ML14225A777
| ML14225A777 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 08/05/2014 |
| From: | Dacimo F Entergy Nuclear Northeast |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-14-093, TAC MD5407, TAC MD5408 | |
| Download: ML14225A777 (31) | |
Text
S'EEntergyEnteray Nuclear NortheastIndian Point Energy Center450 Broadway, GSBP.O. Box 249Buchanan, NY 10511-0249Tel (914) 254-2055Fred DacimnoVice PresidentOperations License RenewalNL-14-093August 05, 2014U.S. Nuclear Regulatory CommissionDocument Control Desk11545 Rockville Pike, TWFN-2 F1Rockville, MD 20852-2738SUBJECT:REFERENCES: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. Entergy letter (NL-14-067), "Reply to Request for Additional InformationRegarding the License Renewal Application," dated June 9, 2014.2. NRC letter, "Request for Additional Information for the Review of theIndian Point Nuclear Generating Unit Nos. 2 and 3, License RenewalApplication, SET 2014-02 (TAC Nos. MD5407 and MD5408)" dated April9, 2014.3. NRC letter, "Summary of Conference Calls Held on June 19 and July 16,2014, Between the U.S. Regulatory Commission and Entergy NuclearOperations, Inc., Entergy's Response to a Request for AdditionalInformation on Applicant/Licensee Action Item 7 from MRP-227-A," datedAugust 1, 2014.Dear Sir or Madam:Entergy Nuclear Operations, Inc. submitted in Reference 1 a response to NRC RAIs 11-C, 16-Band 17-A (Reference 2). Following that submittal Entergy and the NRC held two clarificationcalls concerning the response to RAI 11-C (Reference 3). As a result of those calls Entergy issubmitting, in Attachment 1, a revised response to RAI 11-C. The responses to RAIs 16-B and17-A remain unchanged.The revised response to RAI 11-C includes the deletion of Commitment 47. This commitmentchange is depicted in the list of regulatory commitments provided in Attachment 2.If you have any questions, or require additional information, please contact Mr. Robert Walpoleat 914-254-6710.
Docket Nos. 50-247 & 50-286NL-14-093Page 2 of 2I declare uder penalty of perjury that the foregoing is true and correct. Executed on$ ,2014.FRD/rwAttachment: 1.Reply to NRC Request for Additional Information Regarding the LicenseRenewal Application2. License Renewal Application IPEC List of Regulatory CommitmentsRevision 24cc: 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 IMs. Kimberly Green, 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. John B. Rhodes, President and CEO NYSERDA ATTACHMENT 1 TO NL-14-093REPLY 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-14-093Attachment 1Page 1 of 6REQUEST FOR ADDITIONAL INFORMATION, SET 2014-02RELATED TO INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3LICENSE RENEWAL APPLICATIONREACTOR VESSEL INTERNALS PROGRAM AND INSPECTION PLANRAI 11-CApplicant/Licensee Action Item 7 from the staff's final safety evaluation (SE) of MRP-227requires the applicants/licensees of Babcock & Wilcox (B&W), Combustion Engineering (CE),and Westinghouse reactors to develop plant-specific analyses to be applied for their facilitiesto demonstrate that B&W In-Core Monitoring Instrumentation (IMI) guide tube assemblyspiders and control rod guide tube (CRGT) assembly spacer castings, CE lower supportcolumns, and Westinghouse lower support column bodies will maintain their functionalityduring the period of extended operation, and states that these analyses should also considerthe possible loss of fracture toughness in these components due to thermal embrittlement(TE) and irradiation embrittlement (IE). For Indian Point Nuclear Generating Unit Nos. 2 and 3(IP2 and IP3), the equivalent component to the lower support column bodies are the lowerinternals assembly -column caps (column caps).By letter dated January 28, 2014, Entergy provided plant specific information on the ferritecontent and susceptibility to TE for the Indian Point Nuclear Generating Unit Nos. 2 and 3lower internals assembly -column caps. Based on its evaluation of the plant-specific materialinformation for the column caps, Entergy concluded that the IP2 and IP3 column caps are notsusceptible to TE.Entergy's conclusion related to TE notwithstanding, the column caps remain susceptible to IE.The staff is concerned that the linked Primary component for the column caps, the CRGTassembly lower flange welds, is not a good predictor of IE for the column caps since theCRGT lower flange welds receive substantially lower neutron fluence than the column caps(based on the estimated neutron fluence tabulated in MRP-191 for the two components).Irradiation assisted stress corrosion cracking (IASCC) is the only mechanism of cracking thatscreened in for the column caps. The CRGT lower flange welds are also not a good predictorfor IASCC of the column caps, because the lower flange welds are susceptible to stresscorrosion cracking (SCC) and fatigue cracking, but not IASCC.The staff, therefore, requests that Entergy modify its Reactor Vessel Internals InspectionPlan (RVI Inspection Plan) to provide a link to a Primary component or components that isan appropriate predictor of IE and IASCC of the column caps.Response to RAI 11-CThe following text will be incorporated into the reactor vessel internals (RVI) Inspection Planfor Indian Point Units 2 and 3. This text regards a link to a component that is an appropriatepredictor of irradiation embrittlement (IE) and irradiation-assisted stress corrosion cracking(IASCC) of the column caps (i.e., lower internals assembly -column caps).
