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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Text

Table of Contents tion Title Page INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1

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

.2 Operating Experience . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-3 AGING MANAGEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-3

.1 Aging Management Programs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-3

.2 Time Limited Aging Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-4

.3 Time limited aging analysis Support Programs . . . . . . . . . . . . . . . . . . . . . . . . . 15-5 PROGRAMS THAT MANAGE THE EFFECTS OF AGING . . . . . . . . . . . . . . . 15-5

.1 Alloy 600 Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-5

.2 ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD . . . 15-6

.3 ASME Section XI, Subsection IWE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-7

.4 ASME Section XI, Subsection IWF . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-8

.5 Bolting Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-8

.6 Boric Acid Corrosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-9

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

.8 Closed-Cycle Cooling Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-11

.9 Compressed Air Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-12

.10 External Surfaces Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-12

.11 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-13

.12 Flow-Accelerated Corrosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-14

.13 Flux Thimble Tube Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-15

.14 Fuel Oil Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-15

.15 Fuel Oil Tank Inspections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-16

.16 Inspection of Overhead Heavy Load and Refueling Handling Systems . . . . . . . 15-17

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

.18 Metal Enclosed Bus. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-17

.19 Non-EQ Electrical Cables and Connections . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-18

.20 Non-EQ Electrical Cable Connections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-19

.21 Non-EQ Inaccessible Medium-Voltage Cables . . . . . . . . . . . . . . . . . . . . . . . . . 15-20

.22 Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-21

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

.24 Primary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-22

Table of Contents (continued) tion Title Page

.25 Reactor Containment Leakage Testing 10 CFR 50, Appendix J . . . . . . . . . . . . 15-22

.26 Reactor Head Closure Studs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-23 3.27 Reactor Vessel Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-23

.28 Secondary Water Chemistry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-24

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

.30 Steam Generator Tube Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-25

.31 Structures Monitoring Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-25

.32 Work Control Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-28 TIME-LIMITED AGING ANALYSES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-31

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

.2 Metal Fatigue. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-33

.3 Environmental Qualification of Electric Equipment. . . . . . . . . . . . . . . . . . . . . . 15-36

.4 Containment Fatigue Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-37

.5 Other Plant-Specific Time-Limited Aging Analyses . . . . . . . . . . . . . . . . . . . . . 15-37 TLAA SUPPORT PROGRAMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-40

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

.2 Metal Fatigue of Reactor Coolant Pressure Boundary . . . . . . . . . . . . . . . . . . . . 15-40 EXEMPTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-43 LICENSE RENEWAL COMMITMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-43

List of Tables le Title Page le 15.7-1 License Renewal Commitments. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-44

Intentionally Blank MANAGE THE EFFECTS OF AGING INTRODUCTION The application for a Renewed Operating License is required by 10 CFR 54.21(d) to ude a USAR Supplement. This appendix comprises the USAR supplement and includes the owing 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 nagement programs necessary to provide reasonable assurance that components within the pe of license renewal will continue to perform their intended functions consistent with the rent Licensing Basis (CLB) for the period of extended operation. The period of extended ration extends 20 years from the units original operating license expiration date through ember 21, 2033.

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

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.

Plant-specific and industry operating experience, including past corrective actions resulting rocess enhancements, was considered in development of the aging management programs.

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

15.1 References 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.

NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.

AGING MANAGEMENT

.1 Aging Management Programs The aging management programs for Kewaunee Power Station are described in the sections d below. The list identifies the implementation status of the programs when the ROL was ed.

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

Alloy 600 Inspections [Section 15.3.1] [Existing]

ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD [Section 15.3.2]

[Existing - Requires Enhancement]

ASME Section XI, Subsection IWE [Section 15.3.3] [Existing]

ASME Section XI, Subsection IWF [Section 15.3.4] [Existing]

Bolting Integrity [Section 15.3.5] [Existing - Requires Enhancement]

Boric Acid Corrosion [Section 15.3.6] [Existing]

Buried Piping and Tanks Inspection [Section 15.3.7][Existing - Requires Enhancement]

Closed-Cycle Cooling Water System [Section 15.3.8] [Existing - Requires Enhancement]

External Surfaces Monitoring [Section 15.3.10] [Existing - Requires Enhancement]

Fire Protection [Section 15.3.11] [Existing - Requires Enhancement]

Flow-Accelerated Corrosion [Section 15.3.12] [Existing]

Flux Thimble Tube Inspection [Section 15.3.13] [Existing]

Fuel Oil Chemistry [Section 15.3.14] [Existing - Requires Enhancement]

Fuel Oil Tank Inspections [Section 15.3.15] [Existing - Requires Enhancement]

Inspection of Overhead Heavy Load and Refueling Handling Systems [Section 15.3.16]

[Existing - Requires Enhancement]

Lubricating Oil Analysis [Section 15.3.17] [Existing]

Metal Enclosed Bus [Section 15.3.18] [Existing - Requires Enhancement]

Non-EQ Electrical Cables and Connections [Section 15.3.19] [To Be Developed]

Non-EQ Electrical Cable Connections [Section 15.3.20] [To Be Developed]

Non-EQ Inaccessible Medium-Voltage Cables [Section 15.3.21] [To Be Developed]

Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level Signals

[Section 15.3.22] [To Be Developed]

Open-Cycle Cooling Water System [Section 15.3.23] [Existing - Requires Enhancement]

Primary Water Chemistry [Section 15.3.24] [Existing]

Reactor Containment Leakage Testing 10 CFR 50, Appendix J [Section 15.3.25] [Existing]

Reactor Head Closure Studs [Section 15.3.26] [Existing]

Reactor Vessel Surveillance [Section 15.3.27] [Existing - Requires Enhancement]

Secondary Water Chemistry [Section 15.3.28] [Existing]

Selective Leaching of Materials [Section 15.3.29] [To Be Developed]

Steam Generator Tube Integrity [Section 15.3.30] [Existing]

Structures Monitoring Program [Section 15.3.31] [Existing - Requires Enhancement]

Work Control Process [Section 15.3.32] [To Be Developed]

.2 Time Limited Aging Analyses Reactor Vessel Neutron Embrittlement [Section 15.4.1]

Metal Fatigue [Section 15.4.2]

Containment Fatigue Analysis [Section 15.4.4]

Other Plant-Specific Time-Limited Aging Analyses [Section 15.4.5]

.3 Time limited aging analysis Support Programs Environmental Qualification (EQ) of Electric Components [Section 15.5.1] [Existing]

Metal Fatigue of Reactor Coolant Pressure Boundary [Section 15.5.2] [Existing - Requires Enhancements]

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

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

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

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

gram Description The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program is existing program that corresponds to NUREG-1801,Section XI.M1, ASME Section XI rvice Inspection, Subsections IWB, IWC, and IWD.

