium-Voltage Cables program will manage the aging effects of localized damage and breakdown of insulation leading to electrical failure for non-EQ, inaccessible, low and medium-voltage (> 480 volts) cables within the scope of license renewal that are subject to an adverse localized environment caused by exposure to significant moisture.

Significant moisture is defined as periodic exposures to moisture that last more than a few days, e.g., cable in standing water. Periodic exposures to moisture that last less than a few days, i.e., normal rain and drain, are not significant. An adverse localized environment is a condition in a limited plant area that is significantly more severe than the specified service environment for the cables (power, control, and instrumentation) and connections. An adverse localized environment is significant if it could appreciably increase the rate of aging of a component, or has an immediate adverse effect on operability.

The program will inspect the in-scope manhole east of the tertiary auxiliary transformer, the pulling pit, and the EDG fuel oil storage tank access manholes for water collection that could cause the in-scope cables to be exposed to significant moisture and will remove water, if required.

The program will perform a test on the in-scope non-EQ inaccessible low and medium-voltage cables to provide an indication of the condition of the conductor insulation.

Inspection of the in-scope manhole east of the tertiary auxiliary transformer, the pulling pit, and the EDG fuel oil storage tank access manholes for water collection will be performed based on actual plant experience with water accumulation in the manhole. However, the inspection will be performed at least every two years. The first inspection for license renewal will be performed prior to the period of extended operation.

Testing of the in-scope inaccessible low and medium-voltage cables exposed to significant moisture will be performed prior to the period of extended operation, and the tests will be repeated every ten years thereafter.

Commitments

  • Program Implementation The Non-EQ Inaccessible Medium-Voltage Cables program will be established.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 16.

Revision 2511/26/14 KPS USAR 15-21 15.3.22 Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals Program Description The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals program is a new program that will correspond to NUREG-1801,Section XI.E2, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements.

The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals program will manage the aging effects of reduced insulation resistance and electrical failure for electrical cables and connections subject to sensitive, high-voltage, low-level signals installed in nuclear instrumentation and radiation monitoring circuits within the scope of license renewal that are subject to an adverse localized environment.

The program will perform a proven cable system test for detecting deterioration of the insulation system (such as insulation resistance tests, time domain reflectometry tests, or other testing judged to be effective in determining cable insulation condition) for those electrical cables and connections disconnected during calibration, or will review the results and findings of calibrations for those electrical cables that remain connected during the calibration process.

The first tests and calibration reviews will be completed prior to the period of extended operation and will be repeated every ten years thereafter.

Commitments

  • Program Implementation The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals program will be established.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 17.

15.3.23 Open-Cycle Cooling Water System Program Description The Open-Cycle Cooling Water System program is an existing program that corresponds to NUREG-1801,Section XI.M20, Open-Cycle Cooling Water System.

The Open-Cycle Cooling Water System program manages the aging effects of loss of material and reduction in heat transfer of open-cycle cooling water systems components. The scope of the program includes the components fabricated of copper alloys, stainless steel, and steel in the Service Water System and the portions of the Circulating Water System, which interface with and support the operation of the Service Water System.

Revision 2511/26/14 KPS USAR 15-22 The Open-Cycle Cooling Water System program performs chemical treatment, visual inspections, nondestructive examinations, heat exchanger thermal performance testing, and maintenance, which includes flushing and cleaning, to manage aging of the open-cycle cooling water systems.

Commitments

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 18.

15.3.24 Primary Water Chemistry Program Description The Primary Water Chemistry program is an existing program that corresponds to NUREG-1801,Section XI.M2, Water Chemistry.

The Primary Water Chemistry program manages the aging effects of cracking, loss of material, and reduction of heat transfer for nickel alloys, stainless steel and steel components.

The intent of the Primary Water Chemistry program is to minimize corrosion in order to maintain the primary system pressure boundary integrity.

The Primary Water Chemistry program relies on the periodic monitoring and control of known detrimental contaminants such as chloride, fluoride, dissolved oxygen and sulfate concentrations below the levels known to result in cracking, loss of material, and reduction of heat transfer. Primary water chemistry control is based on the industry guidelines for primary water chemistry.

15.3.25 Reactor Containment Leakage Testing 10 CFR 50, Appendix J Program Description The Reactor Containment Leakage Testing 10 CFR 50, Appendix J program is an existing program that corresponds to NUREG-1801,Section XI.S4, 10 CFR Part 50, Appendix J.

The Reactor Containment Leakage Testing 10 CFR 50, Appendix J program manages the aging effects of cracking, loss of leak tightness, loss of material, loss of sealing and leakage through the Reactor Containment Vessel, including the systems penetrating the Reactor Containment Vessel, penetrations, isolation valves, fittings and access openings made of elastomers, stainless steel, and steel to detect degradation of the pressure boundary.

Revision 2511/26/14 KPS USAR 15-23 The Reactor Containment Leakage Testing 10 CFR 50, Appendix J program is implemented using Option B of Appendix J. The regulatory basis for the program includes NRC Regulatory Guide 1.163 (Reference 16), and NEI 94-01 (Reference 17).

15.3.26 Reactor Head Closure Studs Program Description The Reactor Head Closure Studs program is an existing program that corresponds to NUREG-1801,Section XI.M3, Reactor Head Closure Studs.

The Reactor Head Closure Studs program manages the aging effects of cracking and loss of material for the reactor head closure stud assembly including nuts and washers and for the threads in the reactor vessel flange. The program includes preventive measures to mitigate cracking and loss of material and visual or volumetric examinations to monitor this degradation. The preventive measures implemented by the program are consistent with the measures identified in NRC Regulatory Guide 1.65 (Reference 18). The Reactor Head Closure Studs program visual and volumetric examinations are performed in accordance with the ASME Section XI 1998 Code Edition through the 2000 Addenda, Examination Category B-G-1.

15.3.27 Reactor Vessel Surveillance Program Description The Reactor Vessel Surveillance program is an existing program that corresponds to NUREG-1801,Section XI.M31, Reactor Vessel Surveillance.

The Reactor Vessel Surveillance program manages the aging effects of loss of fracture toughness due to irradiation embrittlement of the reactor pressure vessel low alloy steel material.

Monitoring methods are in accordance with 10 CFR 50, Appendix H. This program includes surveillance capsule removal and specimen mechanical testing/evaluation, radiation analysis, development of pressure-temperature limits, and determination of low-temperature overpressure protection (LTOP) set points. The program ensures the reactor vessel materials meet the fracture toughness requirements of 10 CFR 50, Appendix G, and meet the requirements of Pressurized Thermal Shock (PTS) and upper shelf energy in 10 CFR 50.60 and 10 CFR 50.61, respectively, as modified by the exemption granted to utilize Master Curve methodology.

Commitments

  • Operating Restrictions The Reactor Vessel Surveillance program will be enhanced to include the applicable limitations on operating conditions to which the surveillance capsules were exposed (e.g. neutron flux, spectrum, irradiation temperature, etc.).

Revision 2511/26/14 KPS USAR 15-24 This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 19.

  • Storage of Pulled and Tested Surveillance Capsules The Reactor Vessel Surveillance program will be enhanced to include requirements for storing, and possible recovery, of tested and untested capsules (removed from the Reactor Vessel after August 31, 2000).

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 20.

15.3.28 Secondary Water Chemistry Program Description The Secondary Water Chemistry program is an existing program that corresponds to NUREG-1801,Section XI.M2, Water Chemistry.

The Secondary Water Chemistry program manages the aging effects of cracking, loss of material, and reduction of heat transfer for copper alloys, nickel alloys, stainless steel and steel components.

The intent of the Secondary Water Chemistry program is to minimize the corrosion of secondary-side components to attain their maximum useful life and minimize the fouling of heat transfer surfaces to achieve maximum plant efficiency.

The Secondary Water Chemistry program relies on periodic monitoring and control of known detrimental contaminants such as chloride, dissolved oxygen and sulfate, to ensure the concentrations are below the levels known to result in cracking, loss of material, or reduction of heat transfer. Secondary water chemistry control is based on the industry guidelines for secondary water chemistry.