NL-14-093Attachment 1Page 2 of 6"The cast austenitic stainless steel lower support column bodies (column caps) are listed inMRP-227-A [5, Table 4-6] as "Expansion" components, with the inspection triggered by theobservation of cracking in the control rod guide tube (CRGT) lower flange weld, which is a"Primary" component. The degradation mechanisms that were screened in for thesecomponents are identified in MRP-191 [2], a supporting report for MRP-227-A, and aresummarized in Table 1, below. IE, which has no demonstrated non-destructive examination(NDE) technique, is identified in Table 1 below as a degradation mechanism for the lowersupport column bodies (column caps) and CRGT lower flange weld. The "Primary" to"Expansion" linkage in MRP-227-A was based on consideration of the total relativesusceptibility of the components due to the combined application of all screened in degradationmechanisms. Priority is given to the leading indicator and the inspection technique and timingfocus on initial onset detectability of the effects of the degradation using proven NDEtechniques. The relevant degradation mechanisms that could drive cracking in the lowersupport column bodies (column caps) are IE and IASCC.The "Primary" inspection component for the cast section of the lower support column body(column cap) was originally the cast CRGT lower flange weld. Both of these components wereidentified as potentially subject to thermal embrittlement (TE) in MRP-191 [2]. However, plant-specific evaluations indicate that TE is not a concern for Indian Point Units 2 and 3 because theHull's ferrite number is less than the industry screening criterion of 20%. The ferrite content isbelow 15% for Indian Point Unit 2 and below 12% for Indian Point Unit 3 [4]. Although the lowferrite content eliminates the TE concerns in the lower support column bodies (column caps) atIndian Point Units 2 and 3, the MRP-227-A link to the "Primary" CRGT lower flange weldinspection was maintained to be consistent with the NRC direction in the Safety Evaluation [5]for MRP-227-A.An additional "Primary" inspection link for the lower support column bodies (column caps) isbeing added at both Indian Point units because the 60-year fluence in the top few inches of thelower support column body, which is where the column cap is located, is projected to exceedthe threshold criteria for IE and IASCC of austenitic stainless steel. Therefore, regardless of thematerial type or ferrite content, IE and IASCC must be considered as damage mechanisms inthe lower support column bodies (column caps).A review of the screening results from MRP-1 91 [2] indicates multiple near-core componentsthat potentially experience neutron fluences and stresses over the IE and IASCCthreshold/screening criterion, similar to the lower support column bodies (column caps). Fromthis set of components, as explained below, the lower core barrel cylinder girth weld is anappropriate leading indicator of IE and IASCC in the lower support column bodies (columncaps).MRP-227-A [5] requires an EVT-1 inspection of welds in both the upper and lower core barrelsections. Although many of these core barrel welds do not exceed the neutron fluencethreshold for either IE or IASCC, the lower core barrel cylinder girth weld has been identified inMRP-191 as being subject to both mechanisms. Selection of the lower core barrel cylinder girthweld, which incurs the highest fluence, as a leading indicator of IE and IASCC in the lowersupport columns is based on consideration of material, temperature, fluence, and stressfactors.
