ML15160A194

NYS000497 - Dacimo, Fred, Entergy, Letter to Document Control Desk, Usnrc, Reply to Request for Additional Information Regarding the License Renewal Application, NL-12-089 (June 14, 2012) (Exhibit NRC000153) (Exhibit ENT000554) (ML12167A008

ML15160A194
Person / Time
Site:	Indian Point
Issue date:	06/14/2012
From:	State of NY, Office of the Attorney General
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 27908, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15160A194 (41)

Text

En tergyEnterqy Nuclear NortheastIndian Point Energy Center450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249Tel (914) 254-2055Fred DacimoVice PresidentOperations License RenewalNL-12-089June 14, 2012U.S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

REFERENCE:

Reply to Request for Additional Information Regardingthe License Renewal ApplicationIndian Point Nuclear Generating Unit Nos. 2 & 3Docket Nos. 50-247 and 50-286License Nos. DPR-26 and DPR-641. NRC Letter, "Request for Additional Information for the Review of theIndian Point Nuclear Generating Unit Nos. 2 and 3, License RenewalApplication," dated May 15, 2012

Dear Sir or Madam:

Entergy Nuclear Operations, Inc is providing, in Attachment 1, a reply to the additionalinformation requested in the referenced letter pertaining to NRC review of the License RenewalApplication (LRA) for Indian Point 2 and Indian Point 3. The reply provided in this transmittaladdresses questions on LRA Amendment No. 9 and the Reactor Vessel Internals (RVI)Program.As an initial matter, with regard to the RAls on the RVI Program, Entergy notes that Indian Pointis in a unique position with respect to the timing and implementation of the generic industryguidance for reactor vessel internals aging management (MRP-227-A). The Electric PowerResearch Institute (EPRI) just issued the NRC-approved version of MRP-227-A in January ofthis year, and the industry is working, through EPRI and the Pressurized Water ReactorOwners' Group (PWROG), to develop guidance on the required plant-specific evaluations forsubmittal to the NRC, including evaluations referenced in the RAls. As a result of Indian Point'sunique position, however, Entergy must prepare the requested evaluations in advance of thisguidance which will require additional time beyond the requested 30-day response period.Nevertheless, in this letter Entergy provides responses to RAls 1-5, 8, and 12. Entergy willdevelop the required evaluations and submit responses to the remaining RAls by 09/28/2012.Attachment 2 provides the latest list of regulatory commitments including the commitment madein response to RAI 11 in this letter.If you have any questions, or require additional information, please contact Mr. Robert Walpoleat 914-254-6710.

Docket Nos. 50-247 & 50-286NL-12-089Page 2 of 2I declare under penalty of perjury that the foregoing is true and correct. Executed onSincerely,FRD/rw

Attachment:

1. Reply to NRC Request for Additional Information Regarding the LicenseRenewal Application2. License Renewal Application IPEC List of Regulatory CommitmentsRevision 18.cc: Mr. William Dean, Regional Administrator, NRC Region IMr. Sherwin E. Turk, NRC Office of General Counsel, Special CounselMr. Dave Wrona, NRC Branch Chief, Engineering Review Branch IMr. Robert F. Kuntz, NRC Sr. Project Manager, Division of License RenewalMr. Douglas Pickett, NRR Senior Project ManagerMs. Bridget Frymire, New York State Department of Public ServiceNRC Resident Inspector's OfficeMr. Francis J. Murray, Jr., President and CEO NYSERDA ATTACHMENT 1 TO NL-12-089REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATIONREGARDING THELICENSE RENEWAL APPLICATIONENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3DOCKET NOS. 50-247 AND 50-286 NL-12-089Attachment 1Page 1 of 19INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3LICENSE RENEWAL APPLICATION (LRA)REQUESTS FOR ADDITIONAL INFORMATION (RAI)NRC RAI's Related to License Renewal Application Amendment No. 9 (Ref. 1)RAIlOn page 3 of license renewal application (LRA) Amendment 9 (Ref. 1), it is stated that Table2.3.1-2-1P2 and Table 2.3.1-2-1P3 list the mechanical components subject to aging managementreview and component intended functions for the reactor vessel internals. However, Table2.3.1-2-1P3 (the table for Indian Point Nuclear Generating Unit No. 3 (IP3)), is missing, and thetable for Indian Point Nuclear Generating Unit No. 2 (IP2) listing the components subject toaging management review is numbered Table 2.3.1-4-1P2. Provide Table 2.3.1-2-1P3 andcorrect the numbering of the table for IP2.Response to RAI 1Table 2.3.1.4-1P2 was numbered incorrectly in Amendment 9 and should have been identifiedas Table 2.3.1-2-1P2. Table 2.3.1-2-1P3 was inadvertently omitted from the Amendment 9submittal; however it would have been the same as Table 2.3.1-2-1P2. Tables 2.3.1-2-1P2 and2.3.1-2-1P3 are presented below as they should have appeared in Amendment 9.

NL-12-089Attachment 1Page 2 of 19Table 2.3.1-2-IP2Reactor Vessel InternalsComponents Subject to Aging Management ReviewComponent Type Intended FunctionLower Core Support Structure ..Core baffle/former assembly Structural support-boltsCore baffle/former assembly Structural support* plates Flow distributionShieldingCore barrel assembly Structural support-bolts and screwsCore barrel assembly Structural support* axial flexure plates (thermalshield flexures)Core barrel assembly Structural support* flangeCore barrel assembly Structural support* ring Flow distribution" shell" thermal shield ShieldingCore barrel assembly Structural support" lower core barrel flange weld" upper core barrel flange weldCore barrel assembly Flow distribution* outlet nozzlesLower internals assembly Structural support* clevis insert bolt* clevis insert* fuel alignment pin* lower core support plate columnsleeves* lower core support plate columnbolt* radial key NL-12-089Attachment 1Page 3 of 19Table 2.3.1-2-IP2Reactor Vessel InternalsComponents Subject to Aging Management ReviewComponent Type Intended FunctionLower internals assembly Flow distribution-intermediate diffuser plateLower internals assembly Structural support* lower core plate Flow distribution" lower core support castings" column cap" lower core support" secondary core supportUpper Core Support Structuie-OUpper InternalsAssemblyRCCA guide tube assembly Structural support* boltRCCA guide tube assembly Structural support* guide tube (including lowerflange welds)RCCA guide tube assembly Structural support* guide platesRCCA guide tube assembly Structural support-support pinCore plate alignment pin Structural supportHead / vessel alignment pin Structural supportHold-down spring Structural supportMixing devices Structural support" support column orifice base Flow distribution* support column mixerSupport column Structural supportUpper core plate, fuel alignment Structural supportpin Flow distribution NL-12-089Attachment 1Page 4 of 19Table 2.3.1-2-IP2Reactor Vessel InternalsComponents Subject to Aging Management ReviewComponent Type Intended FunctionUpper support plate, support Structural supportassembly (including ring)Upper support column bolt Structural supportIncore Instnrumentaiion Suport StructureBottom mounted instrumentation Structural supportcolumnFlux thimble guide tube Structural supportThermocouple conduit Structural support NL-12-089Attachment 1Page 5 of 19Table 2.3.1-2-IP3Reactor Vessel InternalsComponents Subject to Aging Management ReviewCmoetType IIntended FunctionLower Core Support StructureCore baffle/former assembly Structural support* boltsCore baffle/former assembly Structural support*plates Flow distributionShieldingCore barrel assembly Structural support-bolts and screwsCore baffrl aseombly Structural supporti floxure plates Flow distributions-hell*the.rmal' hhilldCore barrel assembly Structural support* axial flexure plates (thermalshield flexures)Core barrel assembly Structural support* flangeCore barrel assembly Structural supportSdog Flow distribution" shell" thermal shield ShieldingCore barrel assembly Structural support* lower core barrel flange weld" upper core barrel flange weldCore barrel assembly Flow distribution* outlet nozzles NL-12-089Attachment 1Page 6 of 19Table 2.3.1-2-IP3Reactor Vessel InternalsComponents Subject to Aging Management ReviewComponent TypeIntended FunctionLower internals assembly Structural support" clevis insert bolt* clevis insert* fuel alignment pin" lower core support plate columnbolt* lower core support plate columnsleeves* radial keyLower internals assembly Flow distribution* intermediate diffuser plateLower internals assembly Structural support" lower core plate Flow distribution" lower core support castings* column cap" lower core support" secondary core supportUpper C~ote Support Structure-Upper Internals AssemblyRCC'A" guid. tubo ac......ly Structura cu...p...r tRCCA quide tube assembly Structural support" boltRCCA guide tube assembly Structural support" quide tube (includinq lowerflanqe welds)RCCA guide tube assembly Structural support* guide platesRCCA guide tube assembly Structural supportg support pinCore plate alignment pin Structural supportHead / vessel alignment pin Structural support NL-12-089Attachment 1Page 7 of 19Table 2.3.1-2-IP3Reactor Vessel InternalsComponents Subject to Aging Management ReviewComponent Type Intended FunctionHold-down spring Structural supportMixing devices Structural support* support column orifice base Flow distribution* support column mixerSupport column Structural supportUpper core plate, fuel alignment Structural supportpin Flow distributionUpper support plate, support Structural supportassembly (including ring)Upper support column bolt Structural supportIncore'Istrumentation Support StructureBottom mounted instrumentation Structural supportcolumnFlux thimble guide tube Structural supportThermocouple conduit Structural support NL-12-089Attachment 1Page 8 of 19RAI 2LRA Sections 3.1.2.2.6, 3.1.2.2.9, 3.1.2.2.15, and 3.1.2.2.17, provided in LRA Amendment 9refer to MRP-227. For consistency with the revised LRA Section B.1.42 submitted by letterdated February 17, 2012, the staff requests that the applicant revise the LRA sections listedabove to update the reference to MRP-227-A.Response to RAI 2LRA Sections 3.1.2.2.6, 3.1.2.2.9, 3.1.2.2.15, and 3.1.2.2.17 are revised as shown below toupdate the reference to MRP-227-A. (underline -added)3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void SwellingLoss of fracture toughness due to neutron irradiation embrittlement and change indimensions (void swelling) in stainless steel and nickel alloy reactor vessel internalscomponents exposed to reactor coolant and neutron flux will be managed by the ReactorVessel Internals (RVI) Program. The RVI Program will implement the EPRI PressurizedWater Reactor Internals Inspection and Evaluation Guidelines, MRP-227-A. The RVIProgram will use nondestructive examinations (NDE) and other inspection methods tomanage aging effects for reactor vessel internals.3.1.2.2.9 Loss of Preload due to Stress RelaxationLoss of preload due to thermal stress relaxation (creep) would only be a concern in veryhigh temperature applications (> 7000F) as stated in the ASME Code,Section II, Part D,Table 4. No IPEC internals components operate at > 7000F. Therefore, loss of preload dueto thermal stress relaxation (creep) is not an applicable aging effect for the reactor vesselinternals components. However, irradiation-enhanced creep (irradiation creep) or irradiationenhanced stress relaxation (ISR) is an athermal process that depends on the neutronfluence and stress; and, on void swelling if present. Therefore, loss of preload of stainlesssteel and nickel alloy reactor vessel internals components will be managed by the ReactorVessel Internals (RVI) Program. The RVI Program will implement the EPRI PressurizedWater Reactor Internals Inspection and Evaluation Guidelines, MRP-227-A. The RVIProgram will use nondestructive examinations (NDE) and other inspection methods tomanage aging effects for reactor vessel internals.3.1.2.2.15 Changes in Dimensions due to Void SwellingChanges in dimensions due to void swelling in stainless steel and nickel alloy reactorinternal components exposed to reactor coolant will be managed by the Reactor VesselInternals (RVI) Program. The RVI Program will implement the EPRI Pressurized WaterReactor Internals Inspection and Evaluation Guidelines, MRP-227-A. The RVI Program willuse nondestructive examinations (NDE) and other inspection methods to manage agingeffects for reactor vessel internals.