15.3.29 Selective Leaching of Materials Program Description The Selective Leaching of Materials program is a new program that will correspond to NUREG-1801, XI.M33, Selective Leaching of Materials.

The Selective Leaching of Materials program will manage the aging effects of loss of material on internal and external surfaces of in-scope components such as piping, pumps, valves, and heat exchanger components made of steel (cast iron), and copper alloys (brass, bronze, or aluminum-bronze).

The program will perform a one-time visual inspection, and hardness measurement or qualitative examination such as resonance when struck by another object, scraping, or chipping, as appropriate, of selected components within the scope of license renewal for loss of material due to selective leaching.

Revision 2511/26/14 KPS USAR 15-25 The inspection, and hardness measurement or qualitative examination, as appropriate, will be performed prior to the period of extended operation.

Commitments

  • Program Implementation The Selective Leaching of Materials program will be established.

The commitment is identified in Table 15.7-1 License Renewal Commitments, Item 21.

15.3.30 Steam Generator Tube Integrity Program Description The Steam Generator Tube Integrity program is an existing program that corresponds to NUREG-1801,Section XI.M19, Steam Generator Tube Integrity.

The Steam Generator Tube Integrity program manages the aging effects of cracking and loss of material for the primary and secondary-side steam generator components fabricated of nickel alloys, stainless steel, and steel. The program is based on Technical Specification requirements, meets the intent of NEI 97-06 (Reference 19), and is credited for aging management of the tubes, tube plugs, tube sleeves, tube supports, and secondary-side components whose failure could prevent the steam generator from fulfilling its intended safety function.

Acceptance criteria for inspections performed in accordance with the Steam Generator Tube Integrity program are based on applicable regulations and standards. Corrective actions for conditions that are adverse to quality are performed in accordance with the Corrective Action Program as a part of the Quality Assurance Program.

15.3.31 Structures Monitoring Program Program Description The Structures Monitoring Program is an existing program that corresponds to NUREG-1801, Sections XI.S5, Masonry Wall Program, XI.S6, Structures Monitoring Program, and XI.S7, RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants.

The Structures Monitoring Program manages the aging effects of: (1) cracking, loss of bond, loss of material (spalling, scaling), cracks and distortion, increase in porosity and permeability, loss of strength, and reduction in concrete anchor capacity due to local concrete degradation for concrete, (2) loss of material and loss of mechanical function for steel, (3) loss of material for stainless steel and aluminum, and (4) change in material properties, cracking, increased hardness, shrinkage and loss of strength, loss of sealing, and reduction or loss of isolation function for elastomers.

Revision 2511/26/14 KPS USAR 15-26 The program relies on periodic visual inspections to monitor the condition of structures, structural elements (including component supports), miscellaneous structural commodities, and masonry walls. The program implements the requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, with the guidance of NUMARC 93-01, Revision 2 (Reference 20), and Regulatory Guide 1.160, Revision 2 (Reference 21). For masonry walls within the scope of license renewal, the Structures Monitoring Program manages aging effects based on guidance provided in IE Bulletin 80-11 (Reference 22),

and plant-specific monitoring proposed by NRC Information Notice 87-67 (Reference 23). For water-control structures within the scope of license renewal, the Structures Monitoring Program manages aging effects consistent with the guidelines of RG 1.127 (Reference 24).

Commitments

  • Define In-Scope Structural Elements The Structures Monitoring Program will be enhanced to clearly define structures, structural elements, and miscellaneous structural commodities that are in scope.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 22.

  • Evaluation Criteria The Structures Monitoring Program will be revised to include the evaluation criteria of ACI 349.3R-96, Chapter 5 (Reference 25), as the criteria to be used when evaluating conditions or findings 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 54.

  • Groundwater Monitoring The Structures Monitoring Program will be enhanced to monitor groundwater quality and verify that it remains non-aggressive to below-grade concrete.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 23.

  • Underwater Inspections The Structures Monitoring Program will be enhanced to improve criteria for detection of aging effects for the underwater visual inspections of the in-scope structures.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 24.

  • Leakage Identification and Remediation

Revision 2511/26/14 KPS USAR 15-27 Develop a plan for identification and remediation of reactor refueling cavity liner leakage to be implemented during the period of extended operation.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 33.

Develop an action plan for identification and remediation of spent fuel pool liner leakage to be implemented during the period of extended operation.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 35.

  • Concrete Testing At least one core bore sample will be taken from the waste drumming room reinforced concrete ceiling below the spent fuel pool (SFP). 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 34.

If the results of the core sample testing of the waste drumming room reinforced concrete ceiling leakage site (related to potential SFP liner leakage - Commitment 34) indicate degradation of 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 46.

If SFP liner leakage persists during the period of extended operation, an additional concrete core sample will be taken from the waste drumming room reinforced concrete ceiling below the spent fuel pool. 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 36.

Core samples will be obtained from the inside surface of a concrete wall (below the groundwater table elevation) or from the foundation basemat in the vicinity of the groundwater 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.

Revision 2511/26/14 KPS USAR 15-28 This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 44.

In the event that the chloride content in the groundwater does not decrease to below 500 ppm within the first ten years of the period of extended operation, core samples will be obtained from the inside surface of a concrete wall (below the groundwater table elevation) or from the 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 45.

15.3.32 Work Control Process Program Description The Work Control Process program is a new program that will correspond to NUREG-1801,Section XI.M32, One-Time Inspection, and Section XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components.

One-time inspections will manage the aging effects of cracking, loss of material, and reduction of heat transfer to verify the effectiveness of the Primary Water Chemistry, Secondary Water Chemistry, Closed-Cycle Cooling Water System, Fuel Oil Chemistry, and Lubricating Oil Analysis programs through inspections implemented in accordance with the work management process. The one-time inspections will be performed using NDE techniques that have been determined to be effective for the identification of potential aging effects. The program will use a representative sampling approach to verify degradation is not occurring. The sample size and location for the one-time inspections will be established to ensure that the number and scope of the inspections are sufficient to provide reasonable assurance that the aging effects will not compromise the intended functions during the period of extended operation.

The inspections of internal surfaces in miscellaneous piping and ducting components will manage the aging effects of change in material properties, cracking, hardening and loss of strength, loss of material, loss of sealing, loss of strength, and reduction of heat transfer for the in-scope structures and components through inspections implemented in accordance with the work management process. Periodic surveillance and maintenance activities will be reviewed to select appropriate inspection opportunities which represent the leading indicators used to manage these aging effects. The program will perform visual inspections of piping, piping components, ducting and other components fabricated of aluminum, copper alloys, stainless steel, and steel to detect loss of material, reduction of heat transfer, and cracking. Visual inspections will also manage the degradation of the paper filter elements in the Compressed Air System. The program will include physical manipulation of elastomeric components as a supplement to the visual

Revision 2511/26/14 KPS USAR 15-29 inspections. An enhanced VT-1 NDE examination will be performed to detect cracking of stainless steel diesel exhaust flexible connections.

Commitments

  • Program Implementation The Work Control Process program will be established.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 25.

  • Operating Experience Submit three examples of operating experience associated with the Work Control Process -

Internal Surfaces Monitoring program for NRC staff review in determining the effectiveness of the program to detect and correct the effects of aging prior to the loss of intended function.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 47.

  • One-Time Inspection of Fuel Oil Day Tanks The Work Control Process Program will provide for a one-time-inspection of the Emergency Diesel Generators (EDG) Day Tanks and the Technical Support Center Diesel Generator (TSC DG) Day Tank. An exterior surfaces ultra-sonic testing (UT) 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 31

  • Confirmatory Audit Perform an audit of the Internal Surfaces Monitoring portion of the Work Control Process program inspections to confirm that the components representing the leading indicators of aging for each of the material/environment combinations have been inspected at least once during the audit period.

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

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 50.

