ML083030241

Cooper Nuclear Station, License Renewal Application, Appendix a, Updated Safety Analysis Report Supplement

ML083030241
Person / Time
Site:	Cooper
Issue date:	09/24/2008
From:	Nebraska Public Power District (NPPD)
To:	Office of Nuclear Reactor Regulation
References
NLS2008071
Download: ML083030241 (35)

Text

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-i APPENDIX A UPDATED SAFETY ANALYSIS REPORT SUPPLEMENT TABLE OF CONTENTSA.0INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1A.1AGING MANAGEMENT PROGRAMS AND ACTIVITIES . . . . . . . . . . . . . . . . . . . . . A-1A.1.1Aging Management Programs. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1A.1.1.1Aboveground Steel Tanks Program . . . . . . . . . . . . . . . . . . . . . . . . . . A-2A.1.1.2Bolting Integrity Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2A.1.1.3Buried Piping and Tanks Inspection Program. . . . . . . . . . . . . . . . . . . A-3 A.1.1.4BWR CRD Return Line Nozzle Program. . . . . . . . . . . . . . . . . . . . . . . A-3 A.1.1.5BWR Feedwater Nozzle Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . A-4A.1.1.6BWR Penetrations Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-4A.1.1.7BWR Stress Corrosion Cracking Program . . . . . . . . . . . . . . . . . . . . . A-4 A.1.1.8BWR Vessel ID Attachment Welds Program. . . . . . . . . . . . . . . . . . . . A-4A.1.1.9BWR Vessel Internals Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-4A.1.1.10Containment Inservice Inspection Program. . . . . . . . . . . . . . . . . . . . . A-5 A.1.1.11Containment Leak Rate Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-5A.1.1.12Diesel Fuel Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-6A.1.1.13Environmental Qualification (EQ) of Electric Components Program. . A-6 A.1.1.14External Surfaces Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . A-7A.1.1.15Fatigue Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7A.1.1.16Fire Protection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-8 A.1.1.17Fire Water System Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-9A.1.1.18Flow-Accelerated Corrosion Program. . . . . . . . . . . . . . . . . . . . . . . . . A-10A.1.1.19Inservice Inspection - ISI Program . . . . . . . . . . . . . . . . . . . . . . . . . . . A-10 A.1.1.20Inservice Inspection - IWF Program. . . . . . . . . . . . . . . . . . . . . . . . . . A-11A.1.1.21Masonry Wall Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-11A.1.1.22Metal-Enclosed Bus Inspection Program . . . . . . . . . . . . . . . . . . . . . . A-12 A.1.1.23Neutron Absorber Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . A-12A.1.1.24Non-EQ Bolted Cable Connections Program . . . . . . . . . . . . . . . . . . . A-13A.1.1.25Non-EQ Inaccessible Medium-Voltage Cable Program . . . . . . . . . . . A-13A.1.1.26Non-EQ Instrumentation Circuits Test Review Program. . . . . . . . . . . A-14A.1.1.27Non-EQ Insulated Cables and Connections Program. . . . . . . . . . . . . A-15 Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-iiA.1.1.28Oil Analysis Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-15A.1.1.29One-Time Inspection Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-16A.1.1.30One-Time Inspection - Small-Bore Piping Program . . . . . . . . . . . . . . A-17A.1.1.31Periodic Surveillance and Preventive Maintenance Program . . . . . . . A-18 A.1.1.32Reactor Head Closure Studs Program . . . . . . . . . . . . . . . . . . . . . . . . A-19A.1.1.33Reactor Vessel Surveillance Program. . . . . . . . . . . . . . . . . . . . . . . . . A-19A.1.1.34Selective Leaching Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-20 A.1.1.35Service Water Integrity Program. . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-20A.1.1.36Structures Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-21A.1.1.37Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel Program A-23A.1.1.38Water Chemistry Control - Auxiliary Systems Program . . . . . . . . . . . A-24A.1.1.39Water Chemistry Control - BWR Program. . . . . . . . . . . . . . . . . . . . . A-24 A.1.1.40Water Chemistry Control - Closed Cooling Water Program. . . . . . . . A-24A.1.2Evaluation of Time-Limited Aging Analyses. . . . . . . . . . . . . . . . . . . . . . . . . . . . A-25A.1.2.1Reactor Vessel Neutron Embrittlement. . . . . . . . . . . . . . . . . . . . . . . . A-25A.1.2.1.1Reactor Vessel Fluence. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-25A.1.2.1.2Adjusted Reference Temperature . . . . . . . . . . . . . . . . . . . . . . . A-25A.1.2.1.3Pressure-Temperature Limits. . . . . . . . . . . . . . . . . . . . . . . . . . . A-26A.1.2.1.4Upper-Shelf Energy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-26 A.1.2.1.5Reactor Vessel Circumferential Weld Inspection Relief. . . . . . . A-26A.1.2.2Metal Fatigue. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-27A.1.2.2.1Class 1 Metal Fatigue. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-27A.1.2.2.2Non-Class 1 Metal Fatigue. . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-29A.1.2.2.3Effects of Reactor Water Environment on Fatigue Life. . . . . . . . A-29A.1.2.3Environmental Qualification of Electrical Components . . . . . . . . . . . . A-30 A.1.2.4Fatigue of Primary Containment, Attached Piping, and Components. A-30A.1.2.5Core Plate Plugs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-31A.1.3References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-32A.2LICENSE RENEWAL COMMITMENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-33 Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-1A.0INTRODUCTIONThis appendix provides the information to be submitted in an Updated Safety Analysis Report (USAR) Supplement as required by 10 CFR 54.21(d) for the Cooper Nuclear Station (CNS) License Renewal Application (LRA). Appendix B of the CNS LRA provides descriptions of the programs and activities that manage the effects of aging for the period of extended operation. Section 4 of the LRA documents the evaluations of time-limited aging analyses for the period of extended operation. Appendix B and Section 4 have been used to prepare the summary program and activity descriptions for the CNS USAR Supplement information in this appendix.The information presented in this section will be incorporated into the USAR following issuance of the renewed operating license. Upon inclusion of the USAR Supplement in the CNS USAR, future changes to the descriptions of the programs and activities will be made in accordance with 10 CFR 50.59.The following information will document aging management programs and activities credited in the Cooper Nuclear Station (CNS) license renewal review (Section A.1.1) and time-limited aging analyses evaluated for the period of extended operation (Section A.1.2). References to other sections are to USAR sections, not to sections in the LRA.A.1AGING MANAGEMENT PROGRAMS AND ACTIVITIESThe CNS license renewal application (Reference A.1-1) and information in subsequent related correspondence provided sufficient basis for the NRC to make the findings required by 10 CFR 54.29 (Final Safety Evaluation Report) (Reference A.1-2). As required by 10 CFR 54.21(d), this USAR supplement contains a summary description of the programs and activities for managing the effects of aging (Section A.1.1) and a description of the evaluation of time-limited aging analyses for the period of extended operation (Section A.1.2). The period of extended operation is the 20 years after the expiration date of the original operating license.A.1.1Aging Management ProgramsThe integrated plant assessment for license renewal identified aging management programs necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions consistent with the current licensing basis (CLB) for the period of extended operation. This section describes the aging management programs and activities required during the period of extended operation. All aging management programs will be implemented prior to entering the period of extended operation.CNS quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. The CNS Quality Assurance Program applies to safety-related structures and components. Corrective actions and administrative (document) control for both safety-related and nonsafety-related structures and components are accomplished per the existing CNS Corrective Action Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-2Program and Document Control Program and are applicable to all aging management programs and activities that will be required during the period of extended operation. The confirmation process is part of the Corrective Action Program and includes reviews to assure adequacy of proposed actions, tracking and reporting of open corrective actions, and review of corrective action effectiveness. Any follow-up inspection required by the c onfirmation process is documented in accordance with the Corrective Action Program. The corrective action, confirmation process, and administrative controls of the CNS (10 CFR Part 50, Appendix B) Quality Assurance Program are applicable to all aging management programs and activities required during the period of extended operation.A.1.1.1Aboveground Steel Tanks ProgramThe Aboveground Steel Tanks Program is a new program that will manage loss of material from external surfaces of outdoor, aboveground carbon steel tanks by periodic visual inspection of external surfaces and thickness measurement of locations that are inaccessible for external visual inspection.This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M29, Aboveground Steel Tanks, prior to the period of extended operation.A.1.1.2Bolting Integrity ProgramThe Bolting Integrity Program is an existing program that relies on recommendations for a comprehensive bolting integrity program, as delineated in NUREG-1339, industry recommendations, and Electric Power Research Institute (EPRI) NP-5769, with the exceptions noted in NUREG-1339 for safety-related bolting. The program relies on industry recommendations for comprehensive bolting maintenance, as delineated in EPRI TR-104213 for pressure retaining bolting and structural bolting.The program applies to bolting and torquing practices of safety- and nonsafety-related bolting for pressure retaining components, NSSS component supports, and structural joints. The program addresses all bolting regardless of size except reactor head closure studs, which are addressed by the Reactor Head Closure Studs Program [Section A.1.1.32]. The program includes periodic inspection of closure bolting for signs of leakage that may be due to crack initiation, loss of preload, or loss of material due to corrosion. The program also includes preventive measures to preclude or minimize loss of preload and cracking.The Bolting Integrity Program will be enhanced as follows.

