ML16005A093

South Texas Project, Units 1 and 2 - Response to Request for Additional Information Set 34 for the Review of the License Renewal Application

ML16005A093
Person / Time
Site:	South Texas
Issue date:	12/17/2015
From:	Powell G T South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16005A093	List: ML16005A093 NOC-AE-15003320, Response to Request for Additional Information Set 34 for the Review of the License Renewal Application
References
NOC-AE-15003320, STI: 34253535, TAC ME4936, TAC ME4937
Download: ML16005A093 (34)

Text

Nuclear Operating CompanySouth Te_-as Project Electric GeneratingStatobn P.O Box 289° Wadsworth, Te.as 77483 &A A -December 17, 2015NOC-AE-1 500332010 CER 54File No. G25U. S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555-0001South Texas ProjectUnits 1 and 2Docket Nos. STN 50-498, STN 50-499Response to Request for Additional Information Set 34 for theReview of the South Texas Project, Units 1 and 2,License Renewal A~ppication (TAC Nos. ME4936 and ME4937)References:1. Letter; G. T. Powell to USNRC Document Control Desk; "License Renewal Application;"NOC-AE- 10002607; dated October 25, 2010 (ML1 03010257)2. Letter; J. W. Daily to G. T. Powell; "Request for Additional Information Set 34 for theReview of the South Texas Project, Units 1 and 2, License Renewal Application (TAONos. ME4936 and ME4937);" AE-NOC-1 5002760; dated November 30, 2015(ML1 5308A01 4)By Reference 1, STP Nuclear Operating Company (STPNOC) submitted a License RenewalApplication (LRA). By Reference 2, the NRC staff requested additional information (RAI) fortheir review of the STPNOC LRA. The RAl's contains areas where additional information isneeded to complete the review of the license renewal application. STPNOC's response to theRAI's are provided in Enclosure 1 to this letter.Changes to LRA pages described in Enclosure 1 are depicted as line-in/line-out pages providedin Enclosure 2.In addition, Enclosures 3 and 4 contain supporting documentation referenced in this submittal.Regulatory commitment for item 30 in LRA Table A4-1 is revised and depicted as line-in/line-outpages provided in Enclosure 5. There are no other commitments in this letter.STI: 34253535 NOC-AE- 15003320Page 2 of 3If there are any questions, please contact Arden Aidridge, STP License Renewal Project Lead,at (361) 972-8243 or Rafael Gonzales, STP License Renewal Project regulatory point-of-contact, at (361) 972-4779.I declare under penalty of perjury that the foregoing is true and correct.Executed on ____-_________DateG. T. PowellSite Vice PresidentrjgEnclosures:1. STPNOC Response to RAI2. STPNOC LRA Changes with Line-in/Line-out Annotations3. PWROG-1 4072-NP, Rev. 0, "South Texas Project Units 1 and 2 Summary Report forthe Fuel Design / Fuel Management Assessments to Demonstrate MRP-227-AApplicability," June 3, 2015.4. PWROG-1 5001-NP, Rev. 0, "South Texas Project Unit 1 and Unit 2 Summary Report forApplicant/LicenseeAction Items 1, 2, and 7," June 2015.5. STPNOC Regulatory Commitment NOC-AE-1 5003320Page 3 of 3CC:(paper copy)(electronic copy)Regional Administrator, Region IVU.S. Nuclear Regulatory Commission1600 East Lamar BoulevardArlington, TX 76011-4511Lisa M. RegnerSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North (MS 8 G9A)11555 Rockville PikeRockville, MD 20852NRC Resident InspectorU. S. Nuclear Regulatory CommissionP. O. Box 289, Mail Code: MNl16Wadsworth, TX 77483John W. DailyLicense Renewal Project Manager (Safety)U.S. Nuclear Regulatory CommissionOne White Flint North (MS 011-Fl)Washington, DC 20555-0001Tam TranLicense Renewal Project Manager(Environmental)U. S. Nuclear Regulatory CommissionOne White Flint North (MS 011 F01)Washington, DC 20555-0001Mor~qan, Lewis & Bockius LLPSteve FrantzU.S. Nuclear Requlatory CommissionLisa M. RegnerJohn W. DailyNRG South Texas LPJohn RaganChris O'HaraJim von SuskilCPS EnerqiyKevin PolioCris EugsterL. D. BlaylockCramn Caton & James, P.C.Peter NemethCity of AustinCheryl MeleJohn WesterTexas Dept. of State Health ServicesRichard A. RatliffRobert Free NOC-AE- 15003320Enclosure 1Enclosure 