Revision 2511/26/14 KPS USAR 15-30 15.3 References

1. NUREG-1800, Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants, Rev. 1 U.S. Nuclear Regulatory Commission, September 2005.
2. NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.
3. NUREG-1339, Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants, U.S. Nuclear Regulatory Commission, June 1990.
4. EPRI NP-5769, Degradation and Failure of Bolting in Nuclear Power Plants, May 5, 1988.
5. NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, March 17, 1988.
6. NRC Generic Letter 88-14, Instrument Air Supply System Problems Affecting Safety-Related Equipment, August 8, 1988.
7. ASME OM-S/G-1998, Part 17, Performance Testing of instrument Air Systems Information Notice for Light-Water Reactor Power Plants.
8. EPRI TR-108147, Compressor and Instrument Air Maintenance Guide, March 1998.
9. NFPA 25, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems, 1998 Edition, National Fire Protection Association.
10. EPRI 1011838, Recommendations for an Effective Flow-Accelerated Corrosion Program (NSAC-202L-R3), May 2006.
11. NRC Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors, July 26, 1988.
12. Letter from D.C. Hintz (WPSC) to Document Control Desk (NRC), Responds to the NRC Bulletin 88-009: Thimble Tube Thinning in Westinghouse Reactors, Letter # NRC-88-2, January 6, 1989.
13. ASTM D975-06b, Standard Specification for Diesel Oil Fuels, Revision B, November 1, 2006
14. ASTM D6217, Standard Test Method for Particulate Contamination in Middle Distillate Fuels by Laboratory Filtration, August 2010.
15. NRC Regulatory Guide 1.137, Rev. 1, Fuel-Oil Systems for Standby Diesel Generators, October 1979.
16. NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, September 1995.

Revision 2511/26/14 KPS USAR 15-31

17. NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J.
18. NRC Regulatory Guide1.65, Material and Inspection for Reactor Vessel Closure Studs, October 1973.
19. NEI 97-06, Rev. 2, Steam Generator Program Guidelines, May 2005.
20. NUMARC 93-01, Rev 2, Industry Guideline For Monitoring The Effectiveness Of Maintenance At Nuclear Power Plants, April 1996.
21. NRC Regulatory Guide 1.160, Rev. 2, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.
22. IE Bulletin 80-11, Masonry Wall Design, May 8, 1980.
23. NRC Information Notice 87-67, Lessons Learned from Regional Inspections of Licensee Actions in Response to IE Bulletin 80-11, December 13, 1987.
24. NRC Regulatory Guide 1.127, Rev. 1, Inspection of Water-Control Structures Associated with Nuclear Power Plants, March 1978.
25. American Concrete Institute ACI 349.3R-96, Evaluation of Existing Nuclear Safety-Related Concrete Structures, January 1, 1996.

15.4 TIME-LIMITED AGING ANALYSES As part of the application for a renewed license, 10 CFR 54.21(c) requires that an evaluation of Time-limited Aging Analyses (TLAAs) for the period of extended operation be provided. The following TLAAs have been identified and evaluated to meet this requirement.

15.4.1 Reactor Vessel Neutron Embrittlement The calculation of neutron fluence to which reactor vessel materials are exposed is an important input to the evaluation of reactor vessel neutron embrittlement and is governed by regulatory requirements. WCAP-16641 (Reference 1) provides the calculation of Kewaunee reactor vessel neutron fluence projections to End of License Renewal (EOLR), i.e., 60 year plant lifetime, based on 52.1 Effective Full Power Years (EFPY). Neutron exposure up to Cycle 27 was based upon actual plant operating history, including power uprate that occurred during Cycle 26.

Neutron exposure projections beyond the end of Cycle 27 were based upon an operating scenario that consisted of a series of 18 month operating cycles followed by a 25 day refueling outage. The reactor was considered to be operating at full power for the entire 18 month cycle. This full power period coupled with the 25 day refueling outage resulted in a net capacity factor of 95.6% with a total operating time of 33.0 EFPY at End of Life (EOL) and 52.1 EFPY at EOLR. The neutron exposure projections were also based on the continued use of low neutron leakage fuel management.

Revision 2511/26/14 KPS USAR 15-32 Kewaunee reactor vessel surveillance Capsule T was removed in 2004 (the fifth capsule removed from the reactor) and WCAP-16641 documents the results of the fluence evaluation for the specimens. The fluence calculations concluded that Capsule T surveillance specimens received a fluence of 5.62E+19 n/cm2 (E>1.0 MeV) after irradiation to 24.6 EFPY and the peak reactor vessel clad/base metal interface fluence after 24.6 EFPY of plant operation was 2.60E+19 n/cm2 (E>1.0 MeV). The Capsule T specimens have received a fluence equivalent to slightly greater than 52.1 EFPY. The maximum vessel exposures occur on the intermediate shell base material with all other vessel materials experiencing a lower neutron exposure. Certain materials in the extended beltline (inlet nozzles, inlet and outlet nozzle to upper shell welds, upper shell forging, and intermediate shell to upper shell girth weld) are projected to receive fluence greater than the 10 CFR 50, Appendix H threshold value of 1.0E+17 n/cm2 during the 40 - 60 year operating period.

15.4.1.1 Upper Shelf Energy 10 CFR 50, Appendix G contains screening criteria that establish limits on how far the upper shelf energy (USE) values for a reactor pressure vessel material may be allowed to decrease due to neutron irradiation exposure. The regulation requires the initial USE value to be greater than 75 ft-lbs in the unirradiated condition and that the value be greater than 50 ft-lbs in the fully irradiated condition as determined by Charpy V-notch specimen testing throughout the licensed life of the plant.

Acceptable USE values have been calculated in accordance with Regulatory Guide 1.99, Revision 2 (Reference 2) to the end of the period of extended operation (52.1 EFPY). Calculated USE values for the most limiting reactor pressure vessel forging and weld materials remain greater than 50 ft-lbs.

15.4.1.2 Pressurized Thermal Shock Reactor pressure vessel beltline fluence is one of the factors used in determining the margin of acceptability of the reactor pressure vessel to pressurized thermal shock as a result of radiation embrittlement. The margin is the difference between the maximum nil ductility reference temperature in the limiting beltline material and the screening criteria established in accordance with 10 CFR 50.61(b)(2). The screening criteria for the limiting reactor vessel materials are 270°F for beltline plates, forging, and axial weld materials, and 300°F for beltline circumferential weld materials.

The materials in the reactor vessel extended beltline region have been evaluated using the 10 CFR 50.61 procedure to define RTPTS. None of the materials in the extended beltline were determined to be controlling.

For the circumferential weld metal (heat 1P3571), an exemption to 10 CFR 50.61 was granted (Reference 3) based upon use of the Master Curve method as defined in ASME Code

Revision 2511/26/14 KPS USAR 15-33 Case N-629, coupled with measured fracture toughness data using pre-cracked Charpy specimens.

WCAP-16609 (Reference 4) re-evaluated RTPTS for the circumferential weld metal using fracture toughness data determined from Capsule T specimens, and applying the methodology defined in the NRC Safety Evaluation for the exemption.

The RTTo for a fluence corresponding to EOLR (52.1 EFPY) was determined by making direct measurement of irradiated 1P3571 weld metal fracture toughness using fatigue pre-cracked Charpy surveillance specimens.

The screening criteria of 10 CFR 50.61(b)(2) are met for all beltline and extended beltline materials for a fluence value corresponding to the end of the period of extended operation (52.1 EFPY).

15.4.1.3 Pressure-Temperature Limits 10 CFR Part 50 Appendix G requires that heatup and cooldown of the reactor vessel be accomplished within established pressure-temperature limits. These limits identify the maximum allowable pressure as a function of reactor coolant temperature. As the reactor vessel becomes irradiated and its fracture toughness is reduced, the allowable pressure at low temperatures is reduced. Therefore, in order to heatup and cooldown the vessel, the reactor coolant temperature and pressure must be maintained within the limits of Appendix G as defined by the evaluation of reactor vessel neutron irradiation embrittlement.

Heatup and cooldown limit curves have been calculated using the adjusted RTNDT corresponding to the limiting beltline material of the reactor vessel for the current period of licensed operation. In accordance with 10 CFR 50, Appendix G, updated pressure-temperature limits for the period of extended operation have been developed and will be implemented by the Reactor Vessel Surveillance program prior to the period of extended operation.