• Include guidance from EPRI NP-5769 and EPRI TR-104213 for material selection and testing, bolting preload control, ISI, plant operation and maintenance, and evaluation of the structural integrity of bolted joints.

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• Clarify that actual yield strength is used in selecting materials for low susceptibility to stress corrosion cracking, that the use of lubricants containing MoS 2 is prohibited for bolting at CNS, and that proper gasket compression will be visually verified following assembly.*Include guidance from EPRI NP-5769 and EPRI TR-104213 for replacement of non-Class 1 bolting and disposition of degraded structural bolting.Enhancements will be implemented prior to the period of extended operation.A.1.1.3Buried Piping and Tanks Inspection ProgramThe Buried Piping and Tanks Inspection Program is a new program that will include (a)preventive measures to mitigate corrosion and (b) inspections to manage the effects of corrosion on the pressure-retaining capability of buried carbon steel, gray cast iron, and stainless steel components. Preventive measures will be in accordance with standard industry practice for maintaining external coatings and wrappings. Buried components will be inspected when excavated during maintenance. If trending within the corrective action program identifies susceptible locations, the areas with a history of corrosion problems are evaluated for the need for additional inspection, alternate coating, or replacement.Prior to entering the period of extended operation, plant operating experience will be reviewed to verify that an inspection occurred within the past ten years. If an inspection did not occur, a focused inspection will be performed prior to the period of extended operation. A focused inspection will be performed within the first ten years of the period of extended operation, unless an opportunistic inspection occurs within this ten-year period. A "focused inspection" is defined as an inspection performed in areas with a history of corrosion problems and in areas with the highest likelihood of corrosion problems.This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M34, Buried Piping and Tanks Inspection, prior to the period of extended operation.A.1.1.4BWR CRD Return Line Nozzle ProgramThe BWR Control Rod Drive (CRD) Return Line Nozzle Program is an existing program. Under this program, CNS has cut and capped the CRD return line nozzle to mitigate fatigue cracking and continues Inservice Inspection (ISI) examinations using ASME Section XI to monitor the effects of crack initiation and growth on the intended function of the control rod drive return line nozzle. ISI examinations include ultrasonic inspection of the nozzle inside radius section and nozzle-to-vessel weld. CNS also conducts UT examinations of the CRD return line nozzle-to-cap weld in accordance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) document BWRVIP-75-A as part of the BWR Stress Corrosion Cracking Program [Section A.1.1.7

].

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-4A.1.1.5BWR Feedwater Nozzle ProgramThe BWR Feedwater Nozzle Program is an existing program. Under this program, CNS has removed feedwater nozzle cladding and installed a double piston ring, triple thermal sleeve sparger to mitigate cracking. This program implements enhanced inservice inspection (ISI) of the feedwater nozzles in accordance with the requirements of ASME Section XI, Subsection IWB and the recommendation of General Electric (GE) NE 523-A71-0594-A to detect cracking.A.1.1.6BWR Penetrations ProgramThe BWR Penetrations Program is an existing program that includes (a) inspection and flaw evaluation in conformance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) documents BWRVIP-27-A and BWRVIP-49-A and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 to ensure the long-term integrity of vessel penetrations and nozzles. A.1.1.7BWR Stress Corrosion Cracking ProgramThe BWR Stress Corrosion Cracking Program is an existing program that includes (a) preventive measures to mitigate intergranular stress corrosion cracking (IGSCC), and (b) inspection and flaw evaluation to monitor IGSCC and its effects on reactor coolant pressure boundary components made of stainless steel, CASS, or nickel alloy. CNS has taken actions to prevent IGSCC and will continue to use materials resistant to IGSCC for component replacements and repairs following the reco mmendations delineated in NUREG-0313, Generic Letter 88-01, Generic Letter 88-01 Supplement 1, and the staff-approved BWRVIP-75-A report. Inspection of piping identified in NRC Generic Letter 88-01 to detect and size cracks is performed in accordance with the staff positions on schedule, method, personnel qualification, and sample expansion included in the generic letter and the staff-approved BWRVIP-75-A report.A.1.1.8BWR Vessel ID Attachment Welds ProgramThe BWR Vessel ID Attachment Welds Program is an existing program that includes (a)inspection and flaw evaluation in accordance with the guidelines of staff-approved boiling water reactor vessel and internals project (BWRVIP) BWRVIP-48-A and (b) monitoring and control of reactor coolant water chemistry in accordance with the guidelines of BWRVIP-130 (EPRI Report 1008192) to ensure the long-term integrity and safe operation of reactor vessel inside diameter (ID) attachment welds and support pads.A.1.1.9BWR Vessel Internals ProgramThe BWR Vessel Internals Program is an existing program that includes (a) inspection, flaw evaluation, and repair in conformance with the applicable, staff-approved BWR reactor vessel and internals project (BWRVIP) documents and (b) monitoring and control of reactor coolant Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-5water chemistry in accordance with the guidelines of BWRVIP-130 to ensure the long-term integrity of vessel internal components. In addition, the BWR Vessel Internals Program includes inspection of the steam dryer in accordance with BWRVIP-139 guidance.The BWR Vessel Internals Program will be enhanced as follows.

• Include actions to replace the plugs in the core plate bypass holes based on their qualified life.Enhancements will be implemented prior to the period of extended operation.A.1.1.10Containment Inservice Inspection ProgramThe Containment Inservice Inspection Program is an existing program that manages loss of material and cracking for the primary containment and its integral attachments. The program uses the ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition, through the 2003 Addenda.Visual inspections for IWE monitor loss of material of the steel containment shells and their integral attachments; containment hatches and airlocks; moisture barriers; and pressure-retaining bolting by inspecting surfaces for evidence of flaking, blistering, peeling, discoloration, and other signs of distress. The Containment Inservice Inspection Program will be enhanced as follows.

• Provide guidance for surfaces requiring augmented examination to require accessible areas to be examined using a visual examination method and surface areas not accessible on the side requiring augmented examination to be examined using an ultrasonic thickness measurement method in accordance with IWF-2500 (b).

• Provide guidance to document material loss in a local area exceeding ten percent of the nominal containment wall thickness or material loss in a local area projected to exceed ten percent of the nominal containment wall thickness before the next examination in accordance with IWE-3511.3 for volumetric inspections.Enhancements will be implemented prior to the period of extended operation.A.1.1.11Containment Leak Rate ProgramThe Containment Leak Rate Program is an existing program. As described in 10 CFR Part 50, Appendix J, containment leak rate tests are required to assure that (a) leakage through reactor containment and systems and components penetrating containment shall not exceed allowable values specified in technical specifications or associated bases and (b) periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-6and repairs are made during the service life of containment, and systems and components penetrating containment. The program utilizes 10 CFR 50 Appendix J, Option B, and the guidance in NRC Regulatory Guide 1.163 and NEI 94-01.A.1.1.12Diesel Fuel Monitoring ProgramThe Diesel Fuel Monitoring Program is an existing program that entails sampling to ensure that adequate diesel fuel quality is maintained to prevent loss of material in fuel systems. Exposure to fuel oil contaminants such as water and microbiological organisms is minimized by periodic sampling and analysis, draining and cleaning of tanks, and verifying the quality of new fuel oil before its introduction into the storage tanks. Sampling and analysis activities are in accordance with technical specifications for fuel oil purity and the guidelines of ASTM Standards ASTM D4057 and D975.The One-Time Inspection Program [Section A.1.1.29] describes inspections planned to verify that the Diesel Fuel Monitoring Program has been effective at managing aging effects.The Diesel Fuel Monitoring Program will be enhanced as follows.

• Use ASTM Standard D4057 for sampling of the diesel fire pump fuel oil storage tank.

• Include periodic visual inspections and cleaning, as well as ultrasonic bottom surface thickness measurement, of the diesel fuel oil day tanks, the diesel fuel oil storage tanks, and the diesel fire pump fuel oil storage tank.

• Include periodic multilevel sampling of the diesel fuel oil day tanks and the diesel fire pump fuel oil storage tank.

• Provide the acceptance criterion of < 10 mg/l for the determination of particulates in the diesel fire pump fuel oil storage tank.