The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program ages the aging effects of change in dimensions, cracking, loss of fracture toughness, loss of erial, and loss of preload for the ASME Class 1, 2, and 3 piping, including piping less than r inches nominal pipe size, and components fabricated of nickel alloys, stainless steel, and

l. In addition, the program manages the aging effect of cracking for the steel reactor coolant p motor flywheels.

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

mmitments

  • Aging Management of Reactor Vessel Internals The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program be enhanced to (1) participate in the industry programs for investigating and managing aging cts on reactor internals; (2) evaluate and implement the results of the industry programs as licable to the reactor internals; and (3) upon completion of these programs, but not less than 24 nths before entering the period of extended operation, submit an inspection plan for reactor rnals 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 be enhanced to include identification of the limiting susceptible cast austenitic stainless steel tor vessel internals components from the standpoint of thermal aging susceptibility, neutron nce, and cracking. For each identified component, a plan will be developed, which omplishes aging management through either a supplemental examination or a ponent-specific evaluation. The plan will be submitted for NRC review and approval not less 24 months before entering the period of extended operation.
  • Aging Management of Small Bore Circumferential Welds The ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program be enhanced to ensure that for Examination Category B-J, Item No. B9.21, eight ASME ss 1 small-bore circumferential welds will receive volumetric and surface examinations during h 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 be enhanced to ensure ten volumetric examinations of ASME Class 1 small-bore socket ds will be performed using a demonstrated, nuclear-industry endorsed, inspection hodology that can detect cracking within the specified examination volume, if a methodology omes available. In the event that a demonstrated, nuclear-industry endorsed, inspection hodology is not available, destructive examinations of socket welds will be substituted for umetric non-destructive examinations. Each destructive weld examination will be considered ivalent to performing two volumetric weld examinations, such that a maximum of five ructive examinations will be performed.

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

.3 ASME Section XI, Subsection IWE gram Description The ASME Section XI, Subsection IWE program is an existing program that corresponds to REG-1801,Section XI.S1, ASME Section XI, Subsection IWE.

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

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

ntaining the integrity of the Reactor Containment Vessel pressure boundary and structural grity and ensuring that aging effects are discovered and repaired before the loss of structure or ponent intended functions.

.4 ASME Section XI, Subsection IWF gram Description The ASME Section XI, Subsection IWF program is an existing program that corresponds to REG-1801,Section XI.S3, ASME Section XI, Subsection IWF.

The ASME Section XI, Subsection IWF program manages the aging effects of loss of erial 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, ss 2, and Class 3 component supports. The program support and hanger inspections fulfill the uirements specified by 10 CFR 50.55a(g). Removal, repair, monitoring, or analytical luation are identified as acceptable corrective action options.

.5 Bolting Integrity gram Description The Bolting Integrity program is an existing program that corresponds to NUREG-1801, tion XI.M18, Bolting Integrity.

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

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

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

mmitments

  • Bolting Program Improvements

ustry bolting guidance. Topic enhancements will include proper joint assembly, torque values, ket types, use of lubricants, and other bolting fundamentals.

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

.6 Boric Acid Corrosion gram Description The Boric Acid Corrosion program is an existing program that corresponds to REG-1801,Section XI.M10, Boric Acid Corrosion.

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

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

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

.7 Buried Piping and Tanks Inspection gram Description The Buried Piping and Tanks Inspection program is an existing program that corresponds to REG-1801,Section XI.M34, Buried Piping and Tanks Inspection.

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

sures:

  • 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 orms opportunistic and deliberate visual inspections of the external surface of a representative ple of the material/protective measures combinations of the in-scope buried piping and ponents. The program inspections inspect for evidence of damaged wrapping; coating cts, such as coating perforation, holidays, or other damage; and evidence of loss of material he external surface of the piping or component.

mmitments

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

Visual inspections of the external surface of the components will be performed to identify aged wrapping (if present), degraded or damaged coating (if present), and evidence of loss of erial. 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.

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.

.8 Closed-Cycle Cooling Water System gram Description The Closed-Cycle Cooling Water System program is an existing program that corresponds UREG-1801,Section XI.M21, Closed-Cycle Cooling Water System.

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

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

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

.9 Compressed Air Monitoring gram Description The Compressed Air Monitoring program is an existing program that corresponds to REG-1801,Section XI.M24, Compressed Air Monitoring.

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

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

mmitments

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

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

.10 External Surfaces Monitoring gram Description The External Surfaces Monitoring program is an existing program that corresponds to REG-1801,Section XI.M36, External Surfaces Monitoring.

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

orm the external surface visual inspections. Nuclear Auxiliary Operators perform rounds each t in accessible plant areas and perform general inspections, which include specific inspection ils related to monitoring equipment aging. System Engineers perform comprehensive visual ections during walkdowns of plant systems and components during both normal operation refueling outages. The guidance for System Engineer walkdowns provides a walkdown cklist of attributes to be observed, which includes inspection criteria related to aging nagement. Health Physics technicians routinely perform radiological surveys in the ologically controlled areas of the plant and look for any evidence of boron precipitation and ve radioactive system leaks observed while performing these surveys.

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

mmitments

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

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

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

3.11 Fire Protection gram Description The Fire Protection program is an existing program that corresponds to NUREG-1801, tions XI.M26, Fire Protection and XI.M27, Fire Water System.

king, delamination, increased hardness, loss of material, loss of sealing, loss of strength, nkage, and spalling for the fire protection components and features.

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

mmitments

  • Inspect or Replace Fire Sprinklers The Fire Protection program will be enhanced to test a representative sample of sprinkler ds or to replace all affected sprinkler heads in accordance with the requirements of NFPA 25 ference 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 be revised to include additional inspection criteria for the visual inspection of the system and erform a one-time inspection of the internal surfaces of the reactor coolant pump oil collection This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 10.

.12 Flow-Accelerated Corrosion gram Description The Flow-Accelerated Corrosion program is an existing program that corresponds to REG-1801,Section XI.M17, Flow-Accelerated Corrosion.

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

lies to both safety-related and non-safety-related components.

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

.13 Flux Thimble Tube Inspection gram Description The Flux Thimble Tube Inspection program is an existing program that corresponds to REG-1801,Section XI.M37, Flux Thimble Tube Inspection.

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

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

.14 Fuel Oil Chemistry gram Description The Fuel Oil Chemistry program is an existing program that corresponds to NUREG-1801, tion XI.M30, Fuel Oil Chemistry.

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

ater and microbiological organisms, and verifies the quality of new oil before its introduction the diesel generator fuel oil storage tanks. The program defines specific acceptance criteria contaminant concentrations, which reflect ASTM guidelines for parameters that maintain taminant concentrations below unacceptable levels. Should unacceptable indications be erved, the condition is documented and evaluated using the Corrective Action Program.

mmitments

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

This commitment is identified in Table 15.7-1 License Renewal Commitments, Item 30

.15 Fuel Oil Tank Inspections gram Description The Fuel Oil Tank Inspections program is an existing program that corresponds to REG-1801,Section XI.M30, Fuel Oil Chemistry.