15.4.2 Metal Fatigue 15.4.2.1 Fatigue of ASME Class 1 Components 15.4.2.1.1 Component Design Transient Cycles Operating experience at the plant and other Westinghouse NSSS units has demonstrated that the analyzed numbers of design basis transients are generally conservative for a 40 year life. The Metal Fatigue of Reactor Coolant Pressure Boundary program monitors transients and components to assure that actual plant operation remains bounded by the assumptions used in the design analyses. This program tracks cycles of design basis transient events and evaluates the number of occurrences against the design basis.

The number of occurrences of transient cycles monitored by the Metal Fatigue of Reactor Coolant Pressure Boundary program have been projected to the end of the period of extended operation based on past trends throughout the operating history of the plant. The projection

Revision 2511/26/14 KPS USAR 15-34 provides an estimate of transient occurrences for a 60-year plant lifetime, and provides reasonable assurance that the design basis number of transients will not be exceeded during the period of extended operation. These transients will continue to be tracked in accordance with the Metal Fatigue of Reactor Coolant Pressure Boundary program for the remaining plant life.

15.4.2.1.2 ASME Class 1 Vessels and Surge Line Piping The reactor vessel (including the control rod drive mechanism pressure housings), steam generators, pressurizer, reactor coolant pumps, and the pressurizer surge line, were analyzed for fatigue usage for the original 40-year life of the plant in accordance with ASME Code,Section III, requirements for Class 1 components using transient conditions that are representative of those expected to occur during plant operation and that are sufficiently severe or frequent to be of possible significance to component cyclic behavior. An assumed number of occurrences of each of the design transients during the plant lifetime were used as input to the design basis fatigue calculations. Evaluations have shown that the assumed number of occurrences are conservatively large and are not expected to be exceeded during the period of extended operation.

Therefore, based on these transient cycle projections, that are confirmed by continued cycle counting through the Metal Fatigue of Reactor Coolant Pressure Boundary program, the design fatigue analyses will remain valid for 60 years of plant operation.

15.4.2.1.3 Reactor Coolant Loop Piping The reactor coolant loop piping was designed to the requirements of USAS B31.1.0-1967.

Piping systems designed to this Code were evaluated for thermal expansion cycles, and a thermal expansion stress range reduction factor was to be applied if cycling was determined to be excessive. The Code allows 7000 full temperature thermal expansion cycles without penalty.

The design transients defined for the ASME Class 1 vessels are also applicable to the reactor coolant loop piping. An evaluation of these transients concluded that thermal cycling of the reactor coolant loop piping will remain well below the 7000 thermal expansion cycles allowed by USAS B31.1.0.

15.4.2.1.4 Pressurizer Lower Head and Surge Line The NRC staff has indicated, through Renewal Applicant Action Item 3.3.1.1-1 contained in WCAP-14574-A (Reference 5) safety evaluation report, that insurge/outsurge fatigue effects on the pressurizer lower head and the surge line must be evaluated for license renewal. These effects have been evaluated as part of the Metal Fatigue of Reactor Coolant Pressure Boundary program for critical locations in the pressurizer lower head and in the surge line, including the pressurizer and hot leg nozzles. The highest projected 60-year Cumulative Usage Factor (CUF) for these locations, at a pressurizer heater penetration, is less than the design limit.

Revision 2511/26/14 KPS USAR 15-35 15.4.2.1.5 Effects of Reactor Coolant Environment on Fatigue Life of ASME Class 1 Piping and Components GSI-190 addressed fatigue life of metal components and was closed by the NRC in December 1999 (Reference 6). In the closure letter, however, the NRC concluded that licensees should address the effects of the reactor coolant environment on the fatigue life of selected components as aging management programs are formulated in support of license renewal.

Environmentally-assisted fatigue (EAF) effects for the following plant-specific locations, as identified in NUREG/CR-6260 (Reference 7) for the older vintage Westinghouse plant, have been evaluated.

  • Reactor Vessel Outlet Nozzle
  • Reactor Vessel Inlet Nozzle
  • Reactor Vessel Shell and Lower Head
  • Surge Line Hot Leg Nozzle*
  • Safety Injection Cold Leg Nozzle*
  • Charging Line Nozzle*
  • Since the original design did not include the requirement for fatigue analysis of piping locations, the four piping locations required the development of specific fatigue evaluations, which have been performed based on the guidance of ASME B&PV Code,Section III. The Surge Line Hot Leg Nozzle and the Charging Line Nozzle evaluations were based on the ASME Code,Section III, 2001 edition with addenda through 2003 and the Safety Injection Cold Leg Nozzle and Residual Heat Removal System Tee at Safety Injection Accumulator Line evaluations were based on the ASME Code,Section III, 1989 edition with 1989 addenda.

The evaluation of the effects of the reactor coolant environment on fatigue usage at the NUREG/CR-6260 locations concluded that fatigue limits continue to be met for 60 years, based on projected plant operation. In addition, the Metal Fatigue of Reactor Coolant Pressure Boundary program confirms the assumptions for projected plant operation and manages fatigue for these locations through the period of extended operation.

15.4.2.2 Fatigue of Non-ASME Class 1 Components 15.4.2.2.1 Non-Class 1 Piping Non-Class 1 piping systems were designed and constructed to the requirements of USAS B31.1.0-1967. There is no general requirement in this Code for an explicit fatigue analysis.

However, piping systems are required to be evaluated for thermal expansion cycles, and a thermal

Revision 2511/26/14 KPS USAR 15-36 expansion stress range reduction factor is to be applied if cycling is excessive. The Code allows 7000 full temperature thermal expansion cycles without penalty.

With the exception of the reactor coolant hot leg sample line, all non-Class 1 piping systems remained within the design cycle limit for 60 years of operation. The reactor coolant hot leg sample line was re-analyzed and found to be acceptable for 60 years with the application of the appropriate stress range reduction factor to account for the increased number of thermal expansion cycles.

15.4.2.2.2 Auxiliary Heat Exchangers Heat exchangers in auxiliary systems were designed in accordance with ASME Code,Section III Class C and/or Section VIII rules, which do not require an explicit fatigue analysis.

However, the equipment specification for the residual heat removal, letdown, regenerative, excess letdown, and primary sample heat exchangers included thermal and pressure transient conditions as an input to the component design.

The transient occurrences specified for the design of these auxiliary heat exchangers are either conservatively large when compared to actual operating conditions, are bounded by the transient occurrences monitored, or are directly monitored by the Metal Fatigue of Reactor Coolant Pressure Boundary program.

15.4.3 Environmental Qualification of Electric Equipment Regulation 10 CFR Part 50 requires that certain categories of systems, structures and components be designed to accommodate the effects of both normal and accident environmental conditions, and that design control measures be employed to ensure the adequacy of these designs. Also, 10 CFR 50.49 specifies that electrical equipment that is important to safety and is located in a harsh environment must be qualified for the lifetime of the plant such that the equipment is capable of performing its safety function in the event of a design basis accident.

The qualification of electrical equipment in accordance with 10 CFR 50.49 involves the use of time-limited assumptions such as thermal life, total radiation dose, and component cycling.

As required by 10 CFR 50.49, electrical equipment not qualified for the current license term is to be refurbished, replaced or have their qualification extended prior to reaching the aging limits established in the evaluation. Re-analysis of aging evaluations to extend the qualifications of components is performed on a routine basis as part of the program. Important attributes for the re-analysis of aging evaluations include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria and corrective actions (if acceptance criteria are not met). Continued implementation of the Environmental Qualification (EQ) of Electric Components aging management program for the period of extended operation ensures that the requirements of 10 CFR 50.49 will continue to be met.