• Specify acceptance criterion for UT thickness measurements of the bottom surfaces of the diesel fuel oil day tanks, the diesel fuel oil storage tanks, and the diesel fire pump fuel oil storage tank.Enhancements will be implemented prior to the period of extended operation.A.1.1.13Environmental Qualification (EQ) of Electric Components ProgramThe Environmental Qualification (EQ) of Electric Components Program is an existing program that manages the effects of thermal, radiation, and cyclic aging through the use of aging evaluations based on 10 CFR 50.49(f) qualification methods. As required by 10 CFR 50.49, EQ components are refurbished, replaced, or their qualification is extended prior to reaching the Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-7aging limits established in the evaluation. Some aging evaluations for EQ components are time-limited aging analyses (TLAAs) for license renewal.A.1.1.14External Surfaces Monitoring ProgramThe External Surfaces Monitoring Program is an existing program that inspects external surfaces of components subject to aging management review. The program is also credited with managing loss of material from internal surfaces for situations in which internal and external material and environment combinations are the same such that external surface condition is representative of internal surface condition. This program does not manage aging effects on structures.Surfaces that are inaccessible during plant operations are inspected during refueling outages. Surfaces that are insulated are inspected when the external surface is exposed (i.e., during maintenance). Surfaces are inspected at frequencies to assure the effects of aging are managed such that applicable components will perform their intended function during the period of extended operation.The External Surfaces Monitoring Program will be enhanced as follows.

• Clarify that periodic inspections of systems in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(1) and (a)(3) will be performed. Inspections shall include areas surrounding the subject systems to identify hazards to those systems. Inspections of nearby systems that could impact the subject systems will include SSCs that are in scope and subject to aging management review for license renewal in accordance with 10 CFR 54.4 (a)(2). This enhancement will be implemented prior to the period of extended operation.

A.1.1.15Fatigue Monitoring ProgramThe Fatigue Monitoring Program is an existing program that tracks the number of critical thermal and pressure transients for selected reactor coolant system components, in order not to exceed design limits on fatigue usage. The program ensures the validity of analyses that explicitly assumed a fixed number of thermal and pressure transients by assuring that the actual effective number of transients does not exceed the assumed limit.This program also addresses the effects of the coolant environment on component fatigue life by assessing the impact of the reactor coolant environment on a sample of critical components for the plant.The Fatigue Monitoring Program will be enhanced as follows.

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• Consideration of the effect of the reactor water environment will be accomplished through implementation of one or more of the following options for the feedwater nozzles, core spray nozzles and RHR pipe transition.(1)Update the fatigue usage calculations using refined fatigue analyses to determine valid CUFs less than 1.0 when accounting for the effects of reactor water environment. This includes applying the appropriate F en factors to valid CUFs determined using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case). (2)Repair or replace the affected locations before exceeding a CUF of 1.0.

• The CNS Fatigue Monitoring Program will be enhanced to require the recording of each transient associated with the actuation of a safety/relief valve (SRV).Enhancements will be implemented at least two years prior to entering the period of extended operation.A.1.1.16Fire Protection ProgramThe Fire Protection Program is an existing program that includes a fire barrier inspection and a diesel-driven fire pump inspection. The fire barrier inspection requires periodic visual inspection of fire barrier penetration seals, fire dampers, fire stops, fire wraps, fire barrier walls, ceilings, and floors, and periodic visual inspection and functional tests of fire rated doors to ensure that their operability is maintained. The diesel-driven fire pump inspection requires that the pump and its driver be periodically tested and inspected to ensure that diesel engine fuel supply lines can perform their intended functions. The Fire Protection Program also includes periodic inspection and testing of the CO 2 and Halon fire suppression systems.The Fire Protection Program will be enhanced as follows.

• Explicitly state that the diesel fire pump engine sub-systems (including the fuel supply line) shall be observed while the engine is running. Acceptance criteria will be revised to verify that the diesel engine does not exhibit signs of degradation while running, such as excessive fuel oil, lube oil, coolant, or exhaust gas leakage.

• Specify that diesel fire pump engine carbon steel exhaust components are inspected for evidence of corrosion or cracking at least once every five years.

• Require visual inspections of fire damper framing to check for signs of degradation.

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• Require visual inspections of the Halon and CO 2 fire suppression systems at least once every six months to check for signs of degradation in a manner suitable for trending.

• Include inspection of cardox hose reels for corrosion. Acceptance criteria will be enhanced to verify no unacceptable corrosion.

• Include visual inspections of concrete flood curbs, manways, hatches, and hatch covers on an 18-month basis to check for signs of degradation.Enhancements will be implemented prior to the period of extended operation.A.1.1.17Fire Water System ProgramThe Fire Water System Program is an existing program that applies to water-based fire protection systems that consist of sprinklers, nozzles, fittings, valves, hydrants, hose stations, standpipes, and aboveground and underground piping and components that are tested in accordance with applicable National Fire Protection Association (NFPA) codes and standards. Such testing assures functionality of systems. To determine if significant corrosion has occurred in water-based fire protection systems, periodic flushing, system performance testing and inspections are conducted. Also, many of these systems are normally maintained at required operating pressure and monitored such that leakage resulting in loss of system pressure is immediately detected and corrective actions initiated.In addition, wall thickness evaluations of fire protection piping are periodically performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion.A sample of sprinkler heads will be tested or replaced using the guidance of NFPA-25 (2002 edition), Section 5.3.1.1.1. NFPA-25 states, "Where sprinklers have been in place for 50 years, they shall be replaced or representative samples from one or more sample areas shall be submitted to a recognized testing laboratory for field service testing." This sampling will be repeated every ten years after initial field service testing per the guidance of NFPA-25.The Fire Water System Program will be enhanced as follows.

• Include inspection of hose reels for corrosion. Acceptance criteria will be enhanced to verify no unacceptable corrosion.

• Include visual inspection of spray and sprinkler system internals for evidence of corrosion. Acceptance criteria will be enhanced to verify no unacceptable corrosion.

• Wall thickness evaluations of fire protection piping will be performed on system components using non-intrusive techniques (e.g., volumetric testing) to identify evidence of loss of material due to corrosion. These inspections will be performed before the end Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-10of the current operating term and at intervals thereafter during the period of extended operation. Results of the initial evaluations will be used to determine the appropriate inspection interval to ensure aging effects are identified prior to loss of intended function.

• A sample of sprinkler heads required for 10 CFR 50.48 will be tested or replaced using guidance of NFPA-25 (2002 edition), Section 5.3.1.1.1, before the end of the 50-year sprinkler head service life and at 10-year intervals thereafter during the extended period of operation.Enhancements will be implemented prior to the period of extended operation.A.1.1.18Flow-Accelerated Corrosion ProgramThe Flow-Accelerated Corrosion (FAC) Program is an existing program that applies to safety-related and nonsafety-related carbon steel components and gray cast iron in systems containing high-energy fluids carrying two-phase or single-phase high-energy fluid greater than or equal to two percent of plant operating time per the criteria given in EPRI NSAC-202L.The program, based on EPRI recommendations in NSAC-202L for an effective flow-accelerated corrosion program, predicts, detects, and monitors FAC in plant piping and other pressure retaining components. This program includes (a) an evaluation to determine critical locations, (b)initial operational inspections to determine the extent of thinning at these locations, and (c)follow-up inspections to confirm predictions, or repair or replace components as necessary. The aging effect of loss of material managed by the Flow-Accelerated Corrosion Program is equivalent to the aging effect of wall thinning as defined in NUREG-1801 Volume 2 Table IX.E.The FAC Program will be enhanced as follows.

• Update the System Susceptibility Analysis for the Flow-Accelerated Corrosion Program to reflect the lessons learned and new technology that became available after the publication of NSAC-202L Revision 1.This enhancement will be implemented prior to the period of extended operation.A.1.1.19Inservice Inspection - ISI ProgramThe Inservice Inspection - ISI Program is an existing program that encompasses ASME Section XI Subsection IWB, IWC, and IWD requirements. This program manages loss of material, cracking, and reduction of fracture toughness to assure that the pressure boundary functions of the reactor pressure vessel and reactor coolant system pressure boundary are maintained through the period of extended operation.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-11Regulation 10 CFR 50.55a imposes inservice inspection (ISI) requirements of ASME Code,Section XI, for Class 1, 2, and 3 pressure-retaining components, their integral attachments, and supports in light-water cooled power plants. Inspection, repair, and replacement of these components are covered in Subsections IWB, IWC, and IWD respectively. The program includes periodic visual, surface, and volumetric examination and leakage tests of Class 1, 2, and 3 pressure-retaining components, their integral attachments and supports.The ISI Program is based on ASME Inspection Program B, which has ten-year inspection intervals. Every ten years the program is updated to the latest ASME Section XI code edition and addendum approved by the NRC in 10 CFR 50.55a. On March 1, 2006, CNS entered the fourth ISI interval. The ASME code edition and addenda used for the fourth interval is the 2001 Edition, 2003 Addenda.A.1.1.20Inservice Inspection - IWF ProgramThe Inservice Inspection - IWF Program is an existing program that manages loss of material for ASME Class 1, 2, 3 and MC piping and component supports, bolting, and base plates. The program uses the ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition, 2003 Addenda.The program includes visual inspections of surfaces to manage loss of material. Evidence of corrosion, deformation, misalignment, improper clearances, improper spring settings, damage to close tolerance machined or sliding surfaces, and missing, detached, or loosened support items that may compromise support function or load capacity are detected through visual inspection.The Inservice Inspection - IWF Program will be enhanced as follows.