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

mmitments

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

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

gram Description The Inspection of Overhead Heavy Load and Refueling Handling Systems program is an ting program that corresponds to NUREG-1801,Section XI.M23, Inspection of Overhead vy Load and Light Load (Related to Refueling) Handling Systems.

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

mmitments

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

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

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

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

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

.18 Metal Enclosed Bus gram Description The Metal Enclosed Bus program is an existing program that corresponds to NUREG-1800, tion XI.E4, Metal Enclosed Bus.

stance, electrical failure and loosening of bolted connections for non-segregated metal losed bus (MEB) and internal components within the scope of license renewal.

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

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

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

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

mmitments

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

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

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

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

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

tamination.

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

mmitments

  • 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.

.20 Non-EQ Electrical Cable Connections gram Description The Non-EQ Electrical Cable Connections program is a new program that will correspond UREG-1801,Section XI. E6, Electrical Cable Connections Not Subject To 10 CFR 50.49 ironmental Qualification Requirements (Revised).

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

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

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

mmitments

  • 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.

gram Description The Non-EQ Inaccessible Medium-Voltage Cables program is a new program that will espond to NUREG-1801,Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject to CFR 50.49 Environmental Qualification Requirements.

The Non-EQ Inaccessible Medium-Voltage Cables program will manage the aging effects ocalized damage and breakdown of insulation leading to electrical failure for non-EQ, cessible, low and medium-voltage (> 480 volts) cables within the scope of license renewal 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 s, e.g., cable in standing water. Periodic exposures to moisture that last less than a few days, normal rain and drain, are not significant. An adverse localized environment is a condition in mited plant area that is significantly more severe than the specified service environment for the les (power, control, and instrumentation) and connections. An adverse localized environment ignificant if it could appreciably increase the rate of aging of a component, or has an ediate adverse effect on operability.

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

program will perform a test on the in-scope non-EQ inaccessible low and medium-voltage les 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, the EDG fuel oil storage tank access manholes for water collection will be performed based ctual plant experience with water accumulation in the manhole. However, the inspection will erformed at least every two years. The first inspection for license renewal will be performed r to the period of extended operation.

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

mmitments

  • 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.

gram Description The Non-EQ Instrumentation Circuits Subject to Sensitive, High-Voltage, Low-Level nals program is a new program that will correspond to NUREG-1801,Section XI.E2, ctrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification uirements.

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

The program will perform a proven cable system test for detecting deterioration of the lation system (such as insulation resistance tests, time domain reflectometry tests, or other ing judged to be effective in determining cable insulation condition) for those electrical cables connections disconnected during calibration, or will review the results and findings of brations 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 ration and will be repeated every ten years thereafter.

mmitments

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

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

.23 Open-Cycle Cooling Water System gram Description The Open-Cycle Cooling Water System program is an existing program that corresponds to REG-1801,Section XI.M20, Open-Cycle Cooling Water System.

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

ections, nondestructive examinations, heat exchanger thermal performance testing, and ntenance, which includes flushing and cleaning, to manage aging of the open-cycle cooling er systems.

mmitments

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

.24 Primary Water Chemistry gram Description The Primary Water Chemistry program is an existing program that corresponds to REG-1801,Section XI.M2, Water Chemistry.

The Primary Water Chemistry program manages the aging effects of cracking, loss of erial, 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 ntain the primary system pressure boundary integrity.

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

.25 Reactor Containment Leakage Testing 10 CFR 50, Appendix J gram Description The Reactor Containment Leakage Testing 10 CFR 50, Appendix J program is an existing gram 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 ng effects of cracking, loss of leak tightness, loss of material, loss of sealing and leakage ough the Reactor Containment Vessel, including the systems penetrating the Reactor ntainment Vessel, penetrations, isolation valves, fittings and access openings made of tomers, stainless steel, and steel to detect degradation of the pressure boundary.

g Option B of Appendix J. The regulatory basis for the program includes NRC Regulatory de 1.163 (Reference 16), and NEI 94-01 (Reference 17).

.26 Reactor Head Closure Studs gram Description The Reactor Head Closure Studs program is an existing program that corresponds to REG-1801,Section XI.M3, Reactor Head Closure Studs.

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

.27 Reactor Vessel Surveillance gram Description The Reactor Vessel Surveillance program is an existing program that corresponds to REG-1801,Section XI.M31, Reactor Vessel Surveillance.

The Reactor Vessel Surveillance program manages the aging effects of loss of fracture ghness 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 udes surveillance capsule removal and specimen mechanical testing/evaluation, radiation lysis, development of pressure-temperature limits, and determination of low-temperature rpressure protection (LTOP) set points. The program ensures the reactor vessel materials meet fracture toughness requirements of 10 CFR 50, Appendix G, and meet the requirements of ssurized Thermal Shock (PTS) and upper shelf energy in 10 CFR 50.60 and 10 CFR 50.61, ectively, as modified by the exemption granted to utilize Master Curve methodology.

mmitments

  • Operating Restrictions The Reactor Vessel Surveillance program will be enhanced to include the applicable tations on operating conditions to which the surveillance capsules were exposed (e.g. neutron

, spectrum, irradiation temperature, etc.).

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

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

.28 Secondary Water Chemistry gram Description The Secondary Water Chemistry program is an existing program that corresponds to REG-1801,Section XI.M2, Water Chemistry.

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

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

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

3.29 Selective Leaching of Materials gram Description The Selective Leaching of Materials program is a new program that will correspond to REG-1801, XI.M33, Selective Leaching of Materials.

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

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

erformed prior to the period of extended operation.

mmitments

  • 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.

.30 Steam Generator Tube Integrity gram Description The Steam Generator Tube Integrity program is an existing program that corresponds to REG-1801,Section XI.M19, Steam Generator Tube Integrity.

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

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

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

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

ctural elements (including component supports), miscellaneous structural commodities, and onry walls. The program implements the requirements of 10 CFR 50.65, Requirements for nitoring the Effectiveness of Maintenance at Nuclear Power Plants, with the guidance of MARC 93-01, Revision 2 (Reference 20), and Regulatory Guide 1.160, Revision 2 ference 21). For masonry walls within the scope of license renewal, the Structures Monitoring gram manages aging effects based on guidance provided in IE Bulletin 80-11 (Reference 22),

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

mmitments

  • Define In-Scope Structural Elements The Structures Monitoring Program will be enhanced to clearly define structures, structural ments, 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 I 349.3R-96, Chapter 5 (Reference 25), as the criteria to be used when evaluating conditions indings identified during concrete structure inspections. This will be done prior to the ormance of the next scheduled inspection which will occur prior to the period of extended ration.