NL-14-093Attachment 1Page 3 of 61. Material: The process of melting and solidification that occurs during the weldingprocess produces a steel microstructure that is similar to an austenitic casting with lowferrite content.2. Temperature: Reactor coolant at the core barrel exterior and the lower core supportcolumns is at core inlet temperatures. Gamma heating effects in these components areexpected to be small. The analysis provided in MRP-230 [6] indicates relatively smalltemperature increases in the core barrel and lower core support structure.3. Fluence: Examination of neutron fluence data from the Indian Point units indicates thatthere is a general correspondence between the peak neutron fluence in the core barrel [7]and the fluence range in the top four centimeters of the lower support columns [8]. The 60-year peak fast neutron fluence projected in the lower core barrel after 48 effective fullpower years of operation will be approximately 2.5x1 021 n/cm2 (E > 1 MeV) = 4dpa. Afterthe same period of operation, the peak fast neutron fluence at the top surface of the lowersupport columns near the core centerline will be approximately, 4.1 lx1021 n/cm2 (E >1 MeV) = 6 dpa. However, the attenuation of fluence below the core is steep and the peakfast neutron fluence in the lower support columns at a distance of four centimeters belowthe lower core plate will be 2.07x1 021 n/cm2 (E > 1MeV) = 3 dpa.4. Stress: Although it is difficult to quantify the stresses in the lower core barrel cylindergirth weld and the lower support column, it is reasonable to anticipate large surfacestresses in the lower core barrel cylinder girth weld. Under typical welding procedures, largetensile stresses approaching the unirradiated yield stress can be generated at the weldsurfaces as the metal cools. Additional tensile stresses may be generated on the outersurface of the core barrel due to temperature gradients caused by the attenuation ofgamma heating through the barrel thickness. Axial loads associated with the core supportfunction of the barrel may also contribute to the tensile stress on the core barrel weld. Incontrast, the core support loads on the lower support columns are primarily compressiveand there are no anticipated residual stresses. Thermal deflections caused by heating ofthe lower core plate may impose a surface stress due to bending of the lower supportcolumns. However, these stresses are expected to be small compared to the stress in thelower core barrel cylinder girth weld.Given the similarities between material, temperature, and fluence in the lower supportcolumns and the lower core barrel cylinder girth weld, stress is the leading factor thatdetermines the relative susceptibility to degradation effects caused by IE and IASCC. Giventhe significantly higher stress, the lower core barrel cylinder girth weld is a leading indicatorof IE and IASCC in the lower support columns.The MRP-227-A [5] requirements are augmented by including the MRP-227-A required EVT-1inspection of the lower core barrel cylinder girth weld as an additional "Primary" link for an"Expansion" inspection of the Indian Point lower support column bodies (column caps) formanaging IE and IASCC. Consistent with existing guidance in MRP-227-A, confirmation of asurface crack greater than two inches in length will be taken as an indication of active crackingin the lower core barrel cylinder girth weld. This confirmation will require expansion of theinspection to the Indian Point lower support column bodies (column caps). The link betweenthe CRGT lower flange welds and the lower support column bodies (column caps) provided inMRP-227-A remains unchanged.
NL-14-093Attachment 1Page 4 of 6The expansion inspection will be conducted within two refueling outages of the originalobservation of cracking in the lower core barrel cylinder girth weld. It is reasonable to expect asignificant delay in the onset of cracking in the lower support column bodies (column caps)based on the difference in stress. Existing analysis of laboratory data [3] indicates that largestresses are required to initiate IASCC at fluences below 10 dpa. The lower support columnbodies (column caps) are projected to be primarily in a compressive stress state; therefore,they are expected to have a higher threshold fluence for IE and IASCC than the core barrel.Completion of the "Expansion" inspection within two refueling outages (each fuel cycle is 24months in duration) represents only 13% more than the approximately 32 effective full poweryears of service experienced prior to the inspection of the core barrel cylinder girth weld. Thisprovides reasonable assurance that IE and IASCC degradation, if any, will be detected in atimely fashion."The intent of Commitment 47 was to demonstrate that the lower support column bodies (columncaps) will maintain their functionality during the period of extended operation considering thepossible loss of fracture toughness due to thermal and irradiation embrittlement. As the staffnoted in this RAI 11-C, Entergy has concluded that the IP2 and IP3 lower support column bodies(column caps) are not susceptible to TE and will modify the Reactor Vessel Internals (RVI)Inspection Plan to credit inspection of the lower core barrel cylinder girth weld as an appropriatepredictor of IE and IASCC in the lower support column bodies (column caps). Therefore, thefunctionality of the lower support column bodies (column caps) is assured during the period ofextended operation without the need for a functionality analysis. Therefore, Entergy is deletingCommitment 47 (additions are underlined, deletions lined out).Commitment 47IPE-C- Will pe~form and submit analyses that de.monsrtrate- that the lower support coILumn bodiesAPwRll Rmitain their fUnctigRality drlirg the porild of exteRded operation conRidering the possiblelos~s of fracture toughness due to ther~m.al and- irrad-iationA embrittle-mont. The analyses will beconsistent With the 1P2/!P3 icensing basis. Deleted.References1. U.S. Nuclear Regulatory Commission, "Request for Additional Information for the Reviewof the Indian Point Nuclear Generating Unit Nos. 2 and 3, License Renewal Application,Set 2014-02 (TAC Nos. MD5407 and MD5408)," April 9, 2014. (ADAMS ML14094A173).2. Materials Reliability Program: Screening, Categorization, and Ranking of ReactorInternals Components for Westinghouse and Combustion Engineering PWR Design(MRP-191). EPRI, Palo Alto, CA: 2006. 1013234 (ADAMS ML12335A503).3. Materials Reliability Program: PWR Internals Material Aging Degradation MechanismScreening and Threshold Values (MRP-175). EPRI, Palo Alto. CA: 2005. 1012081(ADAMS ML12335A503).4. Entergy Letter, NL-14-013, "Additional Information Regarding the License RenewalApplication -Action Item 7 from MRP-227-A, Indian Point Nuclear Generating Unit Nos.2 & 3, Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64," January 28,2014 (ADAMS ML14038A150).