Revision 2511/26/14 KPS USAR 15-37 15.4.4 Containment Fatigue Analysis The design specification for the Reactor Containment Vessel provided design input for the number of temperature variations and pressurization cycles during the life of the vessel, which was assumed to be 200 temperature variations, and 40 pressurization cycles. The operating temperature of the vessel stays relatively constant during normal plant operation as the Shield Building effectively isolates the Reactor Containment Vessel from outdoor weather, and temperature variations are only expected during plant shutdown periods. The temperature variations of the Reactor Containment Vessel can be correlated to plant heat-up and cooldown cycles, which are shown to be limited to 200 over 60 years. Therefore, this assumption will remain valid through the period of extended operation. The Reactor Containment Vessel operates at essentially atmospheric pressure, and the vessel would experience a pressurization cycle during integrated leak rate testing (that is currently scheduled at 10-year intervals) or under accident conditions. Therefore, the 40 pressurization cycles specified will remain bounding for 60 years of operation.

Using these assumptions as inputs, a review of paragraph N-415.1 of Subsection B, ASME Section III-1965 W67, determined that a cyclic or fatigue analysis was not required. The number of design cycles used to demonstrate exemption from fatigue in accordance with Articles N-415 (a)-(f) will not be exceeded during the period of extended operation. Therefore, the original evaluation for exemption from fatigue for the Reactor Containment Vessel will remain valid for an additional 20 years of operation.

The penetration assemblies, including the bellows, were designed in accordance with USAS B31.1.0 Power Piping Code. No fatigue analyses or specified cyclic loading limits were identified for the penetration assemblies.

15.4.5 Other Plant-Specific Time-Limited Aging Analyses 15.4.5.1 Crane Load Cycle Limit Overhead cranes were originally designed to Specification 61 of the Electric Overhead Crane Institute (EOCI) (Reference 8). EOCI-61 did not require a specific fatigue or load-cycle analysis. However, cranes subject to the requirements of NUREG-0612 (Reference 9) were subsequently evaluated to the guidelines of Specification 70 of the Crane Manufacturers Association of America (CMAA-70) (Reference 10), which includes an evaluation of load cycles.

Cranes designated as Class A service cranes per CMAA-70, are designed for 20,000 to 100,000 cycles.

Since the expected number of lifts is significantly less than 20,000 through the period of extended operation, it was determined that these cranes are not governed by the fatigue consideration in Table 3.3.3.1.3-1 of CMAA-70.

Revision 2511/26/14 KPS USAR 15-38 15.4.5.2 Reactor Coolant Pump Flywheel The potential for crack propagation in the reactor coolant pump motor flywheel was evaluated due to the potential for flywheel failure that could inhibit pump coastdown or result in missile generation. The original flywheel crack growth analysis was updated to credit the leak-before-break analysis that results in a limited postulated break size and lower reactor coolant pump overspeed conditions, and to account for a 60-year operating life of the motor flywheel.

Following that evaluation, flywheel inspections were required every 10 years.

Subsequently, an additional evaluation provided a technical basis and risk assessment for extending the flywheel inspection interval to 20 years in order to coincide with the typical 10- to 15-year reactor coolant pump motor refurbishment schedule. The reactor coolant pump motor flywheels are currently inspected on a 20-year interval through the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program.

15.4.5.3 Leak-Before-Break 10 CFR 50, Appendix A, Criterion 4 allows for the use of leak-before-break (LBB) methodology for excluding the dynamic effects of postulated ruptures in reactor coolant system piping. The fundamental premise of the LBB methodology is that the materials used in nuclear power plant piping are sufficiently tough that even a large through-wall crack would remain stable and would not result in a double-ended pipe rupture.

The current licensing basis LBB analyses are discussed in USAR Section 4.1.3.4. The analyses were reviewed to determine whether the conclusions were affected by extending the operating life of the plant to 60 years. It was determined that, since the cyclic transient assumptions used as input to the analyses are bounding for the period of extended operation, only thermal aging embrittlement effects on cast austenitic stainless steel (CASS) components needed to be re-evaluated. Additionally, the only CASS components considered in the LBB analyses are the reactor coolant loop piping and elbows. The reactor coolant loop piping LBB analysis was re-evaluated assuming fully-aged (saturated) material properties for the CASS material and found to remain acceptable.

15.4.5.4 Reactor Vessel Underclad Cracking The issue of reactor vessel underclad cracking (cracks in the low-alloy steel vessel wall at the interface with the cladding caused by the weld deposition of the stainless steel cladding material) was addressed for the current licensing basis and for license renewal, and reviewed and approved by the NRC, in topical report WCAP-15338-A (Reference 11). The evaluation concluded that underclad cracks are of no concern relative to structural integrity of the reactor vessel for a period of 60 years.

Revision 2511/26/14 KPS USAR 15-39 15.4.5.5 Reactor Coolant Loop Piping Flaw Tolerance Evaluation NUREG-1801 identifies that CASS reactor coolant system components may be susceptible to reduced fracture toughness in a high temperature environment due to the effects of thermal aging embrittlement of the steel. Reactor coolant loop piping, valves, and pumps are constructed from CASS material. Thermal aging embrittlement of pumps and valves is managed by the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program. An evaluation of the susceptibility of loop piping to thermal aging and the potential for flaw growth in the piping due to reduced fracture toughness has been performed consistent with the recommendations of NUREG-1801,Section XI.M12, Thermal Embrittlement of Cast Austenitic Stainless Steel (CASS).

The evaluation concluded that the CASS reactor coolant loop piping has adequate fracture toughness for a minimum remaining service life of 30 years, which envelopes the period of extended operation. Therefore, there is no requirement to manage the effects of thermal aging embrittlement of CASS reactor coolant loop piping for the period of extended operation.

15.4 References

1. WCAP-16641, Revision 0, Analysis of Capsule T from the Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program, Westinghouse Electric Company, LLC, October, 2006.
2. NRC Regulatory Guide 1.99, Rev 2, Radiation Embrittlement of Reactor Vessel Materials, May 1968.
3. Letter from NRC to M. Reddemann, NMC, Kewaunee Nuclear Power Plant - Exemption from the Requirements of 10 CFR Part 50, Appendix G, Appendix H, and Section 50.61 (TAC No. MA8585), dated May 1, 2001.
4. WCAP-16609, Revision 0, Master Curve Assessment of Kewaunee Power Station Reactor Vessel Weld Metal, Westinghouse Electric Company, LLC, October, 2006.
5. WCAP-14574-A, License Renewal Evaluation: Aging Management Evaluation for Pressurizers, Westinghouse Electric Company, LLC, December, 2000.
6. Memorandum, A. C. Thadani, NRC to W. D. Travers, NRC; Closeout of Generic Safety Issue 190 Fatigue Evaluation of Metal Components for 60-Year Plant Life, dated December 26, 1999.
7. NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.
8. Electric Overhead Crane Institute (EOCI) Specification #61.
9. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.

Revision 2511/26/14 KPS USAR 15-40

10. Crane Manufacturers Association of America (CMAA) Specification #70, Specifications for Electric Overhead Traveling Cranes, 1975
11. WCAP-15338-A, Evaluation of Cracking Associated with Weld Deposited Cladding in PWR Vessels, Westinghouse Electric Company, LLC, October, 2002.

15.5 TLAA SUPPORT PROGRAMS 15.5.1 Environmental Qualification (EQ) of Electric Components Program Description The Environmental Qualification (EQ) of Electric Components program is an existing program that corresponds to NUREG-1801 (Reference 1),Section X.E1, Environmental Qualification (EQ) of Electric Components.

The program manages component thermal, radiation, and cyclical aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components not qualified for the current license term are to be refurbished or replaced, or have their qualification extended prior to reaching the aging limits established in the evaluation.

Aging evaluations for EQ components that specify a qualification of at least 40 years are considered time-limited aging analyses for license renewal.

For the period of extended operation, the necessary qualified life for equipment is an additional 20 years at the maximum normal plant service conditions to which the equipment will be exposed. However, the component lifespan necessary to reach the end of the period of extended operation (or the current operating term) may not always be achieved due to aging limitations and the variations in degradation rates of the materials used in equipment construction.