• Clarify that Class MC piping and component supports are included in the program.

• Clarify that the successive inspection requirements of IWF-2420 and the additional examination requirements of IWF-2430 are applied.Enhancements will be implemented prior to the period of extended operation.A.1.1.21Masonry Wall ProgramThe Masonry Wall Program is an existing program that manages aging effects so that the evaluation basis established for each masonry wall within the scope of license renewal remains valid through the period of extended operation.The program includes visual inspection of all masonry walls identified as performing intended functions in accordance with 10 CFR 54.4. Included components are 10 CFR 50.48-required masonry walls, radiation shielding masonry walls, and masonry walls with the potential to affect Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-12safety-related components. Structural steel components of masonry walls are managed by the Structures Monitoring Program [Section A.1.1.36

].Masonry walls are visually examined at a frequency selected to ensure there is no loss of intended function between inspections. The Masonry Wall Program will be enhanced as follows.

• Clarify that the control house - 161kv switchyard is included in the program.

• Clarify that structures with conditions classified as "acceptable with deficiencies" or "unacceptable" shall be entered into the corrective action program.Enhancements will be implemented prior to the period of extended operation.A.1.1.22Metal-Enclosed Bus Inspection ProgramThe Metal Enclosed Bus Inspection Program is a new program that inspects the following non-segregated phase bus.

• non-segregated bus between the emergency station service transformer and 4.16kV switchgear buses (1F and 1G).

• non-segregated bus between the start-up station service transformer X-winding and 4.16kV switchgear buses (1A and 1B).Inspections of the metal enclosed bus (MEB) will include the bus and bus connections, the bus enclosure assemblies, and the bus insulation and insulators. A sample of the accessible bolted connections will be inspected for loose connections. The bus enclosure assemblies will be inspected for loss of material and elastomer degradation. This program will be used instead of the Structures Monitoring Program for external surfaces of the bus enclosure assemblies. The bus insulation or insulators will be inspected for degradation leading to reduced insulation resistance (IR). These inspections will include visual inspections, as well as quantitative measurements, such as thermography or connection resistance measurements, as required.This program will be implemented prior to the period of extended operation. This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.E4, Metal-Enclosed Bus, prior to the period of extended operation.A.1.1.23Neutron Absorber Monitoring ProgramThe Neutron Absorber Monitoring Program is an existing program that manages loss of material of Boral neutron absorption panels in the spent fuel racks. The program relies on representative Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-13coupon samples mounted in surveillance assemblies located in the spent fuel pool to monitor performance of the absorber material without disrupting the integrity of the storage system.Surveillance assemblies are removed from the spent fuel pool on a prescribed schedule and physical and chemical properties are measured. From this data, the stability and integrity of Boral in the storage cells are assessed.A.1.1.24Non-EQ Bolted Cable Connections ProgramThe Non-EQ Bolted Cable Connections Program is a new program which provides a one-time inspection, on a sampling basis, to confirm the absence of age-related degradation of bolted cable connections due to thermal cycling, ohmic heating, electrical transients, vibration, chemical contamination, corrosion, and oxidation. Connections associated with cables within the scope of license renewal are considered for this program. The factors considered for sample selection will be application (medium and low voltage, defined as <35kV), circuit loading (high loading), and location (high temperature, high humidity, vibration, etc.). The technical basis for the sample selections will be documented. If an unacceptable condition or situation is identified in the selected sample, the corrective action program will be used to evaluate the condition and determine appropriate corrective action.This program will be implemented prior to the period of extended operation.A.1.1.25Non-EQ Inaccessible Medium-Voltage Cable ProgramThe Non-EQ Inaccessible Medium-Voltage Cable Program is a new program that inspects the following underground medium-voltage cables.

• inaccessible medium-voltage cables between the station service water pumps (SWP-1A, 1B, 1C, and 1D) and the 4.16kV safety switchgear

• inaccessible medium-voltage cables between 12.5kV overhead loop and the fire pump motor (FP-MOT-E)

• inaccessible medium-voltage cables between the standby diesel (DG1 and DG2) to the 4.16kV safety busses (1F and 1G)

• inaccessible medium-voltage cables between the 4.16kV non-safety buses (1A and 1B) and the 161kV control house power transformers (located in the 345kV switchyard)The Non-EQ Inaccessible Medium-Voltage Cable Program entails periodic inspections for water collection in cable manholes and periodic testing of cables. In-scope medium-voltage cables (cables with operating voltage from 2kV to 35kV) exposed to significant moisture and voltage will be tested at least once every ten years to provide an indication of the condition of the conductor insulation. Significant moisture is defined as periodic exposures to moisture that last more than a few days (e.g., cable in standing water). Periodic exposures to moisture that lasts less than a few days (i.e., normal rain and drain) are not significant. Significant voltage exposure is defined as being subjected to system voltage for more than twenty-five percent of the time.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-14The program includes inspections for water accumulation in manholes at least once every two years.This program will be implemented prior to the period of extended operation. This new program will be implemented consistent with the corresponding program described in NUREG-1801 Section XI.E3, Inaccessible Medium-Voltage Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, prior to the period of extended operation.A.1.1.26Non-EQ Instrumentation Circuits Test Review ProgramThe Non-EQ Instrumentation Circuits Test Review Program is a new program that inspects the applicable cables in the following systems or sub-systems.

• neutron monitoring system intermediate range monitors

• neutron monitoring system local power range monitors

• neutron monitoring system average power range monitors

• reactor building ventilation exhaust radiation monitors

• main steam line radiation monitorsThe Non-EQ Instrumentation Circuits Test Review Program assures the intended functions of sensitive, high-voltage, low-signal cables exposed to adverse localized equipment environments caused by heat, radiation and moisture (i.e., neutron flux monitoring instrumentation, reactor building ventilation exhaust radiation monitoring, and main steam line radiation monitoring) can be maintained consistent with the current licensing basis through the period of extended operation. Most sensitive instrumentation circuit cables and connections are included in the instrumentation loop calibration at the normal calibration frequency, which provides sufficient indication of the need for corrective actions based on acceptance criteria related to instrumentation loop performance. The review of calibration results will be performed once every ten years, with the first review occurring before the period of extended operation.For sensitive instrumentation circuit cables that are disconnected during instrument calibrations, testing using a proven method for detecting deterioration for the insulation system (such as insulation resistance tests or time domain reflectometry) will occur at least once every ten years, with the first test occurring before the period of extended operation. This program will consider the technical information and guidance provided by the industry. This program will be implemented prior to the period of extended operation. This new program will be implemented consistent with the corresponding program described in NUREG-1801 Section XI.E2, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Used in Instrumentation Circuits.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-15A.1.1.27Non-EQ Insulated Cables and Connections ProgramThe Non-EQ Insulated Cables and Connections Program is a new program that assures the intended functions of insulated cables and connections exposed to adverse localized environments caused by heat, radiation and moisture can be maintained consistent with the current licensing basis through the period of extended operation. An adverse localized environment is significantly more severe than the specified service condition for the insulated cable or connection.A representative sample of accessible insulated cables and connections within the scope of license renewal will be visually inspected for cable and connection jacket surface anomalies such as embrittlement, discoloration, cracking or surface contamination. The program sample consists of all accessible cables and connections in localized adverse environments.This program will be implemented prior to the period of extended operation. This new program will be implemented consistent with the corresponding program described in NUREG-1801 Section XI.E1, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, prior to the period of extended operation.A.1.1.28Oil Analysis ProgramThe Oil Analysis Program is an existing program that maintains oil systems free of contaminants (primarily water and particulates) thereby preserving an environment that is not conducive to loss of material, cracking, or fouling. Activities include sampling and analysis of lubricating oil for detrimental contaminants, water, and particulates.Sampling frequencies are based on vendor recommendations, accessibility during plant operation, equipment importance to plant operation, and previous test results.The One-Time Inspection Program [Section A.1.1.29] utilizes inspections or non-destructive evaluations of representative samples to verify that the Oil Analysis Program has been effective at managing aging effects.The Oil Analysis Program will be enhanced as follows.