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 fy 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 g 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

mplemented 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 mplemented 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 crete ceiling below the spent fuel pool (SFP). The core sample location and depth will be icient to validate the strength of the concrete and the extent of any degradation. The core ple will be tested for compressive strength and will be subject to petrographic examination.

nforcing 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 ing leakage site (related to potential SFP liner leakage - Commitment 34) indicate degradation he structural integrity of the concrete, at least one core bore sample will be taken near at least of the refueling cavity liner leakage indication sites. The core sample location and depth will ufficient to validate the strength of the concrete and the extent of any degradation. The core ple will be tested for compressive strength and will be subject to petrographic examination.

nforcing 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 sample will be taken from the waste drumming room reinforced concrete ceiling below the nt fuel pool. The core sample location and depth will be sufficient to validate the strength of concrete and the extent of any degradation. The core sample will be tested for compressive ngth and will be subject to petrographic examination. Reinforcing steel in the core sample area 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 undwater table elevation) or from the foundation basemat in the vicinity of the groundwater ls for which average sampling results have exceeded the chloride concentration limit of 500

. The concrete core samples will be tested to determine if the chloride content within the crete could cause degradation due to corrosion of reinforcing steel.

In the event that the chloride content in the groundwater does not decrease to below 500 within the first ten years of the period of extended operation, core samples will be obtained m the inside surface of a concrete wall (below the groundwater table elevation) or from the ndation basemat in the vicinity of a groundwater well for which average sampling results have eeded the chloride concentration limit of 500 ppm. The concrete core samples will be tested to rmine if the chloride content within the concrete could cause degradation due to corrosion of forcing steel.

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

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

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

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

nless steel diesel exhaust flexible connections.

mmitments

  • 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 -

rnal Surfaces Monitoring program for NRC staff review in determining the effectiveness of 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 ergency Diesel Generators (EDG) Day Tanks and the Technical Support Center Diesel erator (TSC DG) Day Tank. An exterior surfaces ultra-sonic testing (UT) inspection will be formed to verify wall thickness of the bottom of each day tank. Based upon the UT ections, the most limiting EDG day tank will also be drained, cleaned and visually inspected 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 ram inspections to confirm that the components representing the leading indicators of aging each of the material/environment combinations have been inspected at least once during the t period.

If any scheduled surveillance and maintenance activities which were intended to encompass ponents as leading indicators of aging in each of the material/environment combinations have 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.

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.

NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.

NUREG-1339, Resolution of Generic Safety Issue 29: Bolting Degradation or Failure in Nuclear Power Plants, U.S. Nuclear Regulatory Commission, June 1990.

EPRI NP-5769, Degradation and Failure of Bolting in Nuclear Power Plants, May 5, 1988.

NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, March 17, 1988.

NRC Generic Letter 88-14, Instrument Air Supply System Problems Affecting Safety-Related Equipment, August 8, 1988.

ASME OM-S/G-1998, Part 17, Performance Testing of instrument Air Systems Information Notice for Light-Water Reactor Power Plants.

EPRI TR-108147, Compressor and Instrument Air Maintenance Guide, March 1998.

NFPA 25, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems, 1998 Edition, National Fire Protection Association.

EPRI 1011838, Recommendations for an Effective Flow-Accelerated Corrosion Program (NSAC-202L-R3), May 2006.

NRC Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors, July 26, 1988.

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.

ASTM D975-06b, Standard Specification for Diesel Oil Fuels, Revision B, November 1, 2006 ASTM D6217, Standard Test Method for Particulate Contamination in Middle Distillate Fuels by Laboratory Filtration, August 2010.

NRC Regulatory Guide 1.137, Rev. 1, Fuel-Oil Systems for Standby Diesel Generators, October 1979.

NRC Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, September 1995.

Appendix J.

NRC Regulatory Guide1.65, Material and Inspection for Reactor Vessel Closure Studs, October 1973.

NEI 97-06, Rev. 2, Steam Generator Program Guidelines, May 2005.

NUMARC 93-01, Rev 2, Industry Guideline For Monitoring The Effectiveness Of Maintenance At Nuclear Power Plants, April 1996.

NRC Regulatory Guide 1.160, Rev. 2, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

IE Bulletin 80-11, Masonry Wall Design, May 8, 1980.

NRC Information Notice 87-67, Lessons Learned from Regional Inspections of Licensee Actions in Response to IE Bulletin 80-11, December 13, 1987.

NRC Regulatory Guide 1.127, Rev. 1, Inspection of Water-Control Structures Associated with Nuclear Power Plants, March 1978.

American Concrete Institute ACI 349.3R-96, Evaluation of Existing Nuclear Safety-Related Concrete Structures, January 1, 1996.

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

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

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

oved from the reactor) and WCAP-16641 documents the results of the fluence evaluation for specimens. The fluence calculations concluded that Capsule T surveillance specimens ived a fluence of 5.62E+19 n/cm2 (E>1.0 MeV) after irradiation to 24.6 EFPY and the peak tor vessel clad/base metal interface fluence after 24.6 EFPY of plant operation was 2.60E+19 m2 (E>1.0 MeV). The Capsule T specimens have received a fluence equivalent to slightly ater than 52.1 EFPY. The maximum vessel exposures occur on the intermediate shell base erial with all other vessel materials experiencing a lower neutron exposure. Certain materials he extended beltline (inlet nozzles, inlet and outlet nozzle to upper shell welds, upper shell ing, and intermediate shell to upper shell girth weld) are projected to receive fluence greater n the 10 CFR 50, Appendix H threshold value of 1.0E+17 n/cm2 during the 40 - 60 year rating period.

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

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

4.1.2 Pressurized Thermal Shock Reactor pressure vessel beltline fluence is one of the factors used in determining the margin cceptability of the reactor pressure vessel to pressurized thermal shock as a result of radiation brittlement. The margin is the difference between the maximum nil ductility reference perature in the limiting beltline material and the screening criteria established in accordance h 10 CFR 50.61(b)(2). The screening criteria for the limiting reactor vessel materials are

°F for beltline plates, forging, and axial weld materials, and 300°F for beltline circumferential d materials.

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

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

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

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

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

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

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

.2 Metal Fatigue

.2.1 Fatigue of ASME Class 1 Components

.2.1.1 Component Design Transient Cycles Operating experience at the plant and other Westinghouse NSSS units has demonstrated that analyzed numbers of design basis transients are generally conservative for a 40 year life. The tal Fatigue of Reactor Coolant Pressure Boundary program monitors transients and ponents to assure that actual plant operation remains bounded by the assumptions used in the ign analyses. This program tracks cycles of design basis transient events and evaluates the ber of occurrences against the design basis.

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

rance that the design basis number of transients will not be exceeded during the period of nded operation. These transients will continue to be tracked in accordance with the Metal gue of Reactor Coolant Pressure Boundary program for the remaining plant life.