In these cases, it is acceptable to determine a qualified life of less than the length necessary to envelop the period of extended operation, as long as the equipment is replaced, refurbished, or requalified prior to end of that qualified life. Re-analysis of aging evaluations to extend the qualifications of components is performed on a routine basis as part of the program. Important attributes for the re-analysis of aging evaluations include analytical methods, data collection and reduction methods, underlying assumptions, acceptance criteria and corrective actions (if acceptance criteria are not met).

15.5.2 Metal Fatigue of Reactor Coolant Pressure Boundary Program Description The Metal Fatigue of Reactor Coolant Pressure Boundary program is an existing program that corresponds to NUREG-1801,Section X.M1, Metal Fatigue of the Reactor Coolant Pressure Boundary.

Revision 2511/26/14 KPS USAR 15-41 The Metal Fatigue of Reactor Coolant Pressure Boundary program manages the effects of fatigue for ASME Code Class 1 components. The program monitors and tracks the critical thermal and pressure transients to ensure that cycle occurrence limits are not exceeded such that the ASME Class 1 vessels and pressurizer surge line fatigue analyses assumptions are maintained.

Maintaining cycle limits assumed in the analyses provides assurance that the probability of fatigue cracking of ASME Class 1 components is minimized.

As part of the program, the effects of the reactor coolant environment on component fatigue life have been addressed by assessing the impact of the environment on a sample of critical components as identified in NUREG/CR-6260 (Reference 2) for an older vintage Westinghouse plant. Management of the fatigue effects is required for the hot leg surge line nozzle and the charging nozzle locations when environmental life correction factors are applied. The Metal Fatigue of Reactor Coolant Pressure Boundary program provides fatigue monitoring for these locations to ensure adequate margin against fatigue cracking due to anticipated cyclic strains and the effects of the reactor coolant environment.

In addition, the program monitors thermal cycles associated with selected auxiliary heat exchangers in order to ensure that original equipment specification cycle limits are not exceeded.

The program utilizes fatigue monitoring software (EPRI FatiguePro') to monitor plant transient cycles (in addition to using plant surveillance procedures) and to monitor fatigue usage for selected ASME Class 1 components. The software counts cycles and calculates fatigue usage for selected high usage components. The fatigue monitoring software counts most of the transient cycles that are required to be tracked by monitoring changes in plant instrument readings. Cycles that cannot be counted based on installed instrumentation are counted manually and then incorporated into the fatigue monitoring software database.

The Metal Fatigue of Reactor Coolant Pressure Boundary program provides for corrective actions in response to approaching an Action Limit on cycle counts or fatigue usage. When monitored transients or fatigue usage exceeds 80 percent of the design limit, the condition is evaluated and appropriate corrective action is initiated to ensure the design limit is not exceeded.

Limits are established based on equipment specifications or fatigue evaluation assumptions for cycle counts, ASME Code CUF limit of 1.0, or the CUF limit considering environmental effects, whichever is limiting.

Commitments

  • Routine Update of Cycle Count and Fatigue Usage Status The Metal Fatigue of Reactor Coolant Pressure Boundary program will be enhanced to include a routine assessment of the transient cycle count totals and fatigue usage status for monitored locations, including an action limit for the initiation of corrective action.

Revision 2511/26/14 KPS USAR 15-42 This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 28.

  • Additional Fatigue Evaluations (Surge Line HL Nozzle and Charging Line Nozzle)

A fatigue analysis of the surge line hot leg nozzle and the charging line nozzle in accordance with ASME B&PV Code Section III, Subsection NB-3200 guidance was performed to determine the CUF, considering the effects of the reactor coolant environment. The analysis confirmed that CUF is less than 1.0 at the end of 60 years of plant operation for both cases.

This commitment to perform these evaluations is identified in Table 15.7-1 License Renewal Commitments, Item 41.

  • Additional Fatigue Evaluation (Pressurizer Lower Head)

The Metal Fatigue of Reactor Coolant Pressure Boundary program will perform a fatigue evaluation of the pressurizer lower head and surge line that is consistent with the requirements of ASME B&PV Code,Section III, NB-3200 and will determine the cumulative fatigue usage through the period of extended operation.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 51.

  • Review of Class 1 Component Fatigue Evaluations The Metal Fatigue of Reactor Coolant Pressure Boundary program will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for 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.

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 52.

15.5 References

1. NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.
2. NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.

Revision 2511/26/14 KPS USAR 15-43 15.6 EXEMPTIONS The requirements of 10 CFR 54.21(c) stipulate that the application for a renewed license should include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and that are based on time-limited aging analyses, as defined in 10 CFR 54.3.

An exemption from the requirements of 10 CFR 50.61 and 10 CFR 50 Appendices G and H was granted in a May, 2001 letter from NRC (Reference 1). The exemption remains in effect and is based on a time-limited aging analysis. Specifically, the NRC issued an exemption to: (1) establish the use of a new methodology to meet the requirements of Appendix G to 10 CFR 50; (2) establish the use of a new methodology to meet the requirements of 10 CFR 50.61; and (3) modify the basis for the Kewaunee reactor pressure vessel surveillance program (required by Appendix H to 10 CFR 50) to incorporate the acquisition of fracture toughness data. The new methodology for assessing the RPV circumferential beltline weld is based on the use of the 1997 Edition of ASTM Standard Test Method E-1921 and ASME Code Case N-629. The exemption was necessary for the reactor vessel beltline weld to meet the pressurized thermal shock criterion of 10 CFR 50.61.

15.6 References6

1. Letter from NRC to M. Reddemann, NMC, Kewaunee Nuclear Power Plant - Exemption from the Requirements of 10 CFR Part 50, Appendix G, Appendix H, and Section 50.61 (TAC No. MA8585), May 1, 2001.

15.7 LICENSE RENEWAL COMMITMENTS Table 15.7-1, provides a listing of the license renewal commitment.

Revision 2511/26/14 KPS USAR 15-44 Table 15.7-1 LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea 1

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program will be enhanced 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 to augment the current inspections.

ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD At least 2 years prior to entering the period of extended operation.

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

ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD At least 2 years prior to entering the period of extended operation.

3 The Bolting Integrity program will be enhanced to further incorporate applicable EPRI and industry bolting guidance. Topic enhancements will include proper joint assembly, torque values, gasket types, use of lubricants, and other bolting fundamentals.

Bolting Integrity Prior to the period of extended operation.

Revision 2511/26/14 KPS USAR 15-45 4

The Buried Piping and Tanks Inspection program will be enhanced to perform visual inspections of a representative sample of material/protective measure combinations for in-scope buried piping and tanks.

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

  • Steel (including cast iron)/coated,
  • Steel/coated and wrapped,
  • Steel/uncoated, and
  • Stainless steel/coated and wrapped Visual inspections of the external surface of the components will be performed to identify damaged wrapping (if present),

degraded or damaged coating (if present),

and evidence of loss of material. Each piping inspection will include a minimum of ten linear feet of piping.

Buried Piping and Tanks Inspection Letter 10-548; Response to RAI B2.1.7-3a (Reference 1)

Prior to the period of extended operation, and During the first ten (10) years of the period of extended operation and During the second ten (10) years of the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-46 4 (cont) 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 ten years and the second ten years of the period of extended operation.

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 ten years and the second ten 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 ten years and the second ten 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 ten years of the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-47 4 (cont) 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 ten 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 ten years of the period of extended operation.

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

Compressed Air Monitoring Prior to the period of extended operation.

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

External Surfaces Monitoring Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-48 7

The External Surfaces Monitoring program will be enhanced to provide training for Operations, Engineering, and Health Physics personnel performing the program inspections and walkdowns. The training will address the requirements of the External Surfaces Monitoring program for license renewal, the need to document the identified conditions with sufficient detail to support monitoring and trending the aging effects, and the aging effects monitored by the program and how to identify them.

External Surfaces Monitoring Prior to the period of extended operation.

8 The Fire Protection program will be enhanced to test a representative sample of sprinkler heads or to replace all affected sprinkler heads in accordance with the requirements of NFPA 25.

Fire Protection Prior to the sprinkler heads achieving 50 years of service life.

9 The Fire Protection program fire barrier penetration seal inspections will be revised to include the elastomer Shield Building fire boots.

Fire Protection Prior to the period of extended operation.