NL-12-089Attachment 1Page 9 of 193.1.2.2.17 Cracking due to Stress Corrosion Cracking, Primary Water Stress Corrosion Cracking, andIrradiation-Assisted Stress Corrosion CrackingCracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking(PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) in PWR stainless steeland nickel alloy reactor vessel internals components will be managed by the Reactor VesselInternals (RVI) Program. The RVI Program will implement the EPRI Pressurized WaterReactor Internals Inspection and Evaluation Guidelines, MRP-227-A. The RVI Program willuse nondestructive examinations (NDE) and other inspection methods to manage agingeffects for reactor vessel internals.RAI 3The applicant addressed the further evaluation criteria in Section 3.1.2.2.12 of NUREG-1800,"Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants,"Rev. 1 (SRP-LR) by stating (in the "Discussion" column of Table 3.1.1 Item 3.1.1-30) thatcracking will be managed by the Water Chemistry Control Program (Primary and Secondary)and either the Reactor Vessel Internals (RVI) Program or the Inservice Inspection (ISI) Program.Crediting the ISI Program for managing cracking is inconsistent with LRA Tables 3.1.2-2-1P2and 3.1.2-2-1P3, in which the components aligned with Table 3.1.1 Item 3.1.1-30 only credit theWater Chemistry Control -Primary and Secondary Program and the RVI Program for agingmanagement. Further, LRA Amendment 9 does not include a revised LRA Section 3.1.2.2.12. Inaddition, the use of the Inservice Inspection Program (ISI) Aging Management Program (AMP)is not consistent with the NUREG-1 801, "Generic Aging Lessons Learned Report", Revision 1(GALL Report, Rev. 1), Table 1, Item 30 for this line item or the recommendations of SRP-LRSection 3.1.2.2.12.The staff therefore requests the following information:1. Correct the inconsistency between Table 3.1.1 Item 3.1.1-30 and the associated lineitems in Tables 3.1.2-2-1P2 and 3.1.2-2-:P3.2. Provide a markup to LRA Section 3.1.2.2.12 consistent with the changes in LRA Table3.1.1 provided in LRA Amendment 9.3. If the ISI Program is being used as the AMP to manage cracking for certain RVIcomponents aligned with Table 3.1.1 Item 3.1.1-30, justify the use of the ISI Programrather than the RVI Program for managing aging of the affected components, andmake all the necessary conforming changes to Table 3.1.1, Table 3.1.2-2-1P2, andTable 3.1.2-2-1P3.Response to RAI 31. There is no inconsistency between Table 3.1.1 Item 3.1.1-30 and the associated line itemsin Tables 3.1.2-2-1P2 and 3.1.2-2-1P3. In Tables 3.1.2-2-1P2 and 3.1.2-2-1P3, cracking forthe "Upper support plate, support assembly (including ring)" is managed by the WaterChemistry Control -Primary and Secondary and Inservice Inspection Programs. This itemis compared to NUREG-1 801, Rev. 1, Volume 2 Item IV.B2-42, and aligned to Table 1 Item3.1.1-30. Therefore, the RAI statement that "the components aligned with Table 3.1.1 Item3.1.1-30 only credit the Water Chemistry Control -Primary and Secondary Program and theRVI Program," is incorrect.

NL-12-089Attachment 1Page 10 of 192. LRA Section 3.1.2.2.12 was revised by Letter NL-11-101, dated August 22, 2011, to correctthe omission of the section from Amendment 9. LRA Section 3.1.2.2.12, as revised by NL-11-101 is shown below. Additional revisions are shown, with strikethrough for deletion andunderline for additions, to provide clarification on the use of the Inservice InspectionProgram, and the updated reference to MRP-227-A.3.1.2.2.12 Cracking due to Stress Corrosion Cracking and Irradiation-Assisted StressCorrosion Cracking (IASCC)Cracking due to SCC and IASCC in PWR stainless steel reactor internals exposed to reactorcoolant will be managed by the Water Chemistry Control -Primary and Secondary Programand the Reactor Vessel Internals (RVI) or Inservice Inspection (ISI) Programs. The RVIProgram will implement the EPRI Pressurized Water Reactor Internals Inspection andEvaluation Guidelines, MRP-227-A. The RVI Program will use nondestructive examinations(NDE) and other inspection methods to manage aging effects for reactor vessel internals. TheRVI Program includes inspections of core support structures using the existing ASME SectionXl, ISI Program as delineated in MRP-227-A. Table 4-9. Where credited for the management ofcracking, the existing ISI Proaram is listed in Tables 3.1.2-2-1P2 and 3.1.2-2-1P3 in lieu of theRVI Program.3. In Tables 3.1.2-2-1P2 and 3.1.2-2-1P3, cracking for the "Upper support plate, supportassembly (including ring)" is managed by the Water Chemistry Control -Primary andSecondary and Inservice Inspection Programs. This item is compared to NUREG-1801,Rev. 1, Volume 2 Item IV.B2-42, and aligned to Table 1 Item 3.1.1-30. This itemcorresponds to the matching entry in MRP-227-A, Table 4-9, Westinghouse Plants ExistingPrograms Components. Consistent with MRP-227-A, the ISI program is the "existingprogram" credited to manage cracking for this item. No other changes are required.RAI's Related to Reactor Vessel Internals ProaramRAI 4NUREG-1801, "Generic Aging Lessons Learned Report," Revision 2 (GALL Report, Rev. 2),Section XI.M16A, recommends, under the "Monitoring and Trending" program element, usingthe methods of the latest Nuclear Regulatory Commission (NRC)-approved version of MaterialsReliability Program: Pressurized Water Reactor Internals Inspection and Evaluation GuidelinesMRP-227, Section 6 for monitoring, recording, evaluating and trending the data from theprogram inspection results. MRP-227 Section 6 includes recommendations for flaw depth sizingand crack growth determinations as well as for performing applicable limit load, linear elasticand elastic-plastic fracture analyses of relevant flaw indications.However, in the staff's final safety evaluation (SE) on MRP-227, Revision 0 (Ref. 2), the staffnoted that in a request for additional information (RAI) response, Electric Power ResearchInstitute (EPRI) stated that topical report WCAP-1 7096-NP is the document that will be used asthe framework to develop those generic and plant-specific evaluations triggered by findings inthe RVI examinations, and observed that the NRC staff is currently reviewing WCAP-17096-NP.Revision 2. Therefore, the staff requests that the applicant clarify whether the Indian PointEnergy Center (IPEC) RVI Program will use the guidance of WCAP-1 7096-NP, Rev. 2 (Ref. 3)for evaluating the acceptability of relevant conditions found by the inspections conducted underthe RVI Inspection Plan.