1STPNOC Response to RAI NOC-AE-1 5003320Enclosure 1Page 1 of 13RAI 3.0.3.3.6-1 -Components within the scope of the AMPBack~qround:By letter dated June 30, 2015, the applicant submitted its updated version of the plant-specificPWR Reactor Internals Program (LRA Section B2.1 .35) for staff review. In the "scope ofprogram" program element for the aging management program (AMP), the applicant identifiesthat the program includes the following types of components defined for Westinghouse-designedPWRs in the MRP-227-A report: (a) "Primary" category components, (b) "Expansion" categorycomponents, and (c) "Existing Program" componentsIssue:The population of components in MRP-227-A includes "Primary," "Expansion," "ExistingProgram" and "No Additional Measures" category components, even though "No AdditionalMeasures" components are not included as part of the sample of components that will beinspected in accordance with the MRP-227-A methodology. The "Scope of Program" programelement for the PWR Reactor Internals Program does not include "No Additional Measures"components as part of the population of components that is included within the scope of theAMP. The methodology in MRP-227-A does not preclude the possibility that the samecomponents identified as "No Additional Measures" components in MRP-227-A are ASMESection XI Examination Category B-N-2 or B-N-3 components for the STP units.Request:Justify the basis for omitting "No Additional Measures" components from the scope andpopulation of components in the PWR Reactor Internals Program. Clarify whether any of the RVI"No Additional Measures" components at STP are defined as ASME Section XI ExaminationCategory B-N-2 or B-N-3 components. If so, identify which "No Additional Measures"components are within the scope of the ASME Section XI Examination Category B-N-2 or B-N-3requirements, and clarify whether the components will be inspected in accordance with theASME Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD" Program (LRA AMPB2.1.1).STPNOC Response:License Renewal Application (LRA) Appendix B2.1 .35 and License Renewal (LR) BasisDocument Aging Management Program (AMP) Pressurized Water Reactor Internals (PWRI),Pressurized Water Reactor (PWR) Internals program, have been updated to identify a fourthgroup consisting of those PWR internals components for which the effects of all eight agingmechanisms are below the screening criteria were placed in the No Additional Measures group.No further action is required for managing the aging of the No Additional Measures components.The No Additional Measures components are not core support structures, and therefore are notcovered by an AMP element such as the ASME Section XI Examination Category B-N-2 or B-N-3 requirements. These No Additional Measures components are show in the Aging ManagementReview (AMR) Table 3.1.1 Enclosure 2 of STP letter dated June 30, 2015 pages 33-35.Enclosure 2 provides LIRA Changes with Line-in/Line-out Annotations for B2.1 .35 NOC-AE-1 5003320Enclosure 1Page 2 of 13RAI 3.0.3.3.6-2 -Apparent component categorization inconsistenciesBackqiround:The background information in RAI 3.0.3.3.6-1 apply to this RAI. In addition, the "scope ofprogram" program element in the PWR Reactor Internals Program identifies that the scope of theAMP includes the XL lower core plate as an "Expansion" category component for the AMP.Issue:In MRP-227-A, the EPRI MRP identifies XL lower core plates in Westinghouse-designed PWRas "Existing Program" components that are inspected in accordance with ASME Section XlExamination Category B-N-3 requirements and does not define these components as"Expansion" components. To be consistent with this protocol, the reactor vessel internalsinspection plan (RVIIP), as submitted in the letter of June 30, 2015, identifies that the XL lowercore plates are "Existing Program" components that will be inspected in accordance with ASMESection Xl Examination Category B-N-3 requirements. Thus, there is an apparent inconsistencybetween the category identified for the XL lower core plates in the "Scope of