10 The Fire Protection program inspections of the reactor coolant pump oil collection system will be revised to include additional inspection criteria for the visual inspection of the system and to perform a one-time inspection of the internal surfaces of the reactor coolant pump oil collection tank.

Fire Protection Prior to the period of extended operation.

11 The Fuel Oil Tank Inspections program will be enhanced to provide guidance for the periodic draining, cleaning and inspection activities.

Fuel Oil Tank Inspections Prior to the period of extended operation.

12 The Inspection of Overhead Heavy Load and Refueling Handling Systems program will be enhanced to clarify the requirements of visual inspection of structural members, including structural bolting, of the in-scope heavy load and refueling handling cranes and associated equipment.

Inspection of Overhead Heavy Load and Refueling Handling Systems Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-49 13 The Metal Enclosed Bus program will be enhanced to include augmented periodical visual inspections of the MEB internal surfaces, bus supports, bus insulation, taped joints and boots for signs of degradation or aging.

Metal Enclosed Bus Letter 09-469; Response to RAI B2.1.18-1 (Reference 2)

Prior to the period of extended operation.

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 periodically visually inspect for accessible electrical cables and connections installed in an adverse localized equipment environment. Should an adverse localized environment be observed, a representative sample of electrical cables and connections installed within that environment will be visually inspected for jacket surface anomalies.

Non-EQ Electrical Cables and Connections Prior to the period of extended operation.

Thereafter, the inspections will not exceed a 10-year interval.

15 The Non-EQ Electrical Cable Connections program will be established. The program will perform a one-time inspection, on a sampling basis, to confirm the absence of loosening of bolted connections.

Non-EQ Electrical Cable Connections Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-50 16 The Non-EQ Inaccessible Medium-Voltage Cables program will be established. The program will periodically inspect the in-scope manhole/pulling pit for water collection and will remove water, if required. The program will periodically perform a test on the in-scope cables to provide an indication of the condition of the conductor insulation.

Non-EQ Inaccessible Medium-Voltage Cables Letter 10-447; Second Annual Update (Reference 3)

Letter 10-548 Response to RAI B2.1.7-3a (Reference 1)

Prior to the period of extended operation.

Thereafter, the manhole/pulling pit inspections will not exceed a 2-year interval.

Thereafter, the cable testing will not exceed a 10-year interval.

17 The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals program will be established. The program will periodically perform a proven cable system test for detecting deterioration of the insulation system for those electrical cables and connections disconnected during calibration, or will periodically review the results and findings of calibrations for those electrical cables that remain connected during the calibration process.

Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals Prior to the period of extended operation.

Thereafter, the cable testing and calibration reviews will not exceed a 10-year interval.

18 The Open-Cycle Cooling Water System program will be enhanced to add the applicable aging effects as inspection criteria for the Circulating Water System underwater visual inspections.

Open-Cycle Cooling Water System Prior to the period of extended operation.

19 The Reactor Vessel Surveillance program will be enhanced to include the applicable limitations on operating conditions to which the surveillance capsules were exposed (e.g.

neutron flux, spectrum, irradiation temperature, etc.).

Reactor Vessel Surveillance Prior to the period of extended operation.

20 The Reactor Vessel Surveillance program will be enhanced to include requirements for storing, and possible recovery, of tested and untested capsules (removed from the Reactor Vessel after August 31, 2000).

Reactor Vessel Surveillance Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-51 21 The Selective Leaching of Materials program will be established. The program will perform a one-time visual inspection, and hardness measurement or qualitative examination of selected components within the scope of license renewal for selective leaching.

Selective Leaching of Materials Prior to the period of extended operation.

22 The Structures Monitoring Program will be enhanced to clearly define structures, structural elements, and miscellaneous structural commodities that are in scope.

Defined scope to include the MEB enclosure assemblies, structural supports, and enclosure seals.

Structures Monitoring Program Letter 09-469; Response to RAI B2.1.18-2 (Reference 2)

Prior to the period of extended operation.

23 The Structures Monitoring Program will be enhanced to monitor groundwater quality and verify that it remains non-aggressive to below-grade concrete.

Structures Monitoring Program Prior to the period of extended operation.

24 The Structures Monitoring Program will be enhanced to improve criteria for detection of aging effects for the underwater visual inspections of the in-scope structures.

Structures Monitoring Program Prior to the period of extended operation.

25 The Work Control Process program will be established. The program will perform one-time inspections as a verification of the effectiveness of chemistry control programs.

The 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.

Work Control Process Letter 09-597; Changes to the Work Control Process AMP (Reference 4)

Prior to the period of extended operation.

26 Deleted Letter 09-597; Changes to the Work Control Process AMP (Reference 4)

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-52 27 Deleted Letter 09-597; Changes to the Work Control Process AMP (Reference 4) 28 The Metal Fatigue of Reactor Coolant Pressure Boundary program will be enhanced to include a routine assessment of the transient cycle count totals and fatigue usage status for monitored locations, including an action limit for the initiation of corrective action.

Metal Fatigue of Reactor Coolant Pressure Boundary Prior to the period of extended operation.

29 The following will be further evaluated as part of Dominion's ongoing performance improvement programs:

  • Concurrent implementation of SAMAs 81,160,166 and 167.
  • Implementation of temporary screenhouse ventilation.

Environmental Report

- SAMA Analysis Letter 09-028 (Reference 5) and Letter 09-291 (Reference 6)

Prior to the period of extended operation.

30 Quarterly laboratory testing of fuel oil samples for water, sediment and particulates will be performed on the Emergency Diesel Generators and Technical Support Center Diesel Generator day tanks. The testing acceptance criteria will be consistent with the requirements specified in ASTM D975-06b for water and sediment and ASTM D6217 for particulates.

Letter 09-680; Response to RAI B2.1.14-3 (Reference 7)

Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-53 31 The Work Control Process Program will provide for a one-time-inspection of the Emergency Diesel Generators (EDG) Day Tanks and the Technical Support Center Diesel Generator (TSC DG) Day Tank. An exterior surfaces UT 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.

Letter 09-469 Response to RAI B2.1.15-1 (Reference 2)

Prior to the period of extended operation.

32 The 14 potentially cost beneficial SAMAs identified in LRA Appendix E, Attachment F, will be further evaluated as part of Dominion's ongoing performance improvement programs Environmental Report

- SAMA Analysis Letter 08-0462 (Reference 8)

Prior to the period of extended operation.

33 Develop a plan for identification and remediation of reactor refueling cavity liner leakage to be implemented during the period of extended operation.

Letter 09-760; Response to RAI B2.1.31-4a (Reference 9)

Prior to the period of extended operation.

34 At least one core bore sample will be taken from the waste drumming room reinforced concrete ceiling below the spent fuel pool.

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.

Letter 09-760; Response to RAI B2.1.31-5a (Reference 9)

Letter 10-093; Supplemental Response to RAI B2.1.31-5a (Reference 15)

Prior to the end of 2011 35 Develop an action plan for identification and remediation of spent fuel pool liner leakage to be implemented during the period of extended operation.

Letter 09-760; Response to RAI B2.1.31-5a (Reference 9)

Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-54 36 If SFP liner leakage persists during the period of extended operation, an additional concrete core sample will be taken from the waste drumming room reinforced concrete ceiling below the spent fuel pool. 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.

Letter 09-760; Response to RAI B2.1.31-5a (Reference 9)

Prior to the end of the first ten years of extended operation 37 Perform a VT-1 visual examination of the stainless steel cladding of a safety injection pump for indications of cracking or corrosion due to cladding breach.

Letter 09-777; Response to RAI 3.2.2.2.2 (Reference 10)

Prior to the period of extended operation.

38 The boron carbide surveillance program, which includes neutron attenuation testing, will continue to be performed during the period of extended operation every 3 years.

Letter 09-777; Supplemental Response to RAI 3.3.2.2.6-1 (Reference 10)

During the period of extended operation.

39 A surveillance program will be implemented to perform verification that the Boral spent fuel storage rack neutron absorber B-10 areal density is maintained within the bounds of the spent fuel pool criticality analysis. Alternatively, the criticality analysis for the spent fuel pool will be revised to eliminate credit for the Boral neutron absorber material.