NL-14-093Attachment 1Page 5 of 65. Materials Reliability Program: Pressurized Water Reactor Internals Inspection andEvaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863 (ADAMSML1 20170453).6. Materials Reliability Program: Functionality Analysis for Westinghouse and CombustionEngineering Representative PWR Internals (MRP-230-Rev. 1).EPRI, Palo Alto, CA: 2009.1019091.7. Westinghouse Report, WCAP-17780-P, Rev. 0, "Reactor Internals Aging ManagementMRP-227-A Applicability for Combustion Engineering and Westinghouse PressurizedWater Reactor Designs," June 2013 (ADAMS ML13322A454). (Westinghouse ProprietaryClass 2).8. Westinghouse Calculation Note, CN-REA-1 3-38, Rev. 1, "Lower Core Support Fluence andGamma Heat Generation Rate Evaluation for Aging Management at Indian Point Units 2 &3," October 9, 2013. (Westinghouse Proprietary Class 2).
NL-14-093Attachment 1Page 6 of 6Table 1MRP-191 Screening Table for Westinghouse Reactor Internals [21Assembly Subassembly Component Material None SCC IASCC Wear Fatigue TE IE VS ISRIICControl Rod Guide Flanges -304 SS Weld FatUpper Internals Tube Assemblies LowerAssembly and Flow Flanges -Downcomers Lower CF8 SCC Fat TE IECore barrelCore Barrel Flange 3 SLower Lower Core 304SS Weld IASCC IEInternals Barrel (Weld) IAssembly Lower Support Lower Support CF8 IASCC TE IE VSColumn -Column BodiesAssemblies Lower Support 304 SS IASCC IE VS_ Notes:_ Column BodiesNotes:SCC = stress corrosion crackingVS = void swellingISR/IC = irradiation-induced stress relaxation and irradiation creep ATTACHMENT 2 TO NL-14-093LICENSE RENEWAL APPLICATIONIPEC LIST OF REGULATORY COMMITMENTSRev. 24ENTERGY NUCLEAR OPERATIONS, INC.INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3DOCKET NOS. 50-247 AND 50-286 NL-14-093Attachment 2Page 1 of 21List of Regulatory CommitmentsRev. 24The following table identifies those actions committed to by Entergy in this document.Changes are shown as strikethroughs for deletions 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 Complete A.3.1.1NL-13-122 B.1.1the bottom surfaces of the condensate storage tanks, IP3:city water tank, and fire water tanks once during the December 12,first ten years of the period of extended operation. 2015Enhance the Aboveground Steel Tanks Program forIP2 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 Complete A.3.1.2carifyB.1.2materials for low susceptibility to SCC and clarify the 1P3:prohibition on use of lubricants containing MoS2 for omplete NL-07-153 Audit Itemsbolting. 201,241,The Bolting Integrity Program manages loss of NL-1 3-122 270preload and loss of material for all external bolting. I I NL-14-093Attachment 2Page 2 of 21COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI I / IAUDIT 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.P2:-ompleteP3:December 12,2015NL-07-039NL-13-122NL-07-153NL-09-106NL-09-111A.2.1.5A.3.1.5B.1.6Audit Item173NL-1 1-101 NL-14-093Attachment 2Page 3 of 21COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI__ I_ 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:CompleteI P3:December 12,2015NL-07-039NL-13-122NL-07-153NL-08-057A.2.1.8A.3.1.8B.1.9Audit items128, 129,132,491,492,510 NL-14-093Attachment 2Page 4 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM5 Enhance the External Surfaces Monitoring Program lP2: NL-07-039 A.2.1.10for IP2 and IP3 to include periodic inspections of Complete A.3.1.10systems in scope and subject to aging management IP3:NL-13-122 B.1.11review for license renewal in accordance with 10 CFR DP e1December 12,54.4(a)(1) and (a)(3). Inspections shall include areassurrounding the subject systems to identify hazards tothose 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 IP2: NL-07-039 A.2.1.11monitor steady state cycles and feedwater cycles or omplete A.3.1.11NL-13-122 B.1.12,perform an evaluation to determine monitoring is not NL-03-122 Ad Itemrequired. Review the number of allowed events and 164resolve discrepancies between reference documentsand monitoring procedures. IP3:Enhance the Fatigue Monitoring Program for IP3 to December 12,include all the transients identified. Assure all fatigue 2015analysis transients are included with the lowestlimiting numbers. Update the number of designtransients accumulated to date.