.2.1.2 ASME Class 1 Vessels and Surge Line Piping The reactor vessel (including the control rod drive mechanism pressure housings), steam erators, pressurizer, reactor coolant pumps, and the pressurizer surge line, were analyzed for gue usage for the original 40-year life of the plant in accordance with ASME Code, tion III, requirements for Class 1 components using transient conditions that are representative hose expected to occur during plant operation and that are sufficiently severe or frequent to be ossible significance to component cyclic behavior. An assumed number of occurrences of h of the design transients during the plant lifetime were used as input to the design basis gue calculations. Evaluations have shown that the assumed number of occurrences are servatively 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 nting through the Metal Fatigue of Reactor Coolant Pressure Boundary program, the design gue analyses will remain valid for 60 years of plant operation.

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

ng systems designed to this Code were evaluated for thermal expansion cycles, and a thermal ansion stress range reduction factor was to be applied if cycling was determined to be essive. 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 tor coolant loop piping. An evaluation of these transients concluded that thermal cycling of reactor coolant loop piping will remain well below the 7000 thermal expansion cycles allowed USAS B31.1.0.

.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 CAP-14574-A (Reference 5) safety evaluation report, that insurge/outsurge fatigue effects on pressurizer lower head and the surge line must be evaluated for license renewal. These effects e been evaluated as part of the Metal Fatigue of Reactor Coolant Pressure Boundary program critical locations in the pressurizer lower head and in the surge line, including the pressurizer hot leg nozzles. The highest projected 60-year Cumulative Usage Factor (CUF) for these tions, at a pressurizer heater penetration, is less than the design limit.

and Components GSI-190 addressed fatigue life of metal components and was closed by the NRC in ember 1999 (Reference 6). In the closure letter, however, the NRC concluded that licensees uld address the effects of the reactor coolant environment on the fatigue life of selected ponents as aging management programs are formulated in support of license renewal.

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

  • 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 tions, the four piping locations required the development of specific fatigue evaluations, ch have been performed based on the guidance of ASME B&PV Code,Section III. The Surge e Hot Leg Nozzle and the Charging Line Nozzle evaluations were based on the ASME Code, tion III, 2001 edition with addenda through 2003 and the Safety Injection Cold Leg Nozzle Residual Heat Removal System Tee at Safety Injection Accumulator Line evaluations were ed 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 REG/CR-6260 locations concluded that fatigue limits continue to be met for 60 years, based projected plant operation. In addition, the Metal Fatigue of Reactor Coolant Pressure ndary program confirms the assumptions for projected plant operation and manages fatigue these locations through the period of extended operation.

4.2.2 Fatigue of Non-ASME Class 1 Components

.2.2.1 Non-Class 1 Piping Non-Class 1 piping systems were designed and constructed to the requirements of USAS

.1.0-1967. There is no general requirement in this Code for an explicit fatigue analysis.

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

0 full temperature thermal expansion cycles without penalty.

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

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

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

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

.3 Environmental Qualification of Electric Equipment Regulation 10 CFR Part 50 requires that certain categories of systems, structures and ponents be designed to accommodate the effects of both normal and accident environmental ditions, and that design control measures be employed to ensure the adequacy of these gns. Also, 10 CFR 50.49 specifies that electrical equipment that is important to safety and is ated in a harsh environment must be qualified for the lifetime of the plant such that the ipment 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 me-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 o be refurbished, replaced or have their qualification extended prior to reaching the aging ts established in the evaluation. Re-analysis of aging evaluations to extend the qualifications omponents is performed on a routine basis as part of the program. Important attributes for the nalysis of aging evaluations include analytical methods, data collection and reduction hods, underlying assumptions, acceptance criteria and corrective actions (if acceptance eria are not met). Continued implementation of the Environmental Qualification (EQ) of ctric Components aging management program for the period of extended operation ensures the requirements of 10 CFR 50.49 will continue to be met.

The design specification for the Reactor Containment Vessel provided design input for the ber of temperature variations and pressurization cycles during the life of the vessel, which assumed to be 200 temperature variations, and 40 pressurization cycles. The operating perature of the vessel stays relatively constant during normal plant operation as the Shield lding effectively isolates the Reactor Containment Vessel from outdoor weather, and perature variations are only expected during plant shutdown periods. The temperature ations of the Reactor Containment Vessel can be correlated to plant heat-up and cooldown les, which are shown to be limited to 200 over 60 years. Therefore, this assumption will ain valid through the period of extended operation. The Reactor Containment Vessel operates ssentially atmospheric pressure, and the vessel would experience a pressurization cycle during grated leak rate testing (that is currently scheduled at 10-year intervals) or under accident ditions. Therefore, the 40 pressurization cycles specified will remain bounding for 60 years of ration.

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

The penetration assemblies, including the bellows, were designed in accordance with USAS

.1.0 Power Piping Code. No fatigue analyses or specified cyclic loading limits were identified the penetration assemblies.

.5 Other Plant-Specific Time-Limited Aging Analyses 4.5.1 Crane Load Cycle Limit Overhead cranes were originally designed to Specification 61 of the Electric Overhead ne Institute (EOCI) (Reference 8). EOCI-61 did not require a specific fatigue or load-cycle lysis. However, cranes subject to the requirements of NUREG-0612 (Reference 9) were sequently evaluated to the guidelines of Specification 70 of the Crane Manufacturers ociation of America (CMAA-70) (Reference 10), which includes an evaluation of load cycles.

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

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

The potential for crack propagation in the reactor coolant pump motor flywheel was luated due to the potential for flywheel failure that could inhibit pump coastdown or result in sile generation. The original flywheel crack growth analysis was updated to credit the

-before-break analysis that results in a limited postulated break size and lower reactor coolant p overspeed conditions, and to account for a 60-year operating life of the motor flywheel.

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

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

4.5.3 Leak-Before-Break 10 CFR 50, Appendix A, Criterion 4 allows for the use of leak-before-break (LBB) hodology for excluding the dynamic effects of postulated ruptures in reactor coolant system ng. The fundamental premise of the LBB methodology is that the materials used in nuclear er plant piping are sufficiently tough that even a large through-wall crack would remain stable 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 lyses were reviewed to determine whether the conclusions were affected by extending the rating life of the plant to 60 years. It was determined that, since the cyclic transient mptions used as input to the analyses are bounding for the period of extended operation, only mal aging embrittlement effects on cast austenitic stainless steel (CASS) components needed e re-evaluated. Additionally, the only CASS components considered in the LBB analyses are reactor coolant loop piping and elbows. The reactor coolant loop piping LBB analysis was valuated assuming fully-aged (saturated) material properties for the CASS material and found emain acceptable.