NL-12-089Attachment 1Page 11 of 19Response to RAI 4The IPEC RVI Program plans to use the guidance of WCAP-1 7096-NP, Rev. 2 for evaluatingthe acceptability of relevant conditions found by the inspections conducted under the RVIInspection Plan.RAI_5For baffle-former bolts, MRP-227-A, Table 5-3 states that the examination acceptance criteriafor the ultrasonic test (UT) shall be established as part of the examination technical justification."Materials Reliability Program: Inspection Standard for PWR Internals," (MRP-228)(Ref. 4)provides additional guidance on preparation of technical justifications (TJs). However, the IPECRVI Program does not indicate whether a T J has been or will be developed for the baffle-formerbolts. Therefore, the staff requests the applicant submit a T J for the IP2 and IP3 baffle-formerbolts.Response to RAI 5MRP-227-A and its associated safety evaluation contain no requirement for the submittal of atechnical justification for these inspections with the application to implement MRP-227-A.Baffle-former bolt inspections are not required to be performed until between 25 and 35 EFPY.Currently both IPEC units are at less than 28 EFPY. Therefore, inspections are required prior to2019 at IP2 and 2021 at IP3. As a result, IPEC has not yet finalized the inspection scheduleand has not yet selected the vendor to perform the inspections. Since the technical justificationwill be prepared by the vendor selected to perform the inspections, a technical justification hasnot yet been prepared for IPEC. A technical justification is planned to be developed for thebaffle-former bolts when the inspection vendor is selected but no later than 6 months prior to thebeginning of the outage when the inspections will be performed.RAI's Related to Reactor Vessel Internals Inspection Plan (Ref. 6)RAI 6Applicant/Licensee Action Item 1 from the staff's final SE on MRP-227, Revision 0 requires thatapplicants/licensees submit an evaluation that demonstrates that their plant is bounded by theassumptions regarding plant design and operating history that were made in the failure modes,effects and consequences analyses (FMECA) and functionality analyses for reactors of theirdesign.The applicant's response to Applicant/Licensee Action Item 1 in the RVI inspection planaddresses the core loading assumptions (switch to a low-leakage core) and operational (baseloaded plant) aspects of design and operation that are mentioned in MRP-227-A, Section 2.4.An additional assumption listed in Section 2.4 of MRP-227-A is that there have been no designchanges to the RVI beyond those identified in general industry guidance or recommended bythe original vendors. Section 2.4 of MRP-227-A indicated that these assumptions areconsidered to conservatively represent any U.S. Pressurized Water Reactor operating plantprovided that these three assumptions are met, given the information on design and operationknown to the MRP as of May 2007.MRP-191, Revision 0, "Materials Reliability Program: Screening, Categorization and Ranking ofReactor Internals of Westinghouse and Combustion Engineering PWR Designs," documents the NL-12-089Attachment 1Page 12 of 19screening for susceptibility to aging effects, the FMECA results, and the categorization andranking of the RVI components. In addition to the assumptions listed in Section 2.4 of MRP-227-A, MRP-191 documents additional assumptions that were used. In particular, neutron fluencerange, temperature, and material grade for each generic component of the Westinghousedesign internals were used for input to the screening process. These values were determinedbased on an "expert elicitation" process. Stress values were not explicitly tabulated, but wererecorded as either above the stress threshold (>30 ksi) or not based on the expert interviews.MRP-232, Revision 0, "Materials Reliability Program: Aging Management Strategies forWestinghouse and Combustion Engineering PWR Internals," reported more specific stress,temperature and neutron fluence values based on finite element analyses for selected highconsequence of failure components identified in MRP-191.MRP-227 -A did not verify that the values of fluence, temperature, stress, and material,documented in MRP-191 and MRP-232 were bounding for all individual plants, and in fact MRP-227-A states, "These evaluations were based on representative configurations and operationalhistories, which were generally conservative, but not necessarily bounding in every parameter."Each plant should have access to design information enabling verification that the material foreach RVI component is bounded by the design assumptions of the MRP. In this context, thestaff requests the following information:1 ) To provide reasonable assurance that the RVI components are bounded by assumptions inthe FMECA and functionality analyses supporting the development of MRP-227-A, theapplicant is requested to respond to either 2.a or 2.b of this RAI:2.a)Provide the plant-specific values of neutron fluence (n/cm2, E>1.0 MeV), temperature,stress, and materials for a sample of RVI components. The components selected shouldrepresent a range of neutron fluences, and temperatures. This information should identifywhether the stress is greater or less than 30 ksi. Values of neutron fluence andtemperature may be estimated or analytical values. The values should be the peak valuesof each parameter for each component (e.g., peak end-of-life value for fluence). Providethe method used to estimate the values, or describe the analysis method. An acceptablesample of components is:i) Lower Core Plateii) Core Barrel Flangeiii) Barrel-Former Boltsiv) Upper Core Barrel Weldsv) Lower Core Barrel Weldsvi) Upper Core Plate Alignment Pins2.b) If the sample verification approach in Part (a) is not used, describe the process used toverify that the RVI components at IP2 and IP3 are bounded by the assumptions regardingthe neutron fluence, temperature, stress values, and materials that were made for eachcomponent in the FMECA and functionality analyses supporting the development of MRP-227-A.3) If there are any components at IP2 or IP3 not bounded by assumptions regarding neutronfluence, temperature, stress or material used in the development of MRP-227-A, describe NL-12-089Attachment 1Page 13 of 19how the differences were addressed in the plant-specific RVI Inspection Plan. The staffrequests that the applicant, as a part of its demonstration, discuss whether there would beany changes to the screening, categorization, FMECA process and functionality analyses ifthe plant-specific variables (the neutron fluence, temperature, stress values, plant-specificoperating experience, and materials) are used. This evaluation should address whetheradditional aging mechanisms would become applicable to the component.4) For any non-bounded components, determine if any changes to the inspection requirementsof MRP-227-A are needed. Provide plant-specific inspection requirements or an alternateaging management program, as appropriate. If no changes to the inspection requirementsare proposed, provide a justification for the adequacy of the existing MRP-227 -Ainspections for the unbounded components.5) Identify all design changes to the IP2 and IP3 RVI, and describe (1) if any of these arebeyond those identified in general industry guidance or recommended by the originalvendors, and (2) if any of the design changes were implemented after May 2007. Assess theimpact of these design changes on the recommendations of the RVI Inspection Plan.Provide plant-specific inspection requirements if necessary for the affected components.Response to RAI 6As noted in the cover letter the response to this RAI requires that additional evaluations beperformed. The response will be submitted to the NRC by 09/28/2012.RAI 7The staff reviewed the applicant's response to Applicant/Licensee Action Item 2 from the NRCstaff's final SE on MRP-227, Revision 0. In Section 3.6 of the RVI Inspection Plan (Ref. 5), theapplicant stated that it reviewed the information in Table 4-4 of MRP-191 and determined thatthis table contains all the RVI components that are within the scope of license renewal and thatthis is shown in Table 5-7. The staff notes that Table 5-1 contains a cross-index between thecomponent designations in Entergy Letter NL-1 0-063 (Amendment 9 to the LRA, Ref. 1) and thecomponent names as designated in MRP-191, Table 4-4 (Ref. 6). All the IPEC componentdesignations correlate with an equivalent component designation in MRP-191 (Ref. 7), Table 44with the exception of the Lower Internals Assembly -Column Cap.The staff therefore requests that the applicant verify that the Lower Internals Assembly -ColumnCap would be subject to the same inspection requirements that are applied to the lower supportassembly, lower support column bodies (cast) in MRP-227-A, Table 4-6. If not, provide plant-specific aging management requirements for the Lower Internals Assembly -Column Cap.Response to RAI 7As noted in the cover letter the response to this RAI requires that additional evaluations beperformed. The response will be submitted to the NRC by 09/28/2012.RAI 8The staff requests the following information related to the applicant's response toApplicant/Licensee Action Item 3 from the NRC staffs final SE on MRP-227, Revision 0.