Program" elementand the category for these components identified in the RVIIP.Request:Clarify whether the XL lower core plates (one plate in each unit) are "Expansion" components or"Existing Program" components for the PWR Reactor Internals Program, or both. If the platesare "Expansion" components, identify and justify the basis for selecting the "Primary"components that are linked to the XL lower core plates as "Expansion" components for the AMPand the RVIIP.STPNOC Response:The XL lower core plates are Existing Program components for the PWR Reactor InternalsProgram.LRA Appendix B2.1 .35 and LR Basis Document AMP PWRI, PWR Reactor Internals programScope of Program -Element 1 have been revised to show the XL lower core plate as an"Existing" category component.Enclosure 2 provides LRA Changes with Line-in/Line-out Annotations for B2.1 .35 NOC-AE-1 5003320Enclosure 1Page 3 of 13RAI 3.0.3.3.6-3 -Response to A/LAI #1 -Lack of an MRP Letter 2013-025 AssessmentBackgqround:In the applicant's letter of June 30, 2015, the applicant provides its response toApplicant/Licensee Action Item (A/LAI) #1 on the MRP-227-A report, which requested theapplicant to provide adequate demonstration that the assumptions for establishing the criteria inMRP-227-A are bounding the design of the RVI components at the applicant's facility. The EPRIMRP developed the criteria in EPRI MRP Letter No. 201 3-025 to assist applicants ofWestinghouse-designed PWRs in addressing the A/LAI request. In this letter the EPRI MRPrecommended that applicants owning Westinghouse designed PWRs should provide theirassessments of the following parameters:* Demonstrate that the distance between the top of the active fuel and the upper core plate isgreater than 12.2 inches* Demonstrate that the average core power density is less than 124 watts/cm3* Demonstrate that the heat generation figure of merit, F, is less than or equal to 68 watts/cm3In the letter to the EPRI MRP, the staff agreed that demonstration of conformance with theacceptance criteria for these plant parameters would serve as a valid basis for concluding thatthe assumptions used in MRP-227-A are bounding for the design of the RVI at their facilities.Issue:The letter of June 30, 2015, does not include an assessment of the parameters listed above, asrecommended in MRP Letter No. 201 3-025.Reqiuest:Provide the basis why the response basis to A/LAI #1 in the letter of June 30, 2015, did notinclude an assessment of the three parameters listed above, as recommended in EPRI MRPLetter 2013-025. Justify why such an assessment would not be needed as part of the basis forconcluding that the assumptions used to develop MRP-227-A are bounding for the design of theRVI components at STP Units 1 and 2.STP Response:The three parameters in RAI-3.0.3.3.6-3 and EPRI Letter MRP 201 3-025, "MRP-227-AApplicability Template Guideline," October 14, 2013 [ML1 3322A454] are addressed for STP Unit1 and Unit 2 within PWROG-1 4072-NP, Rev. 0, "South Texas Project Units 1 and 2 SummaryReport for the Fuel Design/Fuel Management Assessments to DemonstrateMRP-227-A Applicability," June 3, 2015, and it is demonstrated that the parameters are met. Acopy of this document has been provided in Enclosure 3 of this submittal.The STP Unit 1 assessment concluded that the nominal distance between the top of the activefuel and bottom of the upper core plate, averaged over the first 19 fuel cycles of operation, wasnot less than 12.2 inches. STP Unit 1 has been operating at a rated power level of 3853 MWt forthe last eight operating fuel cycles (Cycles 12 through 19). For the 193 fuel assembly coregeometry of STP Unit 1, this corresponds to a core power density of 101.2 W/cm3.It was alsodemonstrated that, considering the entire operating lifetime of the reactor, the average powerdensity of the core shall be less than 124 W/cm3 for a period of more than two effective full-power years. For the last eight operating fuel cycles, STP Unit 1 has kept its heat generationfigure of merit range under 68 W/cm3, and this range is representative of anticipated futureoperation.