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

NUREG-1801 identifies that CASS reactor coolant system components may be susceptible educed fracture toughness in a high temperature environment due to the effects of thermal g embrittlement of the steel. Reactor coolant loop piping, valves, and pumps are constructed m CASS material. Thermal aging embrittlement of pumps and valves is managed by the ME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program. An luation of the susceptibility of loop piping to thermal aging and the potential for flaw growth he piping due to reduced fracture toughness has been performed consistent with the mmendations of NUREG-1801,Section XI.M12, Thermal Embrittlement of Cast Austenitic nless Steel (CASS).

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

15.4 References 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.

NRC Regulatory Guide 1.99, Rev 2, Radiation Embrittlement of Reactor Vessel Materials, May 1968.

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.

WCAP-16609, Revision 0, Master Curve Assessment of Kewaunee Power Station Reactor Vessel Weld Metal, Westinghouse Electric Company, LLC, October, 2006.

WCAP-14574-A, License Renewal Evaluation: Aging Management Evaluation for Pressurizers, Westinghouse Electric Company, LLC, December, 2000.

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.

NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.

Electric Overhead Crane Institute (EOCI) Specification #61.

NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.

Electric Overhead Traveling Cranes, 1975 WCAP-15338-A, Evaluation of Cracking Associated with Weld Deposited Cladding in PWR Vessels, Westinghouse Electric Company, LLC, October, 2002.

TLAA SUPPORT PROGRAMS

.1 Environmental Qualification (EQ) of Electric Components gram Description The Environmental Qualification (EQ) of Electric Components program is an existing gram that corresponds to NUREG-1801 (Reference 1),Section X.E1, Environmental lification (EQ) of Electric Components.

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

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

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

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

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

gue for ASME Code Class 1 components. The program monitors and tracks the critical mal and pressure transients to ensure that cycle occurrence limits are not exceeded such that ASME Class 1 vessels and pressurizer surge line fatigue analyses assumptions are maintained.

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

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

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

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

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

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

mmitments

  • Routine Update of Cycle Count and Fatigue Usage Status The Metal Fatigue of Reactor Coolant Pressure Boundary program will be enhanced to ude a routine assessment of the transient cycle count totals and fatigue usage status for nitored locations, including an action limit for the initiation of corrective action.
  • 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 ordance with ASME B&PV Code Section III, Subsection NB-3200 guidance was performed etermine the CUF, considering the effects of the reactor coolant environment. The analysis firmed 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 ewal Commitments, Item 41.

  • Additional Fatigue Evaluation (Pressurizer Lower Head)

The Metal Fatigue of Reactor Coolant Pressure Boundary program will perform a fatigue luation of the pressurizer lower head and surge line that is consistent with the requirements of ME B&PV Code,Section III, NB-3200 and will determine the cumulative fatigue usage ugh 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 ign basis ASME Class 1 component fatigue evaluations to determine whether the REG/CR-6260-based components that have been evaluated for the effects of the reactor lant environment on fatigue usage are the limiting components for the Kewaunee plant figuration. If more limiting components are identified, the most limiting component will be luated 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 NUREG-1801, Generic Aging Lessons Learned, Rev. 1, U.S. Nuclear Regulatory Commission, September 2005.

NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, March 1995.

The requirements of 10 CFR 54.21(c) stipulate that the application for a renewed license uld include a list of plant-specific exemptions granted pursuant to 10 CFR 50.12 and that are ed 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 granted in a May, 2001 letter from NRC (Reference 1). The exemption remains in effect and ased on a time-limited aging analysis. Specifically, the NRC issued an exemption to: (1) blish the use of a new methodology to meet the requirements of Appendix G to 10 CFR 50; establish the use of a new methodology to meet the requirements of 10 CFR 50.61; and (3) dify the basis for the Kewaunee reactor pressure vessel surveillance program (required by endix H to 10 CFR 50) to incorporate the acquisition of fracture toughness data. The new hodology for assessing the RPV circumferential beltline weld is based on the use of the 1997 tion of ASTM Standard Test Method E-1921 and ASME Code Case N-629. The exemption necessary for the reactor vessel beltline weld to meet the pressurized thermal shock criterion 0 CFR 50.61.

15.6 References6 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.

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

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea The ASME Section XI Inservice Inspection, ASME Section XI At least 2 years Subsections IWB, IWC, and IWD program Inservice Inspection, prior to entering will be enhanced to (1) participate in the Subsections IWB, the period of industry programs for investigating and IWC, and IWD extended managing aging effects on reactor internals; operation.

(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.

The ASME Section XI Inservice Inspection, ASME Section XI At least 2 years Subsections IWB, IWC, and IWD program Inservice Inspection, prior to entering will be enhanced to include identification of Subsections IWB, the period of the limiting susceptible cast austenitic IWC, and IWD extended stainless steel reactor vessel internals operation.

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.

The Bolting Integrity program will be Bolting Integrity Prior to the period enhanced to further incorporate applicable of extended EPRI and industry bolting guidance. Topic operation.

enhancements will include proper joint assembly, torque values, gasket types, use of lubricants, and other bolting fundamentals.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea The Buried Piping and Tanks Inspection Buried Piping and Prior to the period program will be enhanced to perform visual Tanks Inspection of extended inspections of a representative sample of operation, Letter 10-548; material/protective measure combinations Response to RAI and for in-scope buried piping and tanks.

B2.1.7-3a During the first The following materials are utilized in (Reference 1) ten (10) years of buried applications with the associated the period of protective measures:

extended

  • Steel (including cast iron)/coated, operation
  • Steel/coated and wrapped, and
  • Steel/uncoated, and During the second
  • Stainless steel/coated and wrapped ten (10) years of Visual inspections of the external surface of the period of the components will be performed to extended identify damaged wrapping (if present), operation.

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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea nt) 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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea nt) 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.

The Compressed Air Monitoring program Compressed Air Prior to the period will be enhanced to incorporate the Monitoring of extended compressed air system testing and operation.

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.

The External Surfaces Monitoring program External Surfaces Prior to the period will be enhanced to inspect the accessible Monitoring of extended external surfaces of in-scope components, operation.

piping, supports, structural members, and structural commodities, in the infrequently accessed areas, consistent with the criteria used in other plant areas.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea The External Surfaces Monitoring program External Surfaces Prior to the period will be enhanced to provide training for Monitoring of extended Operations, Engineering, and Health Physics operation.

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.

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

The Fire Protection program fire barrier Fire Protection Prior to the period penetration seal inspections will be revised of extended to include the elastomer Shield Building fire operation.

boots.

0 The Fire Protection program inspections of Fire Protection Prior to the period the reactor coolant pump oil collection of extended system will be revised to include additional operation.

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.

1 The Fuel Oil Tank Inspections program will Fuel Oil Tank Prior to the period be enhanced to provide guidance for the Inspections of extended periodic draining, cleaning and inspection operation.

activities.