NL-12-089Attachment 1Page 14 of 191. Provide more detail on the operating experience for cold-worked type 316 split pins tosupport the prediction that split pins of this material will last until the end of the period ofextended operation (PEO) for IP3.2. Describe the inspection schedule, methods, and basis for replacement split pins at IP3.If no inspections are planned, provide a justification for not inspecting the split pins.3. Describe the criteria for the replacement split pin material and design for IP2.4. Describe the inspection strategy for the replacement IP2 split pins during the PEO.Response to RAI 81: Cold-worked type 316 split pins have been installed at other nuclear power plants since1997. No plants have experienced failures of cold-worked type 316 split pins to date.2: No inspections are planned for the split pins at IP3. However, based on industryoperating experience, if failures of cold-worked type 316 split pins occur, an IP3 plantspecific evaluation will be performed at that time to determine if inspections are required.Since the IP3 split pins were replaced in 2009 and other plants have installed cold-worked type 316 split pins starting in 1997, failure of other plant split pins would beexpected before potential failures at IP3. Any failure would be evaluated by IP3 todetermine the need for an inspection and other actions. No plants have experienced anyfailures of cold-worked type 316 split pins to date.3: 1P2 plans to use the same replacement split pin material and design that was used forIP3. IP2 plans to use cold-worked type 316 split pins.4: The inspection strategy for the replacement IP2 split pins during the PEO will be thesame as the IP3 inspection strategy. No inspections are planned for the replacementIP2 split pins. However, based on industry operating experience, if failures of cold-worked type 316 split pins occur, an IP2 plant specific evaluation will be performed atthat time to determine if inspections are warranted.RAI 9The applicant's response to Applicant/Licensee Action Item 5 from Revision 1 of the staff's finalSE on MRP-227, states in part that the acceptance criteria will ensure the remainingcompressible height of the spring shall provide hold down forces within the IPEC designtolerance. If a plant specific acceptance criterion is not developed for the hold down spring,IPEC will replace the spring in lieu of performing the first required physical measurement.MRP-227-A, Table 4-3, calls for direct measurement of the hold-down spring height within threecycles of the beginning of the license renewal period. If the first set of measurements is notsufficient to determine life, spring height measurements must be taken during the next twooutages, in order to extrapolate the expected spring height to 60 years.The staff requires clarification of how the applicant will determine whether the first set ofmeasurements could be extrapolated to demonstrate acceptable spring functionality through 60years. Therefore, the staff requests the following information:

NL-12-089Attachment 1Page 15 of 191. Provide the specific acceptance criteria for spring height and/or hold down force from theIP2/IP3 licensing basis.2. Describe the procedure by which the remaining hold down forces will be projected toend-of-life based on one measurement. Address whether the decrease in spring heightor hold-down force is assumed to occur linearly over time or via some other function oftime.3. What results of the first spring measurements would indicate a need for successivemeasurements?Response to RAI 9As noted in the cover letter the response to this RAI requires that additional evaluations beperformed. The response will be submitted to the NRC by 09/28/2012.RAI 10The applicant's response to Applicant/Licensee Action Item 7 indicates that the plant-specificanalysis to demonstrate functionality of the lower support column bodies during the period ofextended operation will be submitted to the NRC prior to the PEO. In the aging managementreview tables submitted in LRA Amendment 9, the applicant credits the "Thermal Aging andNeutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program" formanaging loss of fracture toughness of the lower core support column bodies, as well asseveral other CASS components. NUREG-1 930 indicates that the staff determined this programwas consistent with the Generic Aging Lessons Learned Report, Revision 1, AgingManagement Program (AMP) XI.M13, "Thermal Aging and Neutron Irradiation Embrittlement ofCast Austenitic Stainless Steel (CASS) Program." Per GALL, Rev. 1,Section XI.M13, the"Thermal Aging and Neutron Irradiation Embrittlement of CASS Program" generally requiressupplemental visual inspections (equivalent to an EVT-1) for CASS RVI components that areeither susceptible to thermal aging based on chemistry and other manufacturing parameters, orreceive a neutron fluence > lx1017 n/cm2, unless it can be demonstrated that the stresses onthe component are either compressive or low in magnitude if tensile. The RVI Program iscredited with managing cracking of the core support column bodies and other CASScomponents. Under the RVI Program, the core support column bodies are expansioncomponents that would be subject to an EVT-1 visual examination for cracking due to irradiationassisted stress corrosion cracking if cracking were found in the associated primary component.The staff requests the following information:Since both the plant-specific analysis and Thermal Aging and Neutron Irradiation Embrittlementof CASS Program could both potentially involve screening for thermal or neutron irradiationembrittlement, stress analyses, and flaw tolerance evaluations, and both the RVI Program andThermal Aging and Neutron Irradiation Embrittlement of CASS Program could potentially requireinspections, discuss the relationship of the two programs and the plant-specific analysis.Response to RAI 10As noted in the cover letter the response to this RAI requires that additional evaluations beperformed. The response will be submitted to the NRC by 09/28/2012.

NL-12-089Attachment 1Page 16 of 19RAI 11In response to Applicant/Licensee Action Item 7, the applicant stated that the plant-specificanalyses to demonstrate the lower support column bodies will maintain their functionality duringthe period of extended operation will consider the possible loss of fracture toughness in thesecomponents due to thermal and irradiation embrittlement. The analyses will be consistent withthe IP2/1 P3 licensing basis and the need to maintain the functionality of the lower supportcolumn bodies under all licensing basis conditions of operations.The staff requests the following additional information:1) Section 3.3.7 of Revision 1 of the staff's final SE on MRP-227, Revision 0 lists threepossible options for the type of plant-specific analysis used to fulfill the requirements ofthis action item. The three approaches are 1) functionality analyses of the set of likecomponents, 2) component-specific flaw tolerance evaluations, or 3) a screeningapproach demonstrating that the CASS Components are not susceptible to thermalembrittlement, neutron embrittlement, or the combined effects of both. Discuss which ofthese approaches will be used and why.2) Describe the acceptance criteria for the plant-specific analysis results that are derivedfrom the IP2/IP3 licensing basis.3) Since the applicant stated that the analysis of the core support columns will be submittedprior to the period of extended operation for IP2 and IP3, the staff requests the applicantsubmit a letter documenting this as a formal licensing commitment.Response to RAI 11As noted in the cover letter the response to this RAI requires that additional evaluations beperformed. The response will be submitted to the NRC by 09/28/2012.Commitment 47IPEC will perform and submit analyses that demonstrate that the lower support column bodieswill maintain their functionality during the period of extended operation considering the possibleloss of fracture toughness due to thermal and irradiation embrittlement. The analyses will beconsistent with the IP2/IP3 licensing basis and will be submitted prior to the PEO.RAI 12BackgroundIn its letter dated February 17, 2012, the applicant provided the response to Applicant/LicenseeAction Item 8 of the Staff SE of MRP-227-A. The applicant stated that the RVI AMP descriptionhas been revised to be consistent with MRP-227-A, and the applicant's response toApplicant/Licensee Action Item 8 does not request any deviations from the guidance provided inMRP-227-A. The staff noted that Applicant/Licensee Action Item 8 also addresses cumulativeusage factor (CUF) analyses that are time-limited aging analyses (TLAAs).The applicant's response does not address LRA Section 4.3.1.2, which provides the applicant'sTLAA and associated CUF values for the IP2 and IP3 RVI. The staff noted that in Amendment 3 NL-12-089Attachment 1Page 17 of 19to the LRA dated March 24, 2008, (ADAMS Accession No. ML081070255), the applicantamended LRA Section 4.3.1.2 to state that "fatigue on the reactor vessel internals will bemanaged by the Fatigue Monitoring Program in accordance with 10 CFR 54.21 (c)(1)(iii) forboth I P2 and I P3."IssueThe staff noted that Applicant/Licensee Action Item 8 indicates that RVI Program may be usedas the basis for accepting CUF analyses in accordance with 10 CFR 54.21 (c)(1)(iii) only if theRVI components in the CUF analyses are periodically inspected for fatigue-induced crackingduring the period of extended operation. Applicant/Licensee Action Item 8 also indicates that theFatigue Monitoring Program may be used as the basis for accepting CUF analyses inaccordance with 10 CFR 54.21 (c)(1)(iii), in which case the evaluation requirements of ASMECode Section III, Section NG are to be satisfied.It is not clear to the staff whether the applicant will use (a) its RVI Program, (b) its FatigueMonitoring Program, or (c) a combination of both programs to manage RVI fatigue during theperiod of extended operation.RequestIdentify the aging management program that is used to manage fatigue of the reactor vesselinternals:1) If the RVI Program will be used:a. Verify that each RVI component with a CUF value will be periodically inspected forfatigue-induced cracking during the period of extended operation.b. For each component to be inspected for fatigue-induced cracking:i. Identify the examination method(s).ii. Provide the inspection periodicity, including the initial inspection timing andtiming of subsequent examinations.iii. Justify that the periodicity of the inspections for each RVI component isadequate.2) If the Fatigue Monitoring Program will be used, verify that the requirements of ASMECode Section lil, Subsections NG-2160 and NG-3121, as delineated inApplicant/Licensee Action Item 8, will be satisfied.Response to RAI 12IPEC will use the RVI Program to manage the effects of aging due to fatigue on the reactorvessel internals. The aging management strategy development described in MRP-227-A wasbased on consideration of susceptibility to eight age-related degradation mechanisms. Fatiguewas one of the eight degradation mechanisms considered. As provided in Section 3.5.1 of theNRC's safety evaluation for MRP-227-A, for locations with a fatigue time-limited aging analysis,IPEC will manage the effects of aging due to fatigue through the Fatigue Monitoring Program in NL-12-089Attachment 1Page 18 of 19accordance with 10 CFR 54.21(c)(1)(iii). For locations which do not have a current licensingbasis fatigue analysis, IPEC will rely on the inspection requirements of MRP-227-A to managethe effects of aging due to fatigue.Consistent with 10 CFR 54.21 (c)(1)(iii) and the NRC's safety evaluation for MRP-227-A, theFatigue Monitoring Program will manage the effects of aging due to fatigue on RVI componentswith a fatigue time-limited aging analysis. The Fatigue Monitoring Program as described in LRASection B. 1.12 provides assurance that the CUF remains below the allowable limit of 1.0.Consistent with Section 3.5.1 of the safety evaluation for MRP-227-A, prior to entering theperiod of extended operation the existing RVI fatigue calculations will be reviewed to evaluatethe effects of the reactor coolant system water environment on the CUF. Specifically, underCommitment 43, Entergy will review the IPEC design basis ASME Code Class 1 fatigueevaluations to determine whether the NUREG/CR-6260 locations that have been evaluated forthe effects of the reactor coolant environment on fatigue usage are the limiting locations for theIP2 and IP3 configurations. This review includes ASME Code Class 1 fatigue evaluations forreactor vessel internals. If more limiting locations are identified, the most limiting location will beevaluated for the effects of the reactor coolant environment on fatigue usage.References1. Letter from Fred Dacimo, Entergy, to NRC dated July 14, 2010,