NOC-AE-1 5003320Enclosure 1Page 4 of 13The STP Unit 2 assessment concluded that the nominal distance between the top of the activefuel and bottom of the upper core plate, averaged over the first 17 fuel cycles of operation, wasnot less than 12.2 inches. STP Unit 2 has been operating at a rated power level of 3853 MWt forthe last eight operating fuel cycles (Cycles 10 through 17). For the 193 fuel assembly coregeometry of STP Unit 2, this corresponds to a core power density of 101.2 W/cm3.It was alsodemonstrated that, considering the entire operating lifetime of the reactor, the average powerdensity of the core shall be less than 124 W/cm3 for a period of more than two effective full-power years. For the last eight operating fuel cycles, STP Unit 2 has kept its heat generationfigure of merit range under 68 W/cm3, and this range is representative of anticipated futureoperation.Enclosure 3 contains a copy of PWROG-1 4072-NP, Rev. 0, "South Texas Project Units 1 and 2Summary Report for the Fuel Design/Fuel Management Assessments to DemonstrateMRP-227-A Applicability," June 3, 2015.,RAI 3.0.3.3.6-4- Response to AILAI #2 -Comparison to UFSAR InformationBackgqround:The background information in RAl 3.0.3.3.6-1 apply to this RAI. In the applicant's response toA/LAI #2, the applicant states that the generic scoping and screening of the RVI, assummarized in the MRP-191 and MRP-232 reports (in order to support the inspection criteria inMRP-227-A), are applicable to STP Units 1 and 2 with no modifications for the components.The applicant states that the RVI components in the units are in conformance with theaugmented inspection criteria in MRP-227-A for all components and that the protocols in MRP-227-A do not need to be modified under the criteria in A/LAI #2.Issue:In Section 4.1 of the updated final safety analysis report (UJFSAR), the applicant identifies RVIdesign assembly or component modifications that have been or will be implemented in the units.Based on the UFSAR statements, the staff need to understand: (a) whether the specific RVIassemblies at STP include any design configurations that deviate from the RVI designassemblies and assembly components that were generically evaluated in the MRP-1 91, MRP-232, and MRP-227-A reports or were not evaluated in these reports, and (b) whether thesedeviations (if they exist) should have been more definitively assessed in the response that wasprovided to A/LAI #2. Apparent deviations for lower core support structure components areaddressed in RAl 3.0.3.3.6-5.Request:Identify all RVI design assembly component configurations (other than those for the deviationson lower core support structure assembly components) that have not been evaluated by or differfrom those generically evaluated in the MRP-191, MRP-232, and MRP-227-A reports, other thanthose for lower core support assembly components (which are the topic of RAI 3.0.3.3.6-5). Forcomponents that have corresponding components in the generic MRP evaluations but differ fromthe configurations in the generic evaluation, clarify how the stress levels and neutron fluences forthese components compare to those assessed for corresponding components in the genericMRP design evaluations. Based on this comparison, justify why augmented inspection protocolsfor the components would not need to be proposed for the components on a plant-specific basisfor the AMP. Similarly, for components not analyzed in the MRP reports, justify why plant-NOC-AE-1 5003320Enclosure 1Page 5 of 13specific aging management criteria would not need to be proposed for the components on aplant-specific basis for the AMP.STPNOC Response:The STP Unit 1 and Unit 2 reactor internals plant design was considered within the creation ofMRP-1 91, Materials Reliability Program: Screening, Categorization, and Ranking of ReactorInternals Components for Westinghouse and Combustion Engineering PWR Design (MRP-19 1).EPRI, Palo Alto, CA: 2006, as shown in Table 4-2 of MRP-1 91. Therefore, all RVI designassembly component configurations for STP Unit 1 and 2 were accounted for within MRP-1 91,Table 4-4, as concluded in PWROG Report, PWROG-1 5001-NP, Rev. 0, "South Texas ProjectUnit 1 and Unit 2 Summary Report for Applicant/Licensee Action Items 1, 2, and 7," June 2015.STP Unit 1 and 2 comply with A/LAI 2 of the NRC SE in MRP-227-A for all components.Therefore, STP Unit 1 and Unit 2 meet the requirement for application of MRP-227-A as astrategy