2 The Inspection of Overhead Heavy Load Inspection of Prior to the period and Refueling Handling Systems program Overhead Heavy of extended will be enhanced to clarify the requirements Load and Refueling operation.

of visual inspection of structural members, Handling Systems including structural bolting, of the in-scope heavy load and refueling handling cranes and associated equipment.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 3 The Metal Enclosed Bus program will be Metal Enclosed Bus Prior to the period enhanced to include augmented periodical of extended Letter 09-469; visual inspections of the MEB internal operation.

Response to RAI surfaces, bus supports, bus insulation, taped B2.1.18-1 Thereafter, the joints and boots for signs of degradation or (Reference 2) inspection of all aging.

MEB will not exceed a 10-year interval and the inspection of the sample of bolted connections will not exceed a 5-year interval 4 The Non-EQ Electrical Cables and Non-EQ Electrical Prior to the period Connections program will be established. Cables and of extended The program will periodically visually Connections operation.

inspect for accessible electrical cables and Thereafter, the connections installed in an adverse localized inspections will equipment environment. Should an adverse not exceed a localized environment be observed, a 10-year interval.

representative sample of electrical cables and connections installed within that environment will be visually inspected for jacket surface anomalies.

5 The Non-EQ Electrical Cable Connections Non-EQ Electrical Prior to the period program will be established. The program Cable Connections of extended will perform a one-time inspection, on a operation.

sampling basis, to confirm the absence of loosening of bolted connections.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 6 The Non-EQ Inaccessible Medium-Voltage Non-EQ Inaccessible Prior to the period Cables program will be established. The Medium-Voltage of extended program will periodically inspect the Cables operation.

in-scope manhole/pulling pit for water Letter 10-447; Second Thereafter, the collection and will remove water, if Annual Update manhole/pulling required. The program will periodically (Reference 3) pit inspections perform a test on the in-scope cables to will not exceed a provide an indication of the condition of the Letter 10-548 2-year interval.

conductor insulation. Response to RAI B2.1.7-3a Thereafter, the (Reference 1) cable testing will not exceed a 10-year interval.

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

established. The program will periodically Sensitive, Thereafter, the perform a proven cable system test for High-Voltage, cable testing and detecting deterioration of the insulation Low-Level Signals calibration system for those electrical cables and reviews will not connections disconnected during calibration, exceed a 10-year or will periodically review the results and interval.

findings of calibrations for those electrical cables that remain connected during the calibration process.

8 The Open-Cycle Cooling Water System Open-Cycle Cooling Prior to the period program will be enhanced to add the Water System of extended applicable aging effects as inspection criteria operation.

for the Circulating Water System underwater visual inspections.

9 The Reactor Vessel Surveillance program Reactor Vessel Prior to the period will be enhanced to include the applicable Surveillance of extended limitations on operating conditions to which operation.

the surveillance capsules were exposed (e.g.

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

0 The Reactor Vessel Surveillance program Reactor Vessel Prior to the period will be enhanced to include requirements for Surveillance of extended storing, and possible recovery, of tested and operation.

untested capsules (removed from the Reactor Vessel after August 31, 2000).

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 1 The Selective Leaching of Materials Selective Leaching of Prior to the period program will be established. The program Materials of extended will perform a one-time visual inspection, operation.

and hardness measurement or qualitative examination of selected components within the scope of license renewal for selective leaching.

2 The Structures Monitoring Program will be Structures Monitoring Prior to the period enhanced to clearly define structures, Program of extended structural elements, and miscellaneous operation.

Letter 09-469; structural commodities that are in scope.

Response to RAI Defined scope to include the MEB enclosure B2.1.18-2 assemblies, structural supports, and (Reference 2) enclosure seals.

3 The Structures Monitoring Program will be Structures Monitoring Prior to the period enhanced to monitor groundwater quality Program of extended and verify that it remains non-aggressive to operation.

below-grade concrete.

4 The Structures Monitoring Program will be Structures Monitoring Prior to the period enhanced to improve criteria for detection of Program of extended aging effects for the underwater visual operation.

inspections of the in-scope structures.

5 The Work Control Process program will be Work Control Process Prior to the period established. The program will perform of extended Letter 09-597; one-time inspections as a verification of the operation.

Changes to the Work effectiveness of chemistry control programs.

Control Process AMP The program will also perform visual (Reference 4) 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.

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

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 7 Deleted Letter 09-597; Changes to the Work Control Process AMP (Reference 4) 8 The Metal Fatigue of Reactor Coolant Metal Fatigue of Prior to the period Pressure Boundary program will be Reactor Coolant of extended enhanced to include a routine assessment of Pressure Boundary operation.

the transient cycle count totals and fatigue usage status for monitored locations, including an action limit for the initiation of corrective action.

9 The following will be further evaluated as Environmental Report Prior to the period part of Dominion's ongoing performance - SAMA Analysis of extended improvement programs: operation.

Letter 09-028

  • SAMA 160: Install Emergency Diesel (Reference 5) and Generator exhaust duct insulation. Letter 09-291 (Reference 6)
  • Concurrent implementation of SAMAs 81,160,166 and 167.
  • Implementation of temporary screenhouse ventilation.

0 Quarterly laboratory testing of fuel oil Letter 09-680; Prior to the period samples for water, sediment and particulates Response to RAI of extended will be performed on the Emergency Diesel B2.1.14-3 operation.

Generators and Technical Support Center (Reference 7)

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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 1 The Work Control Process Program will Letter 09-469 Prior to the period provide for a one-time-inspection of the Response to RAI of extended Emergency Diesel Generators (EDG) Day B2.1.15-1 operation.

Tanks and the Technical Support Center (Reference 2)

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.

2 The 14 potentially cost beneficial SAMAs Environmental Report Prior to the period identified in LRA Appendix E, Attachment - SAMA Analysis of extended F, will be further evaluated as part of operation.

Letter 08-0462 Dominion's ongoing performance (Reference 8) improvement programs 3 Develop a plan for identification and Letter 09-760; Prior to the period remediation of reactor refueling cavity liner Response to RAI of extended leakage to be implemented during the period B2.1.31-4a operation.

of extended operation. (Reference 9) 4 At least one core bore sample will be taken Letter 09-760; Prior to the end of from the waste drumming room reinforced Response to RAI 2011 concrete ceiling below the spent fuel pool. B2.1.31-5a The core sample location and depth will be (Reference 9) sufficient to validate the strength of the Letter 10-093; concrete and the extent of any degradation.

Supplemental The core sample will be tested for Response to RAI compressive strength and will be subject to B2.1.31-5a petrographic examination. Reinforcing steel (Reference 15) in the core sample area will be exposed and inspected for material condition.

5 Develop an action plan for identification and Letter 09-760; Prior to the period remediation of spent fuel pool liner leakage Response to RAI of extended to be implemented during the period of B2.1.31-5a operation.

extended operation. (Reference 9)

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 6 If SFP liner leakage persists during the Letter 09-760; Prior to the end of period of extended operation, an additional Response to RAI the first ten years concrete core sample will be taken from the B2.1.31-5a of extended waste drumming room reinforced concrete (Reference 9) operation 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.