Subject:

Amendment 9 toLicense Renewal Application (LRA) -Reactor Vessel Internals Program Indian Point NuclearGenerating Unit Nos. 2 & 3, Docket Nos. 50-247 and 50-286 License Nos. DPR-26 andDPR-64 (ADAMS Accession No. ML102010102)2. Letter from Robert Nelson, NRC, to Neil Wilmshurst, EPRI dated December 16, 2011;

Subject:

Revision 1 of the Final Safety Evaluation of EPRI Report, Materials ReliabilityProgram Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR)Internals Inspection and Evaluation Guidelines" (TAC No. ME0680) (ADAMS Accession No.ML11308A770)3. Reactor Internals Acceptance Criteria Methodology and Data Requirements, WCAP-17096-NP, Rev. 2, Westinghouse Non-Proprietary Class 3 Report, December 2009, ADAMSAccession No. ML10146015704. Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228) 1016609Final Report, July 2009 Electric Power Research Institute, Palo Alto, CA (EPRI Product No.1016609) (ADAMS Accession No. ML092120573)5. Indian Point Energy Center Revised Reactor Vessel Internals Inspection Plan Compliantwith MRP-227-A. Attachment 2 to Entergy Letter NL-1 2-037, Letter from Fred Dacimo toNRC dated February 17, 2012,

Subject:

License Renewal Application -Revised ReactorVessel Internals Program and Inspection Plan Compliant with MRP-227-A, Indian PointNuclear Generating Unit Nos. 2 and 3, Docket Nos. 50-247 and 50-286-License Nos. DPR-26 and DPR-64 (ADAMS Accession No. ML1206A312)6. MRP-191 Revision 0, "Materials Reliability Program: Screening, Categorization and Rankingof Reactor Internals of Westinghouse and Combustion Engineering PWR Designs," ADAMSAccession No. ML091910130 NL-12-089Attachment 1Page 19 of 197. NUREG-1930, Volume 2, "Safety Evaluation Report Related to The License Renewal ofIndian Point Nuclear Generating Unit Nos. 2 and 3, Dockets No. 50-247 and 50-286,November 30, 2009 (ADAMS Accession No. ML093170671)

ATTACHMENT 2 TO NL-12-089LICENSE RENEWAL APPLICATIONIPEC LIST OF REGULATORY COMMITMENTSRev. 18ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3DOCKET NOS. 50-247 AND 50-286 NL-12-089Attachment 2Page 1 of 18List of Regulatory CommitmentsRev. 18The following table identifies those actions committed to by Entergy in this document.Changes are shown as strikethroughs for deletiens and underlines for additions.# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM1 Enhance the Aboveground Steel Tanks Program for P2: NL-07-039 A.2.1.1IP2 and IP3 to perform thickness measurements of September 28, A.3.1.1the bottom surfaces of the condensate storage tanks, 013 B.1.1city water tank, and fire water tanks once during the IP3:first ten years of the period of extended operation. December 12,Enhance the Aboveground Steel Tanks Program for 2015IP2 and IP3 to require trending of thicknessmeasurements when material loss is detected.2 Enhance the Bolting Integrity Program for IP2 and IP3 IP2: NL-07-039 A.2.1.2to clarify that actual yield strength is used in selecting September 28, A.3.1.2materials for low susceptibility to SCC and clarify theprohibition on use of lubricants containing MoS2 for P3: NL-07-153 Audit Itemsbolting. December 12, 201,241,The Bolting Integrity Program manages loss of 015 270preload and loss of material for all external bolting. r I I _I NL-12-089Attachment 2Page 2 of 18COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI I_ 1 I/ AUDIT ITEM3Implement the Buried Piping and Tanks InspectionProgram for IP2 and IP3 as described in LRA SectionB.1.6.This new program will be implemented consistent withthe corresponding program described in NUREG-1801 Section XI.M34, Buried Piping and TanksInspection.Include in the Buried Piping and Tanks InspectionProgram described in LRA Section B.1.6 a riskassessment of in-scope buried piping and tanks thatincludes consideration of the impacts of buried pipingor tank leakage and of conditions affecting the risk forcorrosion. Classify pipe segments and tanks ashaving a high, medium or low impact of leakagebased on the safety class, the hazard posed by fluidcontained in the piping and the impact of leakage onreliable plant operation. Determine corrosion riskthrough consideration of piping or tank material, soilresistivity, drainage, the presence of cathodicprotection and the type of coating. Establishinspection priority and frequency for periodicinspections of the in-scope piping and tanks based onthe results of the risk assessment. Performinspections using inspection techniques withdemonstrated effectiveness.I P2:September 28,2013I P3:December 12,2015NL-07-039NL-07-153NL-09-106NL-09-111A.2.1.5A.3.1.5B.1.6Audit Item173NL-1 1-101 NL-12-089Attachment 2Page 3 of 18COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION________I AUDIT ITEM4Enhance the Diesel Fuel Monitoring Program toinclude cleaning and inspection of the IP2 GT-1 gasturbine fuel oil storage tanks, IP2 and IP3 EDG fuel oilday tanks, IP2 SBO/Appendix R diesel generator fueloil day tank, and IP3 Appendix R fuel oil storage tankand day tank once every ten years.Enhance the Diesel Fuel Monitoring Program toinclude quarterly sampling and analysis of the IP2SBO/Appendix R diesel generator fuel oil day tank,IP2 security diesel fuel oil storage tank, IP2 securitydiesel fuel oil day tank, and IP3 Appendix R fuel oilstorage tank. Particulates, water and sedimentchecks will be performed on the samples. Filterablesolids acceptance criterion will be less than or equalto 10mg/l. Water and sediment acceptance criterionwill be less than or equal to 0.05%.Enhance the Diesel Fuel Monitoring Program toinclude thickness measurement of the bottom of thefollowing tanks once every ten years. IP2: EDG fueloil storage tanks, EDG fuel oil day tanks,SBO/Appendix R diesel generator fuel oil day tank,GT-1 gas turbine fuel oil storage tanks, and diesel firepump fuel oil storage tank; IP3: EDG fuel oil daytanks, EDG fuel oil storage tanks, Appendix R fuel oilstorage tank, and diesel fire pump fuel oil storagetank.Enhance the Diesel Fuel Monitoring Program tochange the analysis for water and particulates to aquarterly frequency for the following tanks. IP2: GT-1gas turbine fuel oil storage tanks and diesel fire pumpfuel oil storage tank; IP3: Appendix R fuel oil day tankand diesel fire pump fuel oil storage tank.Enhance the Diesel Fuel Monitoring Program tospecify acceptance criteria for thicknessmeasurements of the fuel oil storage tanks within thescope of the program.Enhance the Diesel Fuel Monitoring Program to directsamples be taken and include direction to removewater when detected.Revise applicable procedures to direct sampling of theonsite portable fuel oil contents prior to transferringthe contents to the storage tanks.Enhance the Diesel Fuel Monitoring Program to directthe addition of chemicals including biocide when thepresence of biological activity is confirmed.IP2:September 28,2013I P3:December 12,2015NL-07-039NL-07-153NL-08-057A.2.1.8A.3.1.8B.1.9Audit items128,129,132,491,492,510______ ________________________________________________________ I ________________ I NL- 12-089Attachment 2Page 4 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM5 Enhance the External Surfaces Monitoring Program 1P2: NL-07-039 A.2.1.10for IP2 and IP3 to include periodic inspections of September 28, A.3.1.10systems in scope and subject to aging management 013 B.1.11review for license renewal in accordance with 10 CFR IP3:54.4(a)(1) and (a)(3). Inspections shall include areassurrounding the subject systems to identify hazards to Dcmr1those systems. Inspections of nearby systems thatcould impact the subject systems will include SSCsthat are in scope and subject to aging managementreview for license renewal in accordance with 10 CFR54.4(a)(2).6 Enhance the Fatigue Monitoring Program for IP2 to 1P2: NL-07-039 A.2.1.11monitor steady state cycles and feedwater cycles or September 28, A.3.1.11perform an evaluation to determine monitoring is not 013 B.1.12,required. Review the number of allowed events and NL-07-153 Audit Itemresolve discrepancies between reference documents 164and monitoring procedures.Enhance the Fatigue Monitoring Program for IP3 to IP3:include all the transients identified. Assure all fatigue December 12,analysis transients are included with the lowest 2015limiting numbers. Update the number of designtransients accumulated to date.I P2: NL-07-039 A.2.1.127 Enhance the Fire Protection Program to inspect Sptb 28, A.3.1.12external surfaces of the IP3 RCP oil collection Speb 2.1.12systems for loss of material each refueling cycle.Enhance the Fire Protection Program to explicitly IP3:state that the IP2 and IP3 diesel fire pump engine December 12,sub-systems (including the fuel supply line) shall be _015observed while the pump is running. Acceptancecriteria will be revised to verify that the diesel enginedoes not exhibit signs of degradation while running;such as fuel oil, lube oil, coolant, or exhaust gasleakage.Enhance the Fire Protection Program to specify thatthe IP2 and IP3 diesel fire pump engine carbon steelexhaust components are inspected for evidence ofcorrosion and cracking at least once each operatingcycle.Enhance the Fire Protection Program for IP3 tovisually inspect the cable spreading room, 480Vswitchgear room, and EDG room C02 fire suppressionsystem for signs of degradation, such as corrosionand mechanical damage at least once every sixmonths.