• Include viscosity, neutralization number, and flash point determination of oil samples from components that do not have regular oil changes, along with analytical ferrography and elemental analysis for the identification of wear particles.

• Include screening for particulate and water content for oil replaced periodically.

• Formalize preliminary oil screening for water and particulates and laboratory analyses, including defined acceptance criteria for all components included in the scope of the Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-16program. The program will specify corrective actions in the event acceptance criteria are not met.Enhancements will be implemented prior to the period of extended operation.A.1.1.29One-Time Inspection ProgramThe One-Time Inspection Program is a new program that will include measures to verify effectiveness of an aging management program (AMP) and confirm the insignificance of an aging effect. For structures and components that rely on an AMP, this program will verify effectiveness of the AMP by confirming that unacceptable degradation is not occurring and the intended function of a component will be maintained during the period of extended operation. One-time inspections may be needed to address concerns for potentially long incubation periods for certain aging effects on structures and components. There are cases where either (a) an aging effect is not expected to occur but there is insufficient data to completely rule it out, or (b) an aging effect is expected to progress very slowly. For these cases, there will be confirmation that either (a) the aging effect is indeed not occurring, or (b) the aging effect is occurring very slowly as not to affect the component or structure intended function. A one-time inspection of the subject component or structure is appropriate for this verification. The inspections will be nondestructive examinations (including visual, ultrasonic, or surface techniques). The inspection will be performed within the ten years prior to the period of extended operation. The elements of the One-Time Inspection Program include (a) determination of the sample size based on an assessment of materials of fabrication, environment, plausible aging effects, and operating experience; (b) identification of the inspection locations in the system or component based on the aging effect; (c) determination of the examination technique, including acceptance criteria that would be effective in managing the aging effect for which the component is examined; and (d) evaluation of the need for follow-up examinations to monitor the progression of any aging degradation. A one-time inspection activity is used to verify the effectiveness of the Diesel Fuel Monitoring Program by confirming that unacceptable loss of material is not occurring on components within systems managed by the Diesel Fuel Monitoring Program [Section A.1.1.12

].A one-time inspection activity is used to verify the effectiveness of the Oil Analysis Program by confirming that unacceptable cracking, loss of material, and fouling is not occurring on components within systems managed by the Oil Analysis Program [Section A.1.1.28

].A one-time inspection activity is used to verify the effectiveness of the three water chemistry control programs by confirming that unacceptable cracking, loss of material, and fouling is not occurring on components within systems managed by water chemistry control programs [Sections A.1.1.38 , A.1.1.39, and A.1.1.40].

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-17One-time inspection activities on the following are used to confirm that loss of material, cracking, and reduction of fracture toughness, as applicable, are not occurring or are so insignificant that an aging management program is not warranted.

• main steam line flow elements

• reactor recirculation flow elements

• internal surfaces of stainless steel components in the standby gas treatment system containing raw water (drain water)

• internal surfaces of stainless steel tubing in the circulating water system containing raw water (river water)

• internal surfaces of stainless steel tubing and components in the off gas system containing raw water (drain water)

• internal surfaces of stainless steel components in the radwaste system containing raw water (drain water)

• Internal surfaces of stainless steel tubing and components in the service air system exposed to condensationThe program provides for increasing inspection sample size and locations in the event that aging effects are detected. Unacceptable inspection findings are evaluated in accordance with the corrective action process to determine the need for subsequent (including periodic) inspections and for monitoring and trending the results.For specific system components where significant aging effects are not expected, one-time inspection activities are used to confirm that loss of material, cracking, and reduction of fracture toughness, as applicable, are not occurring or are so insignificant that an aging management program is not warranted. When evidence of an aging effect is revealed by a one-time inspection, routine evaluation of the inspection results will identify appropriate corrective actions.This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M32, One-Time Inspection.A.1.1.30One-Time Inspection - Small-Bore Piping ProgramThe One-Time Inspection - Small-Bore Piping Program is a new program applicable to small-bore American Society of Mechanical Engineers (ASME) Code Class 1 piping less than 4 inches nominal pipe size (NPS 4"), which includes pipe, fittings, and branch connections. The ASME Code does not require volumetric examination of Class 1 small-bore piping. The CNS One-Time Inspection of ASME Code Class 1 Small-Bore Piping Program will manage cracking through the use of volumetric examinations.The program will include a sample selected based on susceptibility, inspectability, dose considerations, operating experience, and limiting locations of the total population of ASME Code Class 1 small-bore piping locations.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-18When evidence of an aging effect is revealed by a one-time inspection, evaluation of the inspection results will identify appropriate corrective actions.The inspection will be performed within the ten years prior to the period of extended operation. This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M35, One-Time Inspection of ASME Code Class 1 Small-Bore Piping.A.1.1.31Periodic Surveillance and Preventive Maintenance ProgramThe Periodic Surveillance and Preventive Maintenance Program is an existing program that includes periodic inspections and tests that manage aging effects not managed by other aging management programs. In addition to specific activities in the plant's preventive maintenance program and surveillance program, the Periodic Surveillance and Preventive Maintenance Program includes enhancements to add new activities. The preventive maintenance and surveillance testing activities are generally implemented through repetitive tasks or routine monitoring of plant operations. The program is credited with managing loss of material from external surfaces for situations in which external and internal material and environment combinations are the same such that internal surface condition is representative of external surface condition.Surveillance testing and periodic inspections using visual or other non-destructive examination techniques verify that the following components are capable of performing their intended function.*reactor building monorails, railroad airlock doors, reactor building crane, rails and girders, and refueling bridge equipment assembly

• elastomer seals for railroad airlock doors

• SLC system accumulator shells

• HPCI system turbine lube oil heat exchanger tubes

• ADS piping and T-quenchers in waterline region of the suppression chamber

• RCIC system vacuum pump discharge piping, piping elements, and components

• RCIC system turbine lube oil heat exchanger tubes

• SGT system components

• SGT system fan inlet flexible connections

• plant drain system components

• DG system exhaust gas components

• DG system intercooler tubes and fins

• DG system service air components

• HVAC system flexible duct connections

• HVAC system portable blower fan housings and flexible trunks kept in storage that may be used for ventilation

• HVAC system fan coil unit tubes, fins and drip pan

• PC system equipment and floor drain components Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-19

• piping, piping components, and piping elements in the circulating water system, nonradioactive floor drain system, heating and ventilation system, off gas system, potable water system, radwaste system, diesel generator starting air system, and service air system*service air primary containment penetration X-21

• nitrogen system vaporizer tank and vaporizer coilThe Periodic Surveillance and Preventive Maintenance Program will be enhanced as follows.

• Enhance as necessary to assure that the effects of aging will be managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.

• For each activity that refers to a representative sample, a representative sample will be selected for each unique material and environment combination. The sample size will be determined in accordance with Chapter 4 of EPRI 107514, Age Related Degradation Inspection Method and Demonstration, which outlines a method to determine the number of inspections required for 90% confidence that 90% of the population does not experience degradation (90/90).Enhancements will be implemented prior to the period of extended operation.A.1.1.32Reactor Head Closure Studs ProgramThe Reactor Head Closure Studs Program is an existing program that includes inservice inspection (ISI) in conformance with the requirements of ASME Section XI, Subsection IWB, and preventive measures (e.g., rust inhibitors, stable lubricants, appropriate materials) to mitigate cracking and loss of material of reactor head closure studs, nuts, washers, and bushings.A.1.1.33Reactor Vessel Surveillance ProgramThe Reactor Vessel Surveillance Program is an existing program that manages reduction in fracture toughness of reactor vessel beltline materials to assure that the pressure boundary function of the reactor pressure vessel is maintained through the period of extended operation.CNS has received NRC approval to use the BWR vessel and internals project (BWRVIP) Integrated Surveillance Program (ISP). The Reactor Vessel Surveillance Program monitors changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. As BWRVIP-ISP capsule test reports become available for RPV materials representative of CNS, the actual shift in the reference temperature for nil-ductility transition of the vessel material may be updated. In accordance with 10 CFR 50 Appendices G and H, CNS reviews relevant test reports to assure compliance with fracture toughness requirements and P-T limits.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-20BWRVIP-116, "BWR Vessel and Internals Project Integrated Surveillance Program (ISP) Implementation for License Renewal," describes the design and implementation of the ISP during the period of extended operation. BWRVIP-116 identifies additional capsules of the Supplemental Surveillance Program (SSP), their withdrawal schedule, and contingencies to ensure that the requirements of 10 CFR 50 Appendix H are met through the period of extended operation.The Reactor Vessel Surveillance Maintenance Program will be enhanced as follows.