NL-14-093Attachment 2Page 5 of 21COMMITMENT IMPLEMENTATION SOURCE I RELATEDSCHEDULE LRA SECTION4 I /I AUDIT ITEM7Enhance the Fire Protection Program to inspectexternal surfaces of the IP3 RCP oil collectionsystems for loss of material each refueling cycle.Enhance the Fire Protection Program to explicitlystate that the IP2 and IP3 diesel fire pump enginesub-systems (including the fuel supply line) shall beobserved 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.P2:,ompleteNL-07-039NL-13-122A.2.1.12A.3.1.12B.1.13P3:December 12,2015 NL-14-093Attachment 2Page 6 of 21COMMITMENT IMPLEMENTATION SOURCE I RELATEDSCHEDULE LRA SECTIONI I /I AUDIT ITEM8Enhance the Fire Water Program to include inspectionof IP2 and IP3 hose reels for evidence of corrosion.Acceptance criteria will be revised to verify nounacceptable signs of degradation.Enhance the Fire Water Program to replace all or testa sample of IP2 and IP3 sprinkler heads required for10 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 nosiqnificant corrosion.I P2:CompleteNL-07-039NL-13-122NL-07-153NL-08-014A.2.1.13A.3.1.13B.1.14Audit Items105, 106IP3:December 12,2015 NL-14-093Attachment 2Page 7 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION: I_ I I jI /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 pressure boundary.P2:ZompleteNL-07-039A.2.1.15A.3.1.15B.1.16NL-13-122IP3:December 12,2015 NL-14-093Attachment 2Page 8 of 21COMMITMENT IMPLEMENTATIONI SOURCE I RELATEDSCHEDULE LRA SECTIONI I /I AUDIT ITEM10 Enhance the Heat Exchanger Monitoring Program for IP2: NL-07-039 A.2.1.16IP2 and IP3 to include the following heat exchangers Complete A.3.1.16NL-13-122 B.1.17,in the scope of the program. IP3: NL-07-153 Audit Item* Safety injection pump lube oil heat exchangers December 12, 52* RHR heat exchangers 2015* 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, NL-09-018fouling, or scaling.NL-09-05611 Deleted NL-1 1-101 NL-14-093Attachment 2Page 9 of 21COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM12 Enhance the Masonry Wall Program for IP2 and IP3 IP2: NL-07-039 A.2.1.18to specify that the IP1 intake structure is included in Complete A.3.1.18the program. IP3: NL-13-122 8.1.19Complete13 Enhance the Metal-Enclosed Bus Inspection Program IP2: NL-07-039 A.2.1.19for IP2 and IP3 to visually inspect the external surface Complete A.3.1.19NL-13-122 8.1.20of MEB enclosure assemblies for loss of material at NL-13-122 Aui.tmleast once every 10 years. The first inspection willoccur prior to the period of extended operation and ecember 12, 124the acceptance criterion will be no significant loss of 015 NL-08-057 133,519material. NL-1 3-077Enhance 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.14 Implement the Non-EQ Bolted Cable Connections 1P2: NL-07-039 A.2.1.21Program for IP2 and IP3 as described in LRA Section Complete A.3.1.21NL-13-122 B.1.22B.1.22. IP3:December 12,2015 NL-14-093Attachment 2Page 10 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTION/ AUDIT ITEM15 Implement the Non-EQ Inaccessible Medium-Voltage lP2: NL-07-039 A.2.1.22Cable Program for IP2 and IP3 as described in LRA Complete A.3.1.22CeProg for.23. NL-13-122 B.1.23Section B.1.23. IP3: NL-07-153 Audit itemThis new program will be implemented consistent with December 12, 173the corresponding program described in NUREG- 2015 NL-11-0321801 Section XI.E3, Inaccessible Medium-VoltageCables Not Subject To 10 CFR 50.49 Environmental NL-1 1-096Qualification Requirements.NL-1 1-10116 Implement the Non-EQ Instrumentation Circuits Test P2: NL-07-039 A.2.1.23Review Program for IP2 and IP3 as described in LRA ýomplete A.3.1.23Section B.1.24. NL-13-122 B.1.24SP3: NL-07-153 Audit itemThis new program will be implemented consistent with December 12, 173the corresponding program described in NUREG- 20151801 Section XI.E2, Electrical Cables andConnections Not Subject to 10 CFR 50.49Environmental Qualification Requirements Used inInstrumentation Circuits.17 Implement the Non-EQ Insulated Cables and P2: NL-07-039 A.2.1.24Connections Program for IP2 and IP3 as described in 3omplete A.3.1.24LRA Section B.1.25. NL-13-122 B.1.25LP3: NL-07-153 Audit itemThis new program will be implemented consistent with December 12, 173the corresponding program described in NUREG- 20151801 Section XI.E1, Electrical Cables andConnections Not Subject to 10 CFR 50.49Environmental Qualification Requirements.