for managing age-related material degradation in reactor internals components.Enclosure 4 contains a copy of PWROG Report, PWROG-1 5001-NP, Rev. 0, "South TexasProject Unit 1 and Unit 2 Summary Report for Applicant/Licensee Action Items 1, 2, and 7," June2015.RAI 3.0.3.3.6-5 -Response to AILAI #2 -Lower Core Support Assembly DeviationsBackqround:The background information in RAl 3.0.3.3.6-4 apply to this RAl. Additionally, the "scope ofprogram" program element for the PWR Reactor Internals Program and the tables of the ReactorVessel Internals Inspection Plan (RVIIP) identify that the design of the RVI assemblies at STP donot include: (a) lower core support assemblies, (b) lower core support column bodies, or (c)lower core support column bolts. The lower core support column bodies and column bolts aredefined as "Expansion" components in the MRP-227-A report.Issue:These deviations were not identified in the response to A/LAI #2 and change a number ofgeneric "Primary" to "Expansion" category relationships for the RVIIP from those defined in theMRP-227-A report for these Westinghouse-designed internals.Request:Justify why the response to A/LAI #2 has not identified the lack of a lower core support structureassembly and lower core support column bodies and bolts (MRP-227-A "Expansion"components) as a deviations from the assessments in the MRP-191, MRP-232, and MRP-227-Areports. Clarify how these deviations would change the "Primary" to "Expansion" categoryrelationships that need to be defined for the AMP and RVlIP when compared to those normallydefined in the MRP-227-A report for Westinghouse-designed internals. Provide the basis whyalternative "Expansion" component substitutions for these components would not need to beproposed for the AMP and RVIIP in order to be consistent with the total number of "Expansion"components defined in MRP-227-A for Westinghouse-designed internals.

NOC-AE-1 5003320Enclosure 1Page 6 of 13STPNOC Response:The STP Unit 1 and Unit 2 reactor internals plant designs were considered within the creation ofMRP-1 91, Materials Reliability Program: Screening, Categorization, and Ranking of ReactorInternals Components for Westinghouse and Combustion Engineering PWR Design (MRP-19 1).EPRI, Palo Alto, CA: 2006, as shown in Table 4-2 of MRP-1 91. Therefore, the lower core supportstructure is accounted for by the XL lower core plate within MRP-1 91. There are no alternatecomponent substitutions identified within MRP-227-A for STP Unit 1 and 2 and both units complywith A/LAI 2 of the NRC SE in MRP-227-A for all components. Therefore, STP Unit 1 and Unit 2meet the requirement for application of MRP-227-A as a strategy for managing age-relatedmaterial degradation in reactor internals components.RAI 3.0.3.3.6-6 -Topic -Response to A/LAI #3 -Use of Inspection Data for CRGT Split PinsBackqround:In the applicant's letter of June 30, 2015, the applicant provides its response to A/LAI #3 on theMRP-227-A report, which requested that the applicant assess the need to replace or performaugmented inspections of their control rod guide tube (CRGT) support pins (split pins). Theapplicant's basis for resolving A/LAI #3 and for concluding that augmented inspections of thereplaced CRGT split pins are not currently needed relies, in part, on the applicant's statementthat data from industry inspections of replaced CRGT split pins made from Type 316 cold-workedstainless steel will be obtained from other U.S. (or foreign) licensees, the EPRI MRP, or otherindustry organizations and will be used to assess the need for developing augmented inspectioncriteria of the CRGT split pins at STP Units I and 2.Issue:1. The EPRI MRP has yet to identify in MRP-227-A or in the background reports for MRP-227-A that augmented inspections are part of the programmatic criteria for managingcracking or wear in replaced Westinghouse-design CRGT split pins made from Type 316cold-worked stainless steel materials or that such data will be collected by the EPRI MRPfor distribution to and evaluation by the industry licensee. Thus, some additionalinformation is needed to clarify how the applicant will implement its process for collectingand assessing CRGT split pin inspection data in accordance with the PWR ReactorInternals Program.2. If the CRGT splits pins are defined as ASME Section Xl Examination Category B-N-3removable core support structure components, the applicant will be required to inspect thecomponents