• If the CNS standby capsule is removed from the reactor vessel without the intent to test it, the capsule will be stored in a manner which maintains it in a condition which would permit its future use, including during the period of extended operation, if necessary.

• Ensure that the additional requirements specified in the final NRC safety evaluation for BWRVIP-116 will be addressed before the period of extended operation.Enhancements will be implemented prior to the period of extended operation.A.1.1.34Selective Leaching ProgramThe Selective Leaching Program is a new program that will ensure the integrity of components made of cast iron, bronze, brass, and other alloys exposed to condensation, raw water, steam, treated water, and soil (groundwater) that may lead to selective leaching. The program will include a one-time visual inspection, hardness measurement (where feasible based on form and configuration), or other industry accepted mechanical inspection techniques of selected components that may be susceptible to selective leaching to determine whether loss of material due to selective leaching is occurring, and whether the process will affect the ability of the components to perform their intended function through the period of extended operation.This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M33, Selective Leaching of Materials, prior to the period of extended operation.A.1.1.35Service Water Integrity ProgramThe Service Water Integrity Program is an existing program that relies on implementation of the recommendations of GL 89-13 to ensure that the effects of aging on the service water (SW) system will be managed through the period of extended operation. The program includes component inspections for cracking, erosion, corrosion, wear, and blockage and performance monitoring to verify the heat transfer capability of the safety-related heat exchangers cooled by SW. Periodic cleaning and flushing of redundant or infrequently used loops are the methods used to control or prevent fouling within the heat exchangers and loss of material in SW components.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-21A.1.1.36Structures Monitoring ProgramThe Structures Monitoring Program is an existing program that performs inspections in accordance with 10 CFR 50.65 (Maintenance Rule) as addressed in Regulatory Guide 1.160 and NUMARC 93-01. Periodic inspections are used to monitor the condition of structures and structural commodities to ensure there is no loss of intended function. Since protective coatings are not relied upon to manage the effects of aging for structures included in the Structures Monitoring Program, the program does not address protective coating monitoring and maintenance.The Structures Monitoring Program will be enhanced as follows.

• Clarify that the following structures are included in the program.biological shield wallcontrol room ceiling support systemcrane rails and girdersCRD shootout steeldiesel fuel tank hatch coverdiesel fuel tank retaining wall and slabdrywell fill slabdrywell shell protection panels and jet deflectorsdrywell stabilizer supportsfoundations (buildings)guide wallmanholes and duct banksmonorailsnew fuel storage vaultoffice building (or administration building)oil tank bunker crushed rock fillpump baffle platesreactor building loop seal drain capsreactor building railroad airlock doorsreactor building sump structurereactor cavity floor and wallsreactor cavity linerreactor pedestalsacrificial shield wall (steel portion)sacrificial shield wall lateral supportsservice water pipe slabshield plugsspent fuel pool floor and wallssteam tunnel Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-22sumps and sump linerstransformer yard and switchyard support structures and foundationstransmission towers (galvanized), wooden utility towers, wooden utility poles, and foundationstraveling screen casing and associated framing

• Clarify that, in addition to structural steel and concrete, the following commodities are inspected for each structure as applicable.anchor boltsanchorage/embedmentsbase platesbattery racksbeams, columns, floor slabs, and walls (below grade)blowout panels (including east end of steam tunnel)cable trays and supportscomponent and piping supportsconduits and conduit supportselectrical and instrument panels and enclosuresequipment pads and foundationsexterior wallsflood curbsflood, pressure and specialty doorsflood retention materials (spare parts)HVAC duct supportsinstrument line supportsinstrument racks, frames, and tubing traysmanways, hatches, manhole covers, and hatch coversmissile shieldspenetration sealant (flood, radiation)penetration sleeves and sealant (mechanical/electrical not penetrating PC boundary)pipe whip restraintsseals and gaskets (doors, manways and hatches)stairs and handrails, platforms, grating, decking, and ladderssupport pedestalsvents and louvers

• Inspect inaccessible concrete areas that are submerged or below grade which may become exposed due to excavation, construction or other activities. CNS will also inspect inaccessible concrete areas when observed conditions in accessible areas exposed to the same environment indicate that significant concrete degradation is occurring.

• Inspect elastomers (seals, gaskets, and roof elastomers) to identify cracking and change in material properties.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-23

• Perform an engineering evaluation of groundwater samples to assess aggressiveness of groundwater to concrete on a periodic basis (at least once every five years). CNS will obtain samples from a well that is representative of the groundwater surrounding below-grade site structures. Samples will be monitored for sulfates, pH and chlorides.

• Perform visual structural examinations of wood to identify loss of material and change in material properties.

• Perform visual structural monitoring of the oil tank bunker crushed rock fill to identify loss of form.*Clarify that structures with conditions classified as "acceptable with deficiencies" or "unacceptable" shall be entered into the corrective action program.Enhancements will be implemented prior to the period of extended operation.A.1.1.37Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel ProgramThe Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program is a new program that will as sure reduction of fracture toughness due to thermal aging and reduction of fracture toughness due to radiation embrittlement will not result in loss of intended function. This program will evaluate CASS components in the reactor vessel internals and require non-destructive examinations as appropriate. This program will supplement reactor vessel internals inspections required by the BWR Vessel Internals Program [Section A.1.1.9] and the Inservice Inspection - ISI Program [Section A.1.1.19