NL-12-089Attachment 2Page 5 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM8 Enhance the Fire Water Program to include inspection lP2: NL-07-039 A.2.1.13of IP2 and IP3 hose reels for evidence of corrosion. September 28, A.3.1.132013 B.1.14Acceptance criteria will be revised to verify no NL-07-153 Audit Itemsunacceptable signs of degradation. IP3: 105, 106Enhance the Fire Water Program to replace all or test December 12, NL-08-014a sample of IP2 and IP3 sprinkler heads required for 201510 CFR 50.48 using guidance of NFPA 25 (2002edition), Section 5.3.1.1.1 before the end of the 50-year sprinkler head service life and at 10-yearintervals thereafter during the extended period ofoperation to ensure that signs of degradation, such ascorrosion, are detected in a timely manner.Enhance the Fire Water Program to perform wallthickness evaluations of IP2 and IP3 fire protectionpiping on system components using non-intrusivetechniques (e.g., volumetric testing) to identifyevidence of loss of material due to corrosion. Theseinspections will be performed before the end of thecurrent operating term and at intervals thereafterduring the period of extended operation. Results ofthe initial evaluations will be used to determine theappropriate inspection interval to ensure aging effectsare identified prior to loss of intended function.Enhance the Fire Water Program to inspect theinternal surface of foam based fire suppression tanks.Acceptance criteria will be enhanced to verify nosignificant corrosion.

NL-12-089Attachment 2Page 6 of 18COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM9Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to implement comparisons to wearrates identified in WCAP-12866. Include provisions tocompare data to the previous performances andperform evaluations regarding change to testfrequency and scope.Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to specify the acceptance criteria asoutlined in WCAP-12866 or other plant-specific valuesbased on evaluation of previous test results.Enhance the Flux Thimble Tube Inspection Programfor IP2 and IP3 to direct evaluation and performanceof corrective actions based on tubes that exceed orare projected to exceed the acceptance criteria. Alsostipulate that flux thimble tubes that cannot beinspected over the tube length and cannot be shownby analysis to be satisfactory for continued service,must be removed from service to ensure the integrityof the reactor coolant system rressure boundary.IP2:September 28,2013I P3:December 12,2015NL-07-039A.2.1.15A.3.1.15B.1.16J. a i a.

NL-12-089Attachment 2Page 7 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONSI I / AUDIT ITEM10Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to include the following heat exchangersin the scope of the program.* Safety injection pump lube oil heat exchangers* RHR heat exchangers* RHR pump seal coolers* Non-regenerative heat exchangers* Charging pump seal water heat exchangers* Charging pump fluid drive coolers* Charging pump crankcase oil coolers* Spent fuel pit heat exchangers* Secondary system steam generator samplecoolers* Waste gas compressor heat exchangers* SBO/Appendix R diesel jacket water heatexchanger (IP2 only)Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to perform visual inspection on heatexchangers where non-destructive examination, suchas eddy current inspection, is not possible due to heatexchanger design limitations.Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to include consideration of material-environment combinations when determining samplepopulation of heat exchangers.Enhance the Heat Exchanger Monitoring Program forIP2 and IP3 to establish minimum tube wall thicknessfor the new heat exchangers identified in the scope ofthe program. Establish acceptance criteria for heatexchangers visually inspected to include no indicationof tube erosion, vibration wear, corrosion, pitting,foulinq, or scalinq.IP2:September 28,2013I P3:December 12,2015NL-07-039NL-07-153NL-09-018A.2.1.16A.3.1.16B.1.17,Audit Item5211 IeNL-09-056NL- 11-101 NL-12-089Attachment 2Page 8 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM12 Enhance the Masonry Wall Program for IP2 and IP3 P2: NL-07-039 A.2.1.18to specify that the IP1 intake structure is included in September 28, A.3.1.18the program. 2013 B.1.19IP3:December 12,201513Enhance the Metal-Enclosed Bus Inspection Programto add IP2 480V bus associated with substation A tothe scope of bus inspected.Enhance the Metal-Enclosed Bus Inspection Programfor IP2 and IP3 to visually inspect the external surfaceof MEB enclosure assemblies for loss of material atleast once every 10 years. The first inspection willoccur prior to the period of extended operation andthe acceptance criterion will be no significant loss ofmaterial.Enhance the Metal-Enclosed Bus Inspection Programto add acceptance criteria for MEB internal visualinspections to include the absence of indications ofdust accumulation on the bus bar, on the insulators,and in the duct, in addition to the absence ofindications of moisture intrusion into the duct.Enhance the Metal-Enclosed Bus Inspection Programfor IP2 and IP3 to inspect bolted connections at leastonce every five years if performed visually or at leastonce every ten years using quantitativemeasurements such as thermography or contactresistance measurements. The first inspection willoccur prior to the period of extended operation.The plant will process a change to applicable siteprocedure to remove the reference to "re-torquing"connections for phase bus maintenance and boltedconnection maintenance.I P2:September 28,2013I P3:December 12,2015NL-07-039NL-07-153NL-08-057A.2.1.19A.3.1.19B.1.20Audit Items124,133, 51914 Implement the Non-EQ Bolted Cable Connections 1P2: NL-07-039 A.2.1.21Program for IP2 and IP3 as described in LRA Section September 28, A.3.1.21B.1.22. 2013 B.1.22IP3:December 12,2015 NL-12-089Attachment 2Page 9 of 18COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM15 Implement the Non-EQ Inaccessible Medium-Voltage P2: NL-07-039 A.2.1.22Cable Program for IP2 and IP3 as described in LRA September 28, A.3.1.22Section B.1.23. 2013 B.1.23NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12, NL-1 1-0321801 Section XI.E3, Inaccessible Medium-Voltage 2015Cables Not Subject To 10 CFR 50.49 Environmental NL-1 1-096Qualification Requirements.NL-11-10116 Implement the Non-EQ Instrumentation Circuits Test P2: NL-07-039 A.2.1.23Review Program for IP2 and IP3 as described in LRA September 28, A.3.1.23Section B.1.24. 2013 B.1.24NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801 Section XI.E2, Electrical Cables and 2015Connections Not Subject to 10 CFR 50.49Environmental Qualification Requirements Used inInstrumentation Circuits.17 Implement the Non-EQ Insulated Cables and SP2: NL-07-039 A.2.1.24Connections Program for IP2 and IP3 as described in September 28,LRA Section B.1.25. 2013 B.1.25NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801 Section XI.E1, Electrical Cables and 2015Connections Not Subject to 10 CFR 50.49Environmental Qualification Requirements.18 Enhance the Oil Analysis Program for IP2 to sample IP2: NL-07-039 A.2.1.25and analyze lubricating oil used in the SBO/Appendix eptember 28, A.3.1.25R diesel generator consistent with the oil analysis for 013 NL-11-101 B.1.26other site diesel generators. I P3:Enhance the Oil Analysis Program for IP2 and IP3 to December 12,sample and analyze generator seal oil and turbine 2015hydraulic control oil.Enhance the Oil Analysis Program for IP2 and IP3 toformalize preliminary oil screening for water andparticulates and laboratory analyses including definedacceptance criteria for all components included in thescope of this program. The program will specifycorrective actions in the event acceptance criteria arenot met.Enhance the Oil Analysis Program for IP2 and IP3 toformalize trending of preliminary oil screening resultsas well as data provided from independentlaboratories.