7 Perform a VT-1 visual examination of the Letter 09-777; Prior to the period stainless steel cladding of a safety injection Response to RAI of extended pump for indications of cracking or 3.2.2.2.2 operation.

corrosion due to cladding breach. (Reference 10) 8 The boron carbide surveillance program, Letter 09-777; During the period which includes neutron attenuation testing, Supplemental of extended will continue to be performed during the Response to RAI operation.

period of extended operation every 3 years. 3.3.2.2.6-1 (Reference 10) 9 A surveillance program will be implemented Letter 09-777; Prior to 2017.

to perform verification that the Boral spent Supplemental Surveillance fuel storage rack neutron absorber B-10 Response to RAI program will be areal density is maintained within the 3.3.2.2.6-2 performed every bounds of the spent fuel pool criticality (Reference 10) 10 years analysis. Alternatively, the criticality thereafter.

analysis for the spent fuel pool will be revised to eliminate credit for the Boral neutron absorber material.

0 Implement nitrate monitoring for the Letter 10-008; Prior to the period Component Cooling System on a frequency Response to RAI of extended consistent with the existing monitoring for B2.1.8-3a operation.

ammonia. (Reference 11)

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 1 Perform a fatigue analysis of the surge line Letter 10-324; Prior to the period hot leg nozzle and the charging line nozzle Completion of of extended in accordance with ASME B&PV Code Commitment 41 operation.

Section III, Subsection NB-3200 guidance related to RAI B3.2-2 (Complete) and determine the CUF, considering the (Reference 12) effects of the reactor coolant environment.

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

2 For Examination Category B-J, Item No. Letter 10-033; During each B9.21, eight ASME Class 1 small-bore Supplemental 10-year lSI circumferential welds will receive Response to RAI inspection interval volumetric and surface examinations during B2.1.2-1 during the period each 10-year lSI inspection interval during (Reference 13) of extended the period of extended operation. operation.

3 Ten volumetric examinations of ASME Letter 10-665; Four volumetric Class 1 small-bore socket welds will be Supplemental examinations or performed using a demonstrated, Response to RAI two destructive nuclear-industry endorsed, inspection B2.1.2-2 examinations (or methodology that can detect cracking within (Reference 14) an equivalent the specified examination volume, if a combination of methodology becomes available. In the examinations) event that a demonstrated, nuclear-industry prior to the period endorsed, inspection methodology is not of extended available, destructive examinations of socket operation.

welds will be substituted for volumetric Remaining non-destructive examinations. Each examinations destructive weld examination will be within three years considered equivalent to performing two of entering the volumetric weld examinations, such that a period of maximum of five destructive examinations extended will be performed.

operation.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 4 Core samples will be obtained from the Letter 10-093; Prior to the period inside surface of a concrete wall (below the Supplemental of extended groundwater table elevation) or from the Response to RAI operation foundation basemat in the vicinity of the B2.1.31-3a groundwater wells for which average (Reference 15) 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.

5 In the event that the chloride content in the Letter 10-093; Prior to the end of groundwater does not decrease to below 500 Supplemental the first ten years ppm within the first ten years of the period Response to RAI of extended of extended operation, core samples will be B2.1.31-3a operation obtained from the inside surface of a (Reference 15) 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.

6 If the results of the core sample testing of the Letter 10-093; Prior to the period waste drumming room reinforced concrete Supplemental of extended ceiling leakage site (related to potential SFP Response to RAI operation.

liner leakage - Commitment 34) indicate B2.1.31-4a degradation of the structural integrity of the (Reference 15) 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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 7 Submit three examples of operating Letter 10-286; Within 2 years experience associated with the Work Control Response to RAI following Process - Internal Surfaces Monitoring B2.1.32-5. implementation of program for NRC staff review in (Reference 16) the Work Control determining the effectiveness of the program Process aging to detect and correct the effects of aging management prior to the loss of intended function. program.

8 The cathodic protection system associated Letter 10-548 During the period with the diesel generator fuel oil storage Response to RAI of extended tanks and protected portions of the fuel oil B2.1.7-3a operation.

lines, and the circulating water system (Reference 1) 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.

9 Recognizing that the EPRI SGMP resolution Letter 10-548 Prior to 2023 is still under development, Kewaunee will Response to RAI perform an inspection of each steam 3.1.2.2.13-1a generator to assess the condition of the (Reference 1) 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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 0 Perform an audit of the Internal Surfaces Letter 10-595 Prior to the period Monitoring portion of the Work Control (Supplemental of extended Process program inspections to confirm that Response to RAI operation and the components representing the leading B2.1.32-5a) every 10 years indicators of aging for each of the (Reference 17) thereafter.

material/environment combinations have Deliberate been inspected at least once during the audit focused period.

inspections will If any scheduled surveillance and be performed maintenance activities which were intended within 5 years of to encompass components as leading the completion of indicators of aging in each of the the audits.

material/environment combinations have not been performed, then deliberate focused inspections of these components will be performed.

1 Dominion Energy Kewaunee, Inc. (DEK) Letter 10-595 Prior to the period will perform a fatigue evaluation of the Supplemental of extended pressurizer lower head and surge line that is Response to RAI operation.

consistent with the requirements of ASME B3.2-2a B&PV Code,Section III, NB-3200 and will (Reference 17) determine the cumulative fatigue usage through the period of extended operation.

2 DEK will perform a review of design basis Letter 10-595 Prior to the period ASME Class 1 component fatigue Supplemental of extended evaluations to determine whether the Response to RAI operation.

NUREG/CR-6260-based components that B3.2-2a have been evaluated for the effects of the (Reference 17) 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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 3 DEK will develop a plan to address the Letter 10-595 Develop a plan potential for failure of the (Reference 17) prior to the period primary-to-secondary pressure boundary due of extended to PWSCC cracking of tube-to-tubesheet operation.

welds. The plan will consist of two Implement the resolution options:

requirements of

1. Perform an analytical evaluation of the the plan prior to steam generator tube-to-tubesheet 2023.

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.

LICENSE RENEWAL COMMITMENTS m Commitment Source Schedulea 4 The Structures Monitoring Program will be Letter 10-707; Prior to the period revised to include the evaluation criteria of Response to RAI of extended ACI 349.3R-96, Chapter 5, as the criteria to B2.1.31-9 operation.

be used when evaluating conditions or (Reference 16) 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.

he period of extended operation is the period of 20 years beyond the expiration date of the units original operating cense.

15.7 References 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.

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.

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.

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.

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.

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.

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.

Application for Renewed License, Letter No. 08-0462 dated August 12, 2008, Appendix E, Attachment F.

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.

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.

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.

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.

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.

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.

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.

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.

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.

Information for the Review of the Kewaunee Power Station License Renewal Application, Letter No.10-707, November 23, 2010.

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