] to manage the effects of loss of fracture toughness due to thermal aging and neutron embrittlement of cast austenitic stainless steel (CASS) components. This aging management program includes(a)identification of susceptible components determined to be limiting from the standpoint of thermal aging susceptibility (i.e., ferrite and molybdenum contents, casting process, and operating temperature) and/or neutron irradiation embrittlement (neutron fluence), and (b)for each "potentially susceptible" component, aging management is accomplished through either a supplemental examination of the affected component during the period of extended operation, or a component-specific evaluation to determine its susceptibility to reduction of fracture toughness.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-24This new program will be implemented consistent with the corresponding program described in NUREG-1801,Section XI.M13, Thermal Aging and Neutron Irradiat ion Embrittlement of Cast Austenitic Stainless Steel (CASS) Program, prior to the period of extended operation.A.1.1.38Water Chemistry Control - Auxiliary Systems ProgramThe Water Chemistry Control - Auxiliary Systems Program is an existing program that manages loss of material and cracking for components exposed to treated water and steam.Program activities include sampling and analysis of water in auxiliary condensate drain system components, auxiliary steam system components, and heating and ventilation system components to minimize component exposure to aggressive environments.The One-Time Inspection Program [Section A.1.1.29] utilizes inspections or non-destructive evaluations of representative samples to verify that the Water Chemistry Control - Auxiliary Systems Program has been effective at managing aging effects. A.1.1.39Water Chemistry Control - BWR ProgramThe Water Chemistry Control - BWR Program is an existing program that manages aging effects caused by corrosion and cracking mechanisms. The program relies on monitoring and control of water chemistry based on EPRI Report 1008192 (BWRVIP-130). BWRVIP-130 has three sets of guidelines: one for primary water, one for condensate and feedwater, and one for control rod drive (CRD) mechanism cooling water. EPRI guidelines in BWRVIP-130 also include recommendations for controlling water chemistry in the torus/pressure suppression chamber, condensate storage tank, demineralized water storage tanks, and spent fuel pool.The Water Chemistry Control - BWR Program optimizes the primary water chemistry to minimize the potential for loss of material and cracking. This is accomplished by limiting the levels of contaminants in the reactor coolant system that could cause loss of material and cracking. Additionally, CNS has instituted hydrogen water chemistry and noble metal chemical addition to limit the potential for IGSCC through the reduction of dissolved oxygen in the treated water.The One-Time Inspection Program [Section A.1.1.29] utilizes inspections or non-destructive evaluations of representative samples to verify that the Water Chemistry Control - BWR Program has been effective at managing aging effects. A.1.1.40Water Chemistry Control - Closed Cooling Water ProgramThe Water Chemistry Control - Closed Cooling Water Program is an existing program that includes preventive measures that manage loss of material, cracking, and fouling for components in closed cooling water systems: diesel generator jacket water (DGJW) system, reactor equipment cooling (REC) system, and turbine equipment cooling (TEC) system. These Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-25chemistry activities provide for monitoring and controlling closed cooling water chemistry using CNS procedures and processes based on EPRI guidance for closed cooling water chemistry.The One-Time Inspection Program [Section A.1.1.29] utilizes inspections or non-destructive evaluations of representative samples to verify that the Water Chemistry Control - Closed Cooling Water Program has been effective at managing aging effects.A.1.2Evaluation of Time-Limited Aging AnalysesIn accordance with 10 CFR 54.21(c), an application for a renewed license requires an evaluation of time-limited aging analyses for the period of extended operation. The following time-limited aging analyses have been identified and evaluated to meet this requirement.A.1.2.1Reactor Vessel Neutron EmbrittlementThe reactor vessel neutron embrittlement time-limited aging analyses, including consideration for the measurement uncertainty recapture (MUR) power uprate for cycle 25 and beyond, either have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii) or will be managed for the period of extended operation in accordance with 10CFR 54.24(c)(1)(iii) as summarized below.Based on the plant operating history and assuming 100 percent capacity factor through the period of extended operation, CNS will not surpass 50 EFPY. However, 54 EFPY (90percent capacity factor times 60 years) is conservatively used as the end of the period of extended operation to evaluate reactor vessel neutron embrittlement time-limited aging analyses.A.1.2.1.1Reactor Vessel FluenceCalculated fluence is based on a time-limited assumption defined by the operating term. Therefore, analyses that evaluate reactor vessel neutron embrittlement based on calculated fluence are time-limited aging analyses.The high energy (> 1 MeV) neutron fluence for the welds and shells of the reactor pressure vessel beltline region was determined using the Radiation Analysis Modeling Application (RAMA) fluence method which adheres to the guidance prescribed in Regulatory Guide 1.190.A.1.2.1.2Adjusted Reference TemperatureThe change in reference temperature of nil-ductility transition (RT NDT) and adjusted reference temperature (ART) values were projected to 54 EFPY using the methods described in Regulatory Guide 1.99 Revision 2. Credible surveillance data and the integrated surveillance program (ISP) were used to determine chemistry factors and best-estimate chemistry values for the lower intermediate shell plates. All projected values for ART are below the 200°F suggested in Section 3 of Regulatory Guide 1.99 as an acceptable value of ART for the end of the period of extended operation.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-26The time-limited aging analysis for adjusted reference temperature has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).A.1.2.1.3Pressure-Temperature LimitsAppendix G of 10 CFR 50 requires that the reactor vessel remain within established pressure-temperature (P-T) limits during boltup, hydro-test, pressure tests, normal operation, and anticipated operational occurrences. These limits are calculated using materials and fluence data, including data obtained through the Reactor Vessel Surveillance Program.The P-T limit curves will continue to be updated, as required by Appendix G of 10 CFR Part 50, assuring that limits remain valid through the period of extended operation.The aging effects associated with the reactor vessel pressure-temperature limits will be managed for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).A.1.2.1.4Upper-Shelf EnergyThe predictions for percent drop in Charpy upper shelf energy (C vUSE) values were projected to 54 EFPY using projected beltline fluence values, chemistry and surveillance data, and un-irradiated C vUSE information in accordance with Regulatory Guide 1.99. All projected C vUSE values for 54 EFPY remain above the 50 ft-lb minimum acceptable value specified in Appendix G of 10 CFR 50.The time-limited aging analyses for upper shelf energy have been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).A.1.2.1.5Reactor Vessel Circumferential Weld Inspection ReliefRelief from reactor vessel circumferential weld examination requirements during the fourth ten-year ISI interval for CNS was requested in 2007. The relief request is based on BWRVIP-05 and the associated NRC safety evaluation report (SER), and its supplement (References A.1-4 and A.1-5). The relief request is applicable for the remaining portion of the current operating license.The CNS reactor pressure vessel circumferential weld parameters at 54 EFPY will remain within the NRC's (64 EFPY) bounding CE parameters from the BWRVIP-05 SER. The fact that the values projected to the end of the period of extended operation are less than the 64 EFPY value provided by the NRC leads to the conclusion that the CNS RPV cond itional failure probability is less than the conditional failure probability of the NRC analysis. As such, the conditional probability of failure for circumferential welds remains below that determined during the NRC's final safety evaluation of BWRVIP-05.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-27Axial WeldsA basic assumption in calculating the failure probability of the circumferential welds is the failure probability of the axial welds.The CNS reactor vessel limiting axial weld parameters were compared to those used in the NRC analysis in BWRVIP-05 (Reference A.1-4). The projected 54 EFPY CNS mean ART for axial welds is less than the value shown in the NRC SER for BWRVIP-74 (Reference A.1-6

).The time-limited aging analysis for reactor vessel circumferential weld inspection relief has been projected to the end of the period of extended operation in accordance with 10 CFR 54.21(c)(1)(ii).A.1.2.2Metal FatigueA.1.2.2.1Class 1 Metal FatigueFatigue evaluations were performed in the design of the CNS Class 1 components in accordance with their design requirements. ASME Section III fatigue evaluations are contained in analyses and stress reports, and because they may be based on a number of transient cycles assumed for a 40-year operating term, these evaluations are considered time-limited aging analyses.Design cyclic loadings and thermal conditions for the Class 1 components are defined by the applicable design specifications for each component. The original design specifications provided a set of transients that were used in the design of the components and are included as part of each component analysis or stress report.The Fatigue Monitoring Program tracks and evaluates the cycles and requires corrective actions if limits are approached. The Fatigue Monitoring Program ensures that the numbers of transient cycles experienced by the plant remain within the analyzed numbers of cycles, and hence the component CUFs remain below the code allowable value of 1.0.Reactor VesselThe design code for the reactor vessel is specified in Section IV-2.5.1 of the USAR. Fatigue evaluations for the reactor vessel were performed as part of the vessel design. The fatigue analyses of the reactor vessel are considered time-limited aging anal yses because they are based on numbers of design cycles expected to occur in 40 years of operation.The actual numbers of transient cycles remain within analyzed values used for reactor vessel fatigue analyses. CNS will monitor these transient cycles using the Fatigue Monitoring Program and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the reactor vessel in accordance with 10 CFR 54.21(c)(1)(iii).

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-28Reactor Vessel Feedwater NozzleAs discussed in USAR Section IV-2.5.1.1, the feedwater nozzles were modified in 1980 to remove stainless steel cladding in order to reduce thermal stresses and crack initiation.In 2007, Cooper submitted a Technical Specification change request (Reference A.1-3) that included a re-evaluation of the feedwater nozzle fatigue including MUR. The projected CUF for the nozzle/shell junction, including system cycling and rapid cycling, slightly exceeds 1.0.The feedwater rapid cycling is analyzed based on years of operation, and the number of analyzed years (40) will be exceeded during the period of extended operation. Consequently, the feedwater nozzle CUF cannot be successfully projected for the period of extended operation.

The feedwater nozzle is one of the locations identified by NUREG-6260 for assessment of the effects of the reactor water environment on fatigue. See Section A.1.2.2.3 for a discussion of the environmentally assisted fatigue analysis of the feedwater nozzles and how CNS will manage the aging effect due to fatigue on the feedwater nozzles. CNS will continue to manage fatigue due to rapid cycling using the BWR Feedwater Nozzle Program. As such, the effects of fatigue on the feedwater nozzles will be managed for the period of extended operation in accordance with 10CFR 54.21(c)(1)(iii).Reactor Vessel InternalsThe CNS reactor pressure vessel internals are not Class 1 pressure boundary components. As such, no plant specific fatigue analysis of the entire reactor vessel internals was performed. Fatigue analyses of specific internals piece parts have been performed over the years; however, the only time-limited aging analyses associated with fatigue of the reactor vessel internals at CNS are the core plate plugs addressed below in Section A.1.2.5. A qualitative review of the internals was performed for the MUR power uprate, and it was concluded that the governing stresses for all RPV internal components in the MUR condition remain bounded by the existing values. The shroud support and brackets welded to the vessel are considered part of the vessel and had CUFs calculated in the vessel stress report.Class 1 PipingOriginal piping was designed in accordance with B31.1, "Power Piping." Other Class I S piping is designed to meet the supplementary requirements included in Section A-3.1 of the USAR.Repairs, replacements and modifications are generally performed in accordance with the original code requirements. As permitted by ASME Section XI, later editions of the code or ASME III have been used for some modifications at CNS. To the extent practical, portions of the Class I N piping and nozzle safe ends subject to intergranular stress corrosion cracking (IGSCC) have been replaced with resistant material. The design code for the replaced piping is ASME Section III, 1983 Edition per Section A-3.1 of the USAR.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-29In the B31.1 code, fatigue is addressed by using stress range reduction factors to reduce stress allowable. Components with less than 7000 equivalent full temperature cycles are limited to the calculated stress allowable without reduction per B31.1. Components that exceed 7000 equivalent full temperature cycles have allowable stresses reduced through the application of stress range reduction factors. Since the reactor coolant pressure boundary will not exceed 7000 full temperature cycles in 60 years of operation, the existing stress analyses remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).To the extent practical, portions of the reactor water cleanup (RWCU) and the reactor pressure vessel drain line piping subject to IGSCC have been replaced with IGSCC-resistant material. The design code for the replaced piping is B31.1, as discussed in Section A-3.1 of the USAR.Specific to the ASME Section III piping, a review of CNS fatigue analyses found CUFs calculated for reactor recirculation (RR), residual heat removal (RHR), RWCU, main steam (MS), core spray (CS), reactor feedwater (RF), and reactor pressure vessel level sensing lines.For the ASME Section III piping, CNS will monitor the cycles actually incurred compared to the cycles analyzed using the Fatigue Monitoring Program and assure that action is taken if any of the actual cycles approach their analyzed numbers. As such, the Fatigue Monitoring Program will manage the effects of aging due to fatigue on the ASME Section III piping in accordance with 10 CFR 54.21(c)(1)(iii).A.1.2.2.2Non-Class 1 Metal FatigueThe design of ASME III Code Class 2 and 3 piping systems incorporates the Code stress reduction factor for determining acceptability of piping design with respect to thermal stresses. In general, 7000 thermal cycles are assumed, allowing a stress reduction factor of 1.0 in the stress analyses. CNS evaluated the validity of this assumption for 60 years of plant operation. The results of this evaluation indicate that the 7000 thermal cycle assumption will not be exceeded for 60 years of operation. Therefore, the pipe stress calculations remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).Non-Class 1 components, other than piping system components, required fatigue analyses only if they were built to ASME Section III, NC-3200 or ASME Section VIII, Division 2. CNS has no non-Class 1 components built to these codes and therefore has no associated time-limited aging analyses for components other than piping system components.A.1.2.2.3Effects of Reactor Water Environment on Fatigue Life NUREG/CR-6260 (Reference A.1-9) addresses the application of environmental factors to fatigue analyses (CUFs) and in Section 5.7 identifies the locations most sensitive to environmental effects for CNS vintage General Electric plants. These locations are relevant to