NL-14-093Attachment 2Page 11 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM18 Enhance the Oil Analysis Program for IP2 to sample 1P2: NL-07-039 A.2.1.25and analyze lubricating oil used in the SBO/Appendix Complete A.3.1.25NL-13-122 B.1.26R diesel generator consistent with the oil analysis for NL-11-101other site diesel generators. 3ecember 12,Enhance the Oil Analysis Program for IP2 and IP3 to 2015sample and analyze generator seal oil and turbinehydraulic 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.19 Implement the One-Time Inspection Program for IP2 P2: NL-07-039 A.2.1.26and IP3 as described in LRA Section B.1.27. omplete A.3.1.26NL-13-122 B.1.27This new program will be implemented consistent with IP3: NL-07-153 Audit itemthe corresponding program described in NUREG- ecember 12, 1731801, Section Xl.M32, One-Time Inspection. 01520 Implement the One-Time Inspection -Small Bore P2: NL-07-039 A.2.1.27Piping Program for IP2 and IP3 as described in LRA omplete A.3.1.27Secio B1.8.NL-13-122 B.1.28Section B.1.28. IP3: NL-07-153 Audit itemThis new program will be implemented consistent with December 12, 173the corresponding program described in NUREG- 20151801,Section XI.M35, One-Time Inspection of ASMECode 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 Complete A.3.1.28to assure that the effects of aging will be managed IP3:such that applicable components will continue toperform their intended functions consistent with the ee r1current licensing basis through the period of extendedoperation.
NL-14-093Attachment 2Page 12 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM22 Enhance the Reactor Vessel Surveillance Program for IP2: NL-07-039 A.2.1.31IP2 and IP3 revising the specimen capsule withdrawal Complete A.3.1.31NL-13-122 B.1.32schedules to draw and test a standby capsule to IP3:cover the peak reactor vessel fluence expected December 12,through the end of the period of extended operation. 2015Enhance the Reactor Vessel Surveillance Program forIP2 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. Complete A.3.1.32NL-13-122 B.1.33This new program will be implemented consistent with IP3: NL-07-153 Audit itemthe corresponding program described in NUREG- December 12, 1731801,Section XI.M33 Selective Leaching of Materials. 01524 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 omplete A.3.1.34monitoring assessment are compared to the P3:NL-13-122 B.1.35operational assessment performed for the prior Completeoperating cycle with differences evaluated.25 Enhance the Structures Monitoring Program to JP2: NL-07-039 A.2.1.35explicitly specify that the following structures are Complete A.3.1.35included in the program. NL-13-122 B.1.36" Appendix R diesel generator foundation (IP3) IP3: NL-07-153* Appendix R diesel generator fuel oil tank vault December 12, Audit items(IP3) 2015 86, 87, 88,* Appendix R diesel generator switchgear and NL-08-057 417enclosure (IP3)* city water storage tank foundation* condensate storage tanks foundation (IP3) NL-13-077" 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 (IP2)* new station security building (IP2)* nuclear service building (IP1)" primary water storage tank foundation (iP3).6 refueling water storage tank foundation (IP3)
NL-14-093Attachment 2Page 13 of 21COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONII I I I/ AUDIT ITEM" 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* monorails" new fuel storage racks* sumps NL-13-077Enhance the Structures Monitoring Program for IP2and IP3 to inspect inaccessible concrete areas thatare exposed by excavation for any reason. IP2 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 Monitorina Proaram for IP2Enhance the Structures Monitorina Proaram for 1P2 .- .- I I NL-14-093Attachment 2Page 14 of 21COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI_ I / AUDIT ITEMand 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.Enhance the Structures Monitoring Program toinclude more detailed quantitative acceptance criteriafor inspections of concrete structures in accordancewith ACI 349.3R, "Evaluation of Existing NuclearSafety-Related Concrete Structures" prior to theperiod of extended operation.NL-08-127NL-1 1-032NL- 11-101Audit Item360Audit Item35826 Implement the Thermal Aging Embrittlement of Cast I P2: NL-07-039 A.2.1.36Austenitic Stainless Steel (CASS) Program for IP2 omplete A.3.1.36NL-13-122 B.1.37and IP3 as described in LRA Section B.1.37. 1P3: NL-07-153 Audit itemThis new program will be implemented consistent with December 12, 173the corresponding program described in NUREG- 20151801,Section XI.M12, Thermal Aging Embrittlementof Cast Austenitic Stainless Steel (CASS) Program.