Letter 09-777; Supplemental Response to RAI 3.3.2.2.6-2 (Reference 10)

Prior to 2017.

Surveillance program will be performed every 10 years thereafter.

40 Implement nitrate monitoring for the Component Cooling System on a frequency consistent with the existing monitoring for ammonia.

Letter 10-008; Response to RAI B2.1.8-3a (Reference 11)

Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-55 41 Perform a fatigue analysis of the surge line hot leg nozzle and the charging line nozzle in accordance with ASME B&PV Code Section III, Subsection NB-3200 guidance and determine the CUF, considering the effects of the reactor coolant environment.

Confirm that CUF is less than 1.0 at the end of 60 years of plant operation.

Letter 10-324; Completion of Commitment 41 related to RAI B3.2-2 (Reference 12)

Prior to the period of extended operation.

(Complete) 42 For Examination Category B-J, Item No.

B9.21, eight ASME Class 1 small-bore circumferential welds will receive volumetric and surface examinations during each 10-year lSI inspection interval during the period of extended operation.

Letter 10-033; Supplemental Response to RAI B2.1.2-1 (Reference 13)

During each 10-year lSI inspection interval during the period of extended operation.

43 Ten volumetric examinations of ASME Class 1 small-bore socket welds will be performed using a demonstrated, nuclear-industry endorsed, inspection methodology that can detect cracking within the specified examination volume, if a methodology becomes available. In the event that a demonstrated, nuclear-industry endorsed, inspection methodology is not available, destructive examinations of socket welds will be substituted for volumetric non-destructive examinations. Each destructive weld examination will be considered equivalent to performing two volumetric weld examinations, such that a maximum of five destructive examinations will be performed.

Letter 10-665; Supplemental Response to RAI B2.1.2-2 (Reference 14)

Four volumetric examinations or two destructive examinations (or an equivalent combination of examinations) prior to the period of extended operation.

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

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-56 44 Core samples will be obtained from the inside surface of a concrete wall (below the groundwater table elevation) or from the foundation basemat in the vicinity of the groundwater 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.

Letter 10-093; Supplemental Response to RAI B2.1.31-3a (Reference 15)

Prior to the period of extended operation 45 In the event that the chloride content in the groundwater does not decrease to below 500 ppm within the first ten years of the period of extended operation, core samples will be obtained from the inside surface of a concrete wall (below the groundwater table elevation) or from the 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.

Letter 10-093; Supplemental Response to RAI B2.1.31-3a (Reference 15)

Prior to the end of the first ten years of extended operation 46 If the results of the core sample testing of the waste drumming room reinforced concrete ceiling leakage site (related to potential SFP liner leakage - Commitment 34) indicate degradation of 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.

Letter 10-093; Supplemental Response to RAI B2.1.31-4a (Reference 15)

Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-57 47 Submit three examples of operating experience associated with the Work Control Process - Internal Surfaces Monitoring program for NRC staff review in determining the effectiveness of the program to detect and correct the effects of aging prior to the loss of intended function.

Letter 10-286; Response to RAI B2.1.32-5.

(Reference 16)

Within 2 years following implementation of the Work Control Process aging management program.

48 The cathodic protection system associated with the diesel generator fuel oil storage tanks and protected portions of the fuel oil lines, and the circulating water system recirculation piping, will 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.

Letter 10-548 Response to RAI B2.1.7-3a (Reference 1)

During the period of extended operation.

49 Recognizing that the EPRI SGMP resolution is still under development, Kewaunee will perform an inspection of each steam generator to assess 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 will continue to actively participate in the EPRI SGMP studies.

Letter 10-548 Response to RAI 3.1.2.2.13-1a (Reference 1)

Prior to 2023 Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-58 50 Perform an audit of the Internal Surfaces Monitoring portion of the Work Control Process program inspections to confirm that the components representing the leading indicators of aging for each of the material/environment combinations have been inspected at least once during the audit period.

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

Letter 10-595 (Supplemental Response to RAI B2.1.32-5a)

(Reference 17)

Prior to the period of extended operation and every 10 years thereafter.

Deliberate focused inspections will be performed within 5 years of the completion of the audits.

51 Dominion Energy Kewaunee, Inc. (DEK) will perform a fatigue evaluation of the pressurizer lower head and surge line that is consistent with the requirements of ASME B&PV Code,Section III, NB-3200 and will determine the cumulative fatigue usage through the period of extended operation.

Letter 10-595 Supplemental Response to RAI B3.2-2a (Reference 17)

Prior to the period of extended operation.

52 DEK will perform a review of design basis ASME Class 1 component fatigue evaluations to determine whether the NUREG/CR-6260-based components that have been evaluated for the effects of the reactor coolant environment on fatigue usage are the limiting components for 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.

Letter 10-595 Supplemental Response to RAI B3.2-2a (Reference 17)

Prior to the period of extended operation.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-59 53 DEK will develop a plan to address the potential for failure of the primary-to-secondary pressure boundary due to PWSCC cracking of tube-to-tubesheet welds. The plan will consist of two resolution options:

1. Perform an analytical evaluation of the steam generator tube-to-tubesheet welds in order to:

a.

Establish a technical basis which concludes that the structural integrity of the steam 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 identify PWSCC cracking.

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.

Letter 10-595 (Reference 17)

Develop a plan prior to the period of extended operation.

Implement the requirements of the plan prior to 2023.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-60 15.7 References

1. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-548, September 23, 2010.
2. Letter from Steven E. Scace (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application - Aging Management Programs, Letter No.09-469, August 17, 2009.
3. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), License Renewal Application Second Annual Update Required by 10 CFR 54.21(b), Letter No.10-447, August 9, 2010.
4. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Supplemental Information for the Review of the Kewaunee Power Station License Renewal Application -

Changes to the Work Control Process Aging Management Program, Letter No.09-597, September 25, 2009.

5. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Request for Additional Information Regarding Severe Accident Mitigation Alternatives for Kewaunee Power Station License Renewal Application, Letter No.09-028, March 9, 2009.
6. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Follow-up Questions Regarding The Severe Accident Mitigation Alternatives for Kewaunee Power Station, Letter No.09-291, June 1, 2009.

54 The Structures Monitoring Program will be revised to include the evaluation criteria of ACI 349.3R-96, Chapter 5, as the criteria to be used when evaluating conditions or findings 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.

Letter 10-707; Response to RAI B2.1.31-9 (Reference 16)

Prior to the period of extended operation.

a. The period of extended operation is the period of 20 years beyond the expiration date of the units original operating license.

Table 15.7-1 (continued)

LICENSE RENEWAL COMMITMENTS Item Commitment Source Schedulea

Revision 2511/26/14 KPS USAR 15-61

7. Letter from William R. Mathews (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application - Aging Management Review Results, Letter No.09-680, November 13, 2009.
8. Application for Renewed License, Letter No. 08-0462 dated August 12, 2008, Appendix E, Attachment F.
9. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.09-760, December 28, 2009.
10. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application - Aging Management Review/Aging Management Program, Letter No.09-777, January 21, 2010.
11. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application - Aging Management Review/Aging Management Program, Letter No.10-008, January 10, 2010.
12. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Completion of Kewaunee Power Station License Renewal Commitment 41, Letter No.10-324, June 1, 2010.
13. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application - Aging Management Review/Aging Management Program, Letter No.10-033, February 2, 2010.
14. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Supplemental Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-665, November 9, 2010.
15. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Supplemental Information for the Review of the Kewaunee Power Station License Renewal Application -

Aging Management Review/Aging Management Program, Letter No.10-093, February 15, 2010.

16. Letter from Leslie N. Hartz (DEK) to Document Control Desk (NRC), Response to Request for Additional Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-286, May 13, 2010.
17. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Supplemental Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-595, October 20, 2010.

Revision 2511/26/14 KPS USAR 15-62

18. Letter from J. Alan Price (DEK) to Document Control Desk (NRC), Supplemental Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-707, November 23, 2010.
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