CNS.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-30(1)Reactor vessel shell and lower head(2)Reactor vessel feedwater nozzle(3)Reactor recirculation piping (including inlet and outlet nozzles)(4)Core spray line reactor vessel nozzles and associated Class 1 piping (5)Residual heat removal return line Class 1 piping(6)Feedwater line Class 1 pipingCNS evaluated these six locations using environmentally assisted fatigue correction factors (F en).Based on the analysis, only the feedwater nozzle, core spray nozzle, and RHR piping transition piece have environmentally adjusted CUFs greater than 1.0. These CUFs will exceed 1.0 at the beginning of the period of extended operation when environmental effects are considered. Due to the factor of safety included in the ASME code, a CUF of greater than 1.0 does not indicate that fatigue cracking is expected. However, there is a higher potential for fatigue cracking during the period of extended operation at locations having CUFs exceeding 1.0.The effects of environmentally assisted fatigue will be managed by the Fatigue Monitoring Program for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).A.1.2.3Environmental Qualification of Electrical ComponentsThe CNS Environmental Qualification (EQ) of Electric Components Program implements the requirements of 10 CFR 50.49 (as further defined by the Division of Operating Reactors Guidelines, NUREG-0588, and Reg. Guide 1.89). The program requires action before individual components exceed their qualified life. In accordance with 10 CFR 54.21(c)(1)(iii),

implementation of the EQ Program provides reasonable assurance that the effects of aging on components associated with EQ time-limited aging analyses will be adequately managed such that the intended functions can be maintained for the period of extended operation.A.1.2.4Fatigue of Primary Containment, Attached Piping, and Components Analyses of the CNS containment are included in the Plant Unique Analysis Report (PUAR)

(Reference A.1-7) and the generic Mark 1 containment report, MPR-751 (Reference A.1-8

).The CUF for the torus shell was determined to be 0.51 at the butt weld between the torus shell plates of unequal thickness at the torus equator. However, the initial analysis was redone in 1997, including the limiting ASME Code fatigue reduction factor of 5 for the entire shell. This new analysis yielded a CUF of 0.947. Rather than projecting this analysis, CNS will manage the Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-31aging effects due to fatigue of the torus shell using cycle-based fatigue monitoring. Thus the Fatigue Monitoring Program will manage the aging effects due to fatigue on the torus shell for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(iii).The fatigue analysis of the S/RV discharge line piping is bounded by MPR-751, the GE Mark 1 containment program (Reference A.1-8). MPR-751 was prepared to bound all BWR plants which utilize the Mark I containment design. The analysis concluded that for all plants and piping systems considered, in all cases the fatigue usage factors for an assumed 40-year plant life was less than 0.5. In a worst-case scenario, extending plant life by an additional 20 years would produce usage factors below 0.75. Since this is less than 1.0, the fatigue criteria are satisfied.For torus attached piping (internal and external to the torus), the results of the generic GE Mark I containment program (based on 40 years of operation) were that the torus attached piping would have cumulative usage factors of less than 0.5. In particular, the locations reported for CNS were all less than 0.3. Conservatively multiplying the CUFs by 1.5 demonstrates that for 60 years of operation, the torus attached piping for CNS would have CUFs below 0.75.Therefore, the analysis for the S/RV discharge piping and other attached piping has been projected to the end of the period of extended operation in accordance with 10 CFR 50.21(c)(ii).For Type 1 torus piping penetration assemblies, including expansion joint bellows, the cycle life is specified to be a minimum of 7000 cycles over a period of 40 years. The 7000 thermal cycle assumption is valid and bounding for 60 years of operation. Therefore, the torus piping penetration stress calculations remain valid for the period of extended operation in accordance with 10 CFR 54.21(c)(1)(i).A.1.2.5Core Plate PlugsThe 88 core plate bypass holes were plugged in the mid-1970s to eliminate in core instrument vibration that was causing damage to fuel channels. A stress analysis was performed on the plug considering normal operating conditions, pressure and thermal transients, and installation/removal operations. The results show acceptable stress levels in all plug components during normal operation and transients.The analysis of the fatigue life of these plugs is a time-limited aging analysis. The evaluation concluded that the predicted core plate plug life for both spring relaxation and for stress (fatigue) cracking was 32 EFPY and that the cumulative usage factor (CUF) at 32 EFPY is approximately 0.94, based on plotted data. The slope of the curve in this evaluation is such that it appears the CUF would exceed 1.0 prior to 54 EFPY. Cracking due to fatigue of the core plate plugs will be managed by the BWR Vessel Internals Program, with enhancement to include management of plugs in the core plate bypass holes, for the period of extended operation in accordance with 10CFR 54.21(c)(1)(iii).

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-32A.1.3ReferencesA.1-1[CNS License Renewal Application-later]A.1-2[NRC SER for CNS License Renewal-later]

A.1-3NPPD Letter NLS2007069 to USNRC, "License Amendment Request to Revise Technical Specifications - Appendix K Measurement Uncertainty Recapture Power Uprate Cooper Nuclear Station Docket 50-298, DPR-46," November 19, 2007.A.1-4Strosnider, J. (NRC) to C. Terry (BWRVIP Chairman), "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC NO. 92925)," letter dated July 28, 1998.A.1-5Strosnider, J. (NRC) to C. Terry (BWRVIP Chairman), Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC NO. MA3395)," letter dated March 7, 2000.A.1-6Wagoner, V., and T. Mulford (NRC) to All BWRVIP Committee Members, "Acceptance for Referencing of BWRVIP-74 in License Renewal Applications," letter dated October 31, 2001A.1-7Nebraska Public Power District, Cooper Nuclear Station, Plant Unique Analysis Report , Mark I Containment Program, revised February 26, 2007.A.1-8Technical Report MPR-751, Mark I Containment Program Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus Attached and SRV Piping Systems, November 1982.A.1-9NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, February 1995.

Cooper Nuclear StationLicense Renewal ApplicationTechnical InformationAppendix A Updated Safety Analysis Report SupplementPage A-33A.2LICENSE RENEWAL COMMITMENTSThe preliminary list of commitments made in the License Renewal Application for Cooper Nuclear Station has been provided in the letter transmitting the LRA to the NRC. The commitments reflect the contents of the LRA as submitted but are considered preliminary in that the specific wording of some commitments may change, and additional commitments may be

made, during the NRC review of the LRA. An y other actions discuss ed in the LRA should be considered intended or planned actions. These other actions are included for informational purposes but are not considered regulatory commitments.The final commitments as submitted by NPPD, and accepted by NRC, are expected to be confirmed in the NRC's Safety Evaluation Report (SER) for the renewed operating licenses. These final commitments, as confirmed in the SER, will become effective upon NRC issuance of the renewed operating license.