NL-14-093Attachment 2Page 15 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM27 Implement the Thermal Aging and Neutron Irradiation IP2: NL-07-039 A.2.1.37Embrittlement of Cast Austenitic Stainless Steel Complete A.3.1.37NL-13-122 B.1.38(CASS) Program for IP2 and IP3 as described in LRA IP3: NL-07-153 Audit itemSection B.1.38.Complete 173This new program will be implemented consistent withthe corresponding program described in NUREG-1801 Section XI.M13, Thermal Aging and NeutronEmbrittlement of Cast Austenitic Stainless Steel(CASS) Program.28 Enhance the Water Chemistry Control -Closed IP2: NL-07-039 A.2.1.39Cooling Water Program to maintain water chemistry of Complete A.3.1.39NL-13-122 B.1.40the IP2 SBO/Appendix R diesel generator cooling IP3: NL-08-057 Audit itemsystem per EPRI guidelines. Complete 509Enhance the Water Chemistry Control -ClosedCooling Water Program to maintain the IP2 and IP3security generator and fire protection diesel coolingwater pH and glycol within limits specified by EPRIguidelines.29 Enhance the Water Chemistry Control -Primary and IP2: NL-07-039 A.2.1.40Secondary Program for IP2 to test sulfates monthly in Complete B.1.41the RWST with a limit of <150 ppb. NL-13-122IP2: NL-07-039 A.2.1.4130 For aging management of the reactor vessel internals,IPEC will (1) participate in the industry programs for omplete A.3.1.41NL-13-122investigating and managing aging effects on reactor P3:internals; (2) evaluate and implement the results ofthe industry programs as applicable to the reactor ompleteinternals; and (3) upon completion of these programs,but not less than 24 months before entering the periodof extended operation, submit an inspection plan forreactor internals to the NRC for review and approval.1P2: NL-07-039 A.2.2.1.231 Additional P-T curves will be submitted as requiredper 10 CFR 50, Appendix G prior to the period of 3omplete A.3.2.1.2NL-13-122 4.2.3extended operation as part of the Reactor Vessel P3:Surveillance Program. 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.
NL-14-093Attachment 2Page 16 of 21COMMITMENT IMPLEMENTATION1 SOURCE RELATEDSCHEDULE LRA SECTIONII I / AUDIT ITEM33 At least 2 years prior to entering the period of IP2: NL-07-039 A.2.2.2.3extended operation, for the locations identified in LRA Complete A.3.2.2.3NL-13-122 4.3.3Table 4.3-13 (IP2) and LRA Table 4.3-14 (IP3), under NL-13-122 Adtiethe Fatigue Monitoring Program, IP2 and IP3 will p 146implement one or more of the following: Complete L-08-02 146(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 accounting NL-1 0-082for 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 theASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine avalid CUF.(2) Consistent with the Fatigue Monitoring Program,Corrective Actions, repair or replace the affectedlocations before exceeding a CUF of 1.0.34 IP2 SBO / Appendix R diesel generator will be NL-13-122 2.1.1.3.5installed and operational by April 30, 2008. This Complete NL-07-078committed change to the facility meets the NL-08-074requirements of 10 CFR 50.59(c)(1) and, therefore, alicense amendment pursuant to 10 CFR 50.90 is not NL-1 1-101required. NL -11-10 NL-14-093Attachment 2Page 17 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONSIAUDIT ITEM35 Perform a one-time inspection of representative IP2: NL-08-127 Audit Itemsample area of IP2 containment liner affected by the Complete 271973 event behind the insulation, prior to entering theperiod of extended operation, to assure linerdegradation is not occurring in this area. IP3: NL- 1-101Perform a one-time inspection of representative December 12,sample area of the IP3 containment steel liner at the 2015juncture with the concrete floor slab, prior to enteringthe 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.36 Perform a one-time inspection and evaluation of a IP2: NL-08-127 Audit Itemsample of potentially affected IP2 refueling cavity omplete NL-11-101 359concrete 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 will 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 of extended operation, a sample ofleakage fluid will be analyzed.37.I. P2: NL-08-127 Audit Item37 Enhance the Containment Inservice Inspection (Cl- , N0ompleteIWL) Program to include inspections of the NL-13-122containment using enhanced characterization of IP3:degradation (i.e., quantifying the dimensions of noted Cpltindications through the use of optical aids) during the ompleteperiod of extended operation. The enhancementincludes 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-14-093Attachment 2Page 18 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEMlP2: NL-08-143 4.2.138 For Reactor Vessel Fluence, should future coreloading patterns invalidate the basis for the projected Complete NL-13-122values of RTpts or CVUSE, updated calculations will lP3:be provided to the NRC. ecember 12,201539 Deleted NL-09-07940 Evaluate plant specific and appropriate industry 1P2: NL-09-106 B.1.6operating experience and incorporate lessons learned -omplete B.1.22NL-13-122 8.1.23in establishing appropriate monitoring and inspection lP3: B.1,24frequencies to assess aging effects for the new aging December 12, B.1.25management programs. Documentation of the 015 B.1.27operating experience evaluated for each new program B.1.28will be available on site for NRC review prior to the B.1.33period of extended operation. B.1.37I _B.1.381P2: NL-11-032 N/A41 IPEC will inspect steam generators for both units to fte theassess the condition of the divider plate assembly. egn tThe examination technique used will be capable of beginning of thedetecting PWSCC in the steam generator divider plate e'Em and prior toassembly. The IP2 steam generator divider plate S02 eNL1 1-074inspections will be completed within the first ten years 023of the period of extended operation (PEO). The IP3 P& NL-1 1-090steam generator divider plate inspections will becompleted within the first refueling outage following of the first NL-11-101the beginning of the PEO. refueling outageollowing theeginning of thePEQ.