in accordance with their ISI program requirements for B-N-3 inspections,independent of the position taken in MRP-227-A for replaced split pins made from Type316 cold-worked stainless steel materials.Request:1. Identify the plants that will be performing inspections of their replaced Type 316 cold-worked CRGT split pins which the applicant will use as the lead operating experience formanaging aging in the CRGT split pins at STP Units I and 2. Identify the process orprocesses that will be used in accordance with the "Administrative Controls" or"Confirmation Process" elements of the PWR Reactor Internals Program to collect andcompile the inspection data from these plants. Identify the criteria that will be implementedin accordance with the "monitoring and trending" program element of the AMP. Identify the NOC-AE-1 5003320Enclosure 1Page 7 of 13plant-specific "acceptance criteria" that will be used to assess such data and the "correctiveactions" that will be taken if the acceptance criteria are not met.2. Clarify whether the replaced CRGT split pins at STP are categorized as ASME Section XlExamination Category B-N-3 components (i.e. ASME removable core support structurecomponents). If the split pins are defined as ASME removable core support structurecomponents, justify why the components would not need to be inspected and managed foraging using either the "Existing Program" criteria in the PWR Reactor Internals Program(LRA B.2.1 .35) or the ASME Section XI Inservice Inspection, Subsections IWB, IWC, andIWD Program (i.e., the ISI Program in LRA Section B2.1.1).STPNOC Response:STP follows operating experience (OE) and adjusts the program if needed, based on OE. Theindustry as a whole continues to share OE through the auspices of the MRP and PWROG. Anynew found OE-would be used by STPNOC to assess the need for developing augmentedinspections for the split pins.MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection andEvaluation Guidelines (MRP-22 7-A). EPRI, Palo Alto, CA: 2011, subsection 4.4.3, guidance forguide tube support pins (split pins) in Westinghouse plants is limited to plant-specificrecommendations. The owner is directed to review and follow the original equipmentmanufacturer (OEM) recommendations for aging management and subsequent performancemonitoring. Results of the detailed categorization and ranking of internals components containedin MRP-227-A, Table 3-3, identify only X-750 split pins as requiring specific actions to managematerial aging in the period of extended operation (PEO); thus, no inspection or monitoring of the316 stainless steel (SS) variant for control rod guide tube (CRGT) support pins is included inMRP-227-A, Table 4-9. As described In response to requests for additional information Set 28,dated June 30, 2015, RAI B2.1.35-1, STP Units I and 2 followed the OEM recommendation toreplace the originally installed X-750 CRGT support pins with support pins fabricated from strain-hardened 316 SS material.As listed in MRP-1 91, Materials Reliability Program: Screening, Categorization, and Ranking ofReactor Internals Components for Westinghouse and Combustion Engineering PWR Design(MRP-19 1). EPRI, Palo Alto, CA: 2006, Table 5-1, the 316 SS guide tube support pins werescreened in for the aging degradation mechanisms of wear, fatigue, and irradiation stressrelaxation/irradiation creep (ISRIIC). In MRP-232, Materials Reliability Program: AgingManagement Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232, Revision 1). EPRI, Palo Alto, CA: 2012, the 316 S5 CRGT support pins were categorizedas MRP-191 Category A, "no additional measures." The OEM recommendations do not requiresubsequent inspection of the 316 SS support pins. Long-term material behavior has beenextensively studied from past testing and field experience; all identified degradation mechanisms,including irradiation-assisted stress corrosion cracking (IASCC), SCC, wear, fatigue, ISR/IC andembrittlement, have been assessed. Westinghouse Report, WCAP-1 5028-NP, Rev. 1, "GuideTube Cold-Worked 316 Replacement Support Pin Development Program," June 24, 2011,concluded that the 316 SS CRGT support pins will perform all intended functions for thedesignated PEO with no requirement for post-installation inspections. STPNOC agingmanagement programs comply with the OEM recommendations and MRP-227-A adequacyevaluation requirements for aging management of reactor internals CRGT 316 SS support pins.