NL-12-089Attachment 2Page 10 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM1P2: NL-07-039 A.2.1.2619 Implement the One-Time Inspection Program for IP2 A.2.1.26and IP3 as described in LRA Section B.1.27. September 28, A.3.1.262013 B.1.27This new program will be implemented consistent with NL-07-153 Audit itemthe corresponding program described in NUREG- P3: 1731801,Section XI.M32, One-Time Inspection. December 12,201520 Implement the One-Time Inspection -Small Bore P2: NL-07-039 A.2.1.27Piping Program for IP2 and IP3 as described in LRA September 28, A.3.1.27Section B.1.28. 2013 B.1.28NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801,Section XI.M35, One-Time Inspection of ASME 2015Code Class I Small-Bore Piping.21 Enhance the Periodic Surveillance and Preventive IP2: NL-07-039 A.2.1.28Maintenance Program for IP2 and IP3 as necessary September 28, A.3.1.28to assure that the effects of aging will be managedsuch that applicable components will continue to 3:perform their intended functions consistent with thecurrent licensing basis through the period of extended December 12,operation. 01522 Enhance the Reactor Vessel Surveillance Program for P2: NL-07-039 A.2.1-31IP2 and IP3 revising the specimen capsule withdrawal September 28, A.3.1.31schedules to draw and test a standby capsule to 013 B.1.32cover the peak reactor vessel fluence expected IP3:through the end of the period of extended operation. December 12,Enhance the Reactor Vessel Surveillance Program for 2015IP2 and IP3 to require that tested and untestedspecimens from all capsules pulled from the reactorvessel are maintained in storage.23 Implement the Selective Leaching Program for IP2 P2: NL-07-039 A.2.1.32and IP3 as described in LRA Section B.1.33. September 28, A.3.1.322013 B.1.33This new program will be implemented consistent with NL-07-153 Audit itemthe corresponding program described in NUREG- IP3: 1731801,Section XI.M33 Selective Leaching of Materials. December 12,201524 Enhance the Steam Generator Integrity Program for IP2: NL-07-039 A.2.1.34IP2 and IP3 to require that the results of the condition September 28, A1.34monitoring assessment are compared to theoperational assessment performed for the prior IP3:operating cycle with differences evaluated. December 12,2015 NL-12-089Attachment 2Page 11 of 18COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LLRA SECTION/ AUDIT ITEM25Enhance the Structures Monitoring Program toexplicitly specify that the following structures areincluded in the program.* Appendix R diesel generator foundation (IP3)* Appendix R diesel generator fuel oil tank vault(IP3)* Appendix R diesel generator switchgear andenclosure (IP3)" city water storage tank foundation* condensate storage tanks foundation (IP3)* containment access facility and annex (IP3)* discharge canal (IP2/3)* emergency lighting poles and foundations (IP2/3)* fire pumphouse (IP2)* fire protection pumphouse (IP3)" fire water storage tank foundations (IP2/3)" gas turbine 1 fuel storage tank foundation* maintenance and outage building-elevatedpassageway (I P2)* new station security building (IP2)* nuclear service building (IP1)" primary water storage tank foundation (IP3)* refueling water storage tank foundation (IP3)* security access and office building (IP3)" service water pipe chase (IP2/3)* service water valve pit (IP3)* superheater stack* transformer/switchyard support structures (IP2)* waste holdup tank pits (IP2/3)Enhance the Structures Monitoring Program for IP2and IP3 to clarify that in addition to structural steeland concrete, the following commodities (includingtheir anchorages) are inspected for each structure asapplicable.* cable trays and supports* concrete portion of reactor vessel supports" conduits and supports* cranes, rails and girders* equipment pads and foundations* fire proofing (pyrocrete)* HVAC duct supports* jib cranes* manholes and duct banks" manways, hatches and hatch covers* monorailsIP2:September 28,20131P3:December 12,2015NL-07-039NL-07-153NL-08-057A.2.1.35A.3.1.35B.1.36Audit items86, 87, 88,417 NL-12-089Attachment 2Page 12 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LLRA SECTION/AUDIT ITEM00new fuel storage rackssumps, sump screens, strainers and flow barriersEnhance the Structures Monitoring Program for IP2and IP3 to inspect inaccessible concrete areas thatare exposed by excavation for any reason. I P2 andIP3 will also inspect inaccessible concrete areas inenvironments where observed conditions inaccessible areas exposed to the same environmentindicate that significant concrete degradation isoccurring.Enhance the Structures Monitoring Program for IP2and IP3 to perform inspections of elastomers (seals,gaskets, seismic joint filler, and roof elastomers) toidentify cracking and change in material propertiesand for inspection of aluminum vents and louvers toidentify loss of material.Enhance the Structures Monitoring Program for IP2and IP3 to perform an engineering evaluation ofgroundwater samples to assess aggressiveness ofgroundwater to concrete on a periodic basis (at leastonce every five years). IPEC will obtain samples fromat least 5 wells that are representative of the groundwater surrounding below-grade site structures andperform an engineering evaluation of the results fromthose samples for sulfates, pH and chlorides.Additionally, to assess potential indications of spentfuel pool leakage, IPEC will sample for tritium ingroundwater wells in close proximity to the IP2 spentfuel pool at least once every 3 months.Enhance the Structures Monitoring Program for IP2and IP3 to perform inspection of normally submergedconcrete portions of the intake structures at least onceevery 5 years. Inspect the baffling/grating partition andsupport platform of the IP3 intake structure at leastonce every 5 years.Enhance the Structures Monitoring Program for IP2and IP3 to perform inspection of the degraded areasof the water control structure once per 3 years ratherthan the normal frequency of once per 5 years duringthe PEO.NL-08-127Audit Item360Audit Item358 NL-12-089Attachment 2Page 13 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMEnhance the Structures Monitoring Program toinclude more detailed quantitative acceptance criteria NL-1 1-032for inspections of concrete structures in accordancewith ACI 349.3R, "Evaluation of Existing NuclearSafety-Related Concrete Structures" prior to theperiod of extended operation. NL-1 1-10126 Implement the Thermal Aging Embrittlement of Cast P2: NL-07-039 A.2.1.36Austenitic Stainless Steel (CASS) Program for IP2 September 28, A.3.1.36and IP3 as described in LRA Section B.1.37. 2013 B.1.37NL-07-153 Audit itemThis new program will be implemented consistent with IP3: 173the corresponding program described in NUREG- December 12,1801,Section XI.M12, Thermal Aging Embrittlement 2015of Cast Austenitic Stainless Steel (CASS) Program.27 Implement the Thermal Aging and Neutron Irradiation IP2: NL-07-039 A.2.1.37Embrittlement of Cast Austenitic Stainless Steel September 28, A.3.1.37(CASS) Program for IP2 and IP3 as described in LRA 013 B.1.38Section B.1.38. NL-07-153 Audit itemIP3: 173This new program will be implemented consistent with December 12,the corresponding program described in NUREG- 20151801 Section XI.M13, Thermal Aging and NeutronEmbrittlement of Cast Austenitic Stainless Steel(CASS) Program.28 Enhance the Water Chemistry Control -Closed 1P2: NL-07-039 A.2.1.39Cooling Water Program to maintain water chemistry of September 28, A.3.1.39the IP2 SBO/Appendix R diesel generator cooling 2013 B.1. .40system per EPRI guidelines. IP3: 509Enhance the Water Chemistry Control -Closed December 12,Cooling Water Program to maintain the IP2 and IP3 2015security generator and fire protection diesel coolingwater pH and glycol within limits specified by EPRIguidelines.29 Enhance the Water Chemistry Control -Primary and P2: NL-07-039 A.2.1.40Secondary Program for IP2 to test sulfates monthly in September 28, B.1.41the RWST with a limit of <150 ppb. 01330 For aging management of the reactor vessel internals, P2: NL-07-039 A.2.1.41IPEC will (1) participate in the industry programs for September 28, A.3.1.41investigating and managing aging effects on reactor 011internals; (2) evaluate and implement the results of P3:the industry programs as applicable to the reactor December 12,internals; and (3) upon completion of these programs, 013but not less than 24 months before entering "the periodof extended operation, submit an inspection plan forreactor internals to the NRC for review and approval. Complete NL-11-107 NL-12-089Attachment 2Page 14 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM31 Additional P-T curves will be submitted as required 1P2: NL-07-039 A.2.2.1.2per 10 CFR 50, Appendix G prior to the period of September 28, A.3.2.1.2extended operation as part of the Reactor Vessel 013 4.2.3Surveillance Program. I P3:December 12,201532 As required by 10 CFR 50.61 (b)(4), IP3 will submit a IP3: NL-07-039 A.3.2.1.4plant-specific safety analysis for plate B2803-3 to the December 12, 4.2.5NRC three years prior to reaching the RTPTS 2015 NL-08-127screening criterion. Alternatively, the site may chooseto implement the revised PTS rule when approved.33At least 2 years prior to entering the period ofextended operation, for the locations identified in LRATable 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), underthe Fatigue Monitoring Program, IP2 and IP3 willimplement one or more of the following:(1) Consistent with the Fatigue Monitoring Program,Detection of Aging Effects, update the fatigue usagecalculations using refined fatigue analyses todetermine valid CUFs less than 1.0 when accountingfor the effects of reactor water environment. Thisincludes applying the appropriate Fen factors to validCUFs determined in accordance with one of thefollowing:1. For locations in LRA Table 4.3-13 (IP2) and LRATable 4.3-14 (1P3), with existing fatigue analysis validfor the period of extended operation, use the existingCUF.2. Additional plant-specific locations with a valid CUFmay be evaluated. In particular, the pressurizer lowershell will be reviewed to ensure the surge nozzleremains the limiting component.3. Representative CUF values from other plants,adjusted to or enveloping the IPEC plant specificexternal loads may be used if demonstrated applicableto IPEC.4. An analysis using an NRC-approved version of the ASMEcode or NRC-approved alternative (e.g., NRC-approved codecase) may be performed to determine a valid CUF.(2) Consistent with the Fatigue Monitoring Program,Corrective Actions, repair or replace the affectedlocations before exceeding a CUF of 1.0.I P2:September 28,20111P3:December 12,2013CompleteNL-07-039NL-07-153NL-08-021NL-10-082A.2.2.2.3A.3.2.2.34.3.3Audit item146 NL-12-089Attachment 2Page 15 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM34 IP2 SBO / Appendix R diesel generator will be April 30, 2008 NL-07-078 2.1.1.3.5installed and operational by April 30, 2008. This Complete NL-08-074committed change to the facility meets therequirements of 10 CFR 50.59(c)(1) and, therefore, a NL-11-101license amendment pursuant to 10 CFR 50.90 is notrequired. ___35 Perform a one-time inspection of representative P2: NL-08-127 Audit Itemsample area of IP2 containment liner affected by the September 28, 271973 event behind the insulation, prior to entering the 013period of extended operation, to assure linerdegradation is not occurring in this area. NL-11-101Perform a one-time inspection of representative IP3:sample area of the IP3 containment steel liner at the December 12,juncture with the concrete floor slab, prior to entering 2015the period of extended operation, to assure linerdegradation is not occurring in this area.Any degradation will be evaluated for updating of the NL-09-018containment liner analyses as needed.1P2: NL-08-127 Audit Item36 Perform a one-time inspection and evaluation of a Spt r NL-08-101 359sample of potentially affected IP2 refueling cavity 2810concrete prior to the period of extended operation.The sample will be obtained by core boring therefueling cavity wall in an area that is susceptible toexposure to borated water leakage. The inspectionwill include an assessment of embedded reinforcingsteel.Additional core bore samples wi!l be taken, if the NL-09-056leakage is not stopped, prior to the end of the first tenyears of the period of extended operation.A sample of leakage fluid will be analyzed to NL-09-079determine the composition of the fluid. If additionalcore samples are taken prior to the end of the first tenyears of the period o6i extended operation, a sample ofleakage fluid will be analyzed.IP2: NL-08-127 Audit Item37 Enhance the Containment Inservice Inspection (CII-IWL) Program to include inspections of the September 28, 361containment using enhanced characterization of 013degradation (i.e., quantifying the dimensions of noted P3:indications through the use of optical aids) during the December 12,period of extended operation. The enhancement 015includes obtaining critical dimensional data ofdegradation where possible through directmeasurement or the use of scaling technologies forphotographs, and the use of consistent vantage pointsfor visual inspections.