NL-14-093Attachment 2Page 19 of 21COMMITMENT IMPLEMENTATION SOURCE I 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 generators.NL-1 1-032NL-1 1-074NL-1 1-090NL-1 1-096N/AI P2:Prior to March2024IP3: Prior to theand of the firstrefueling outagefollowing thebeginning of thePEO.IP2:Between March2020 and March20241P3: Prior to theend of the firstrefueling outagefollowing thebeginning of thePEO.43IPEC 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.P2:.ompleteP3: Prior toDecember 12,0115NL-1 1-032NL-13-122NL-1 1-1014.3.3 NL-14-093Attachment 2Page 20 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM44 IPEC will include written explanation and justification oP2: NL-11-032 N/Aof any user intervention in future evaluations using the omplete NL-1 1-101WESTEMS "Design CUF" module. lP3: Prior to NL-13-122December 12,201545 IPEC will not use the NB-3600 option of the oP2: NL-1 1-032 N/AWESTEMS program in future design calculations until omplete NL-1 1-101the issues identified during the NRC review of the IP3: Prior to NL-13-122program have been resolved. December 12,201546 Include in the IP2 ISI Program that IPEC will perform IP2: NL-1 1-032 N/Atwenty-five volumetric weld metal inspections of Completesocket welds during each 10-year ISI interval NL-11-074scheduled as specified by IWB-2412 of the ASME NL-13-122Section 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.47 IPEC" will peform and ,-,,it anay6ec that R21. NL 12 089 N/Ademoecr.ate that the loWe.r SWP199F column bod8eFlF t9Ima.intain their fuctionality during the period of ,L 13 522evtenlded operatn GQcnideFrig the pocibilIoiloss otfracture toughness due to thenrmal anRd irradoWiAtiRon ....... , 4 6embrittlement. The analyspcs will be consistent withthe 12!ID3 l.cen;ing basic. Deleted. NL-14-093 NL-14-093Attachment 2Page 21 of 21# COMMITMENT IMPLEMENTATION SOURCE RELATEDSCHEDULE LRA SECTIONI AUDIT ITEM48 Entergy will visually inspect IPEC underground piping oP2: NL-12-174 N/Awithin the scope of license renewal and subject to ompleteaging management review prior to the period of IP3: Prior to NL-13-122extended operation and then on a frequency of at ecember 12,least once every two years during the period ofextended operation. This inspection frequency will bemaintained unless the piping is subsequently coatedin accordance with the preventive actions specified inNUREG-1801 Section XI.M41 as modified by LR-ISG-2011-03. Visual inspections will be supplementedwith surface or volumetric non-destructive testing ifindications of significant loss of material areobserved. Consistent with revised NUREG-1 801Section XI.M41, such adverse indications will beentered into the plant corrective action program forevaluation of extent of condition and for determinationof appropriate corrective actions (e.g., increasedinspection frequency, repair, replacement).49 Recalculate each of the limiting CUFs provided in P2: NL-13-052 A.2.2.2section 4.3 of the LRA for the reactor vessel internals Complete A.3.2.2to include the reactor coolant environment effects IP3: Prior to NL-13-122(Fen) as provided in the IPEC Fatigue Monitoring December 12,Program using NUREG/CR-5704 or NUREG/CR- emr16909. In accordance with the corrective actionsspecified in the Fatigue Monitoring Program,corrective actions include further CUF re-analysis,and/or repair or replacement of the affectedcomponents prior to the CUFen reaching 1.0.50 Replace the IP2 split pins during the 2016 IP2: NL-13-122 A.2.1.41refueling outage (2R22). Prior to B. 1.42completion of NL-14-0672R22IP3: N/A