NL-12-089Attachment 2Page 16 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEMIP2: NL-08-143 4.2.138 For Reactor Vessel Fluence, should future coreloading patterns invalidate the basis for the projected September 28,values of RTpts or CRUSE, updated calculations will 013be provided to the NRC. IP3:December 12,201539 Deleted NL-09-07940 Evaluate plant specific and appropriate industry P2: NL-09-106 B.1.6operating experience and incorporate lessons learned September 28, B.1.22in establishing appropriate monitoring and inspection 013 B.1.23frequencies to assess aging effects for the new aging IP3: B.1.25management programs. Documentation of the De r1 B.1.25operating experience evaluated for each new program December 12, B.1.27will be available on site for NRC review prior to the B.1.28period of extended operation. B.1.33B.1. 37B. 1.38IP2: NL-11-032 N/A41 IPEC will inspect steam generators for both units to fter theassess the condition of the divider plate assembly. beg othThe examination technique used will be capable of beginning of thedetecting PWSCC in the steam generator divider plate September 28,assembly. The IP2 steam generator divider plate 2023 28,inspections will be completed within the first ten years 023 NL-1 1-074of the period of extended operation (PEO). The IP3steam generator divider plate inspections will be Pro No 1-090completed within the first refueling outage following of the first NL-11-101the beginning of the PEO. refueling outageollowing thebeginning of thePEO. I I NL-12-089Attachment 2Page 17 of 18COMMITMENT 1IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI_ I I / AUDIT ITEM42IPEC will develop a plan for each unit to address thepotential for cracking of the primary to secondarypressure boundary due to PWSCC of tube-to-tubesheet welds using one of the following twooptions.Option 1 (Analysis)IPEC will perform an analytical evaluation of thesteam generator tube-to-tubesheet welds in order toestablish a technical basis for either determining thatthe tubesheet cladding and welds are not susceptibleto PWSCC, or redefining the pressure boundary inwhich the tube-to-tubesheet weld is no longerincluded and, therefore, is not required for reactorcoolant pressure boundary function. The redefinitionof the reactor coolant pressure boundary must beapproved by the NRC as a license amendmentrequest.Option 2 (Inspection)IPEC will perform a one-time inspection of arepresentative number of tube-to-tubesheet welds ineach steam generator to determine if PWSCCcracking is present. If weld cracking is identified:a. The condition will be resolved through repairor engineering evaluation to justify continuedservice, as appropriate, andb. An ongoing monitoring program will beestablished to perform routine tube-to-tubesheet weld inspections for the remaininglife of the steam aenerators.I P2:Prior to March?0241P3: Prior to theend of the firstrefueling outagefollowing thebeginning of thePEO.I P2:Between March2020 and March20241P3: Prior to theend of the firstrefueling outagelollowing thebeginning of thePEO.NL-1 1-032NL-1 1-074NL-1 1-090NL-1 1-096N/A+ I I 443IPEC will review design basis ASME Code Class 1fatigue evaluations to determine whether theNUREG/CR-6260 locations that have been evaluatedfor the effects of the reactor coolant environment onfatigue usage are the limiting locations for the IP2 andIP3 configurations. If more limiting locations areidentified, the most limiting location will be evaluatedfor the effects of the reactor coolant environment onfatigue usage.IPEC will use the NUREG/CR-6909 methodology inthe evaluation of the limiting locations consisting ofnickel alloy, if any.I P2:Prior toSeptember 28,20131P3: Prior toDecember 12,2015NL-1 1-032NL-1 1-1014.3.3 NL-12-089Attachment 2Page 18 of 18# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM44 IPEC will include written explanation and justification P2: NL-11-032 N/Aof any user intervention in future evaluations using the Prior toWESTEMS "Design CUF" module. 2013IP3: Prior toDecember 12,20151P2: NL-11-032 N/A45 IPEC will not use the NB-3600 option of the Pro toWESTEMS program in future design calculations untilthe issues identified during the NRC review of the September 28, NL-1 1-101program have been resolved. 2013IP3: Prior toDecember 12,2015IP2: NL-11-032 N/A46 Include in the IP2 ISI Program that IPEC will perform Prior totwenty-five volumetric weld metal inspections ofsocket welds during each 10-year ISI interval September 28, NL-1 1-074scheduled as specified by IWB-2412 of the ASME 013Section Xl Code during the period of extendedoperation.In lieu of volumetric examinations, destructiveexaminations may be performed, where onedestructive examination may be substituted for twovolumetric examinations.I P2: N L- 12-089 N/A47 IPEC will perform and submit analyses that Pro todemonstrate that the lower support column bodies will Se temaintain their functionality during the period of September 28,extended overation considering the possible loss of 013fracture toughness due to thermal and irradiation 1P& Prior toembrittlement. The analyses will be consistent with December 12the IP2/IP3 licensing basis. 2015