NOC-AE-1 5003320Enclosure 1Page 8 of 13The CRGT split pins are defined as ASME Section Xl, Examination Category B-N-3 componentsat South Texas Project (STP) Units 1 and 2 as part of the upper internals assembly. MRP-227-Aspecifically states that there are no impacts to the plant-specific ASME Section Xl program butthat only specific items that may be part of the program were credited as part of the internalsaging management inspection requirements [Inservice Inspection Program Plan for the SouthTexas Project Electric Generating Stations Units 1 & 2 Commercial Operation Dates Unit 1August 25, 1988 Unit 2 June 19, 1989 Effective 10/08/20121. There is no impact to the ASMESection XI inspection requirements as a result of MRP-227-A implementation at STP Units 1 and2. STP Units 1 and 2 have therefore demonstrated that aging for split pins are adequatelymanaged during the period of extended operation.RAI 3.0.3.3.6-8 -Response to A/LAI #7 -Thermal Aging of CASS Upper InternalsBackqirou nd:In the applicant's letter of June 30, 2015, the applicant provides its response to A/LAI #7 on theMRP-227-A report, which addressed the issue of thermal aging embrittlement and neutronirradiation embrittlement in RVI components made from cast austenitic stainless steel (CASS).Issue:The response to AILAI #7 uses the criteria in NRC License Renewal Issue 08-0030 (dated May19, 2000) as the basis for concluding that thermal aging embrittlement will not be an agingmanagement issue for RVI upper internals assembly support columns or column bases.Additional data is necessary to verify that thermal aging embrittlement will not be an agingmechanism of concern for these components during the period of extended operation.Request:Provide the plant-specific delta-ferrite contents for the CASS CF8 materials used to fabricateupper internals assembly support columns or column bases, and the equational criteria and plantspecific chemistry alloy content data used to calculate the delta-ferrite contents of thesecomponents. As an alternative basis for resolving this issue (if applicable), the applicant maydemonstrate that these components were appropriately evaluated in MRP-227-A or thebackground reports for MRP-227-A and were placed into FMECA Category A and "No AdditionalMeasures" categories based on the conclusions that there are no consequences on RVIcomponent intended functions if these components fail to maintain their structural integrity.STPNOC Response:With the exception of the upper internal support columns, CASS CF8 materials were used tofabricate the upper internal assembly as outlined in PWROG Report, PWROG-1 5001-NP, Rev.0, "~South Texas Project Unit 1 and Unit 2 Summary Report for Applicant/Licensee Action Items1, 2, and 7," June 2015.STP Unit 1 has fifty upper internals assembly -upper support columns, column bases that arecomprised of CASS Grade CF8 material. Forty-eight of the fifty column bases are not susceptibleto thermal embrittlement since they have a ferrite content less than or equal to 20 percent basedon certified material test report (CMTR) data. Two of the fifty column bases have the potential tobe susceptible to thermal embrittlement since CMTRs were not identified, and a conservativeapproach was taken. In MRP-1 91, Materials Reliability Program: Screening, Categorization, andRanking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR NOC-AE-1 5003320Enclosure 1Page 9 of 13Design (MRP-19 1). EPRI, Palo Alto, CA: 2006, Table 5-1, the upper support column, columnbases CF8 are screened in for the material degradation effects of stress corrosion cracking(SCC), thermal embrittlement (TE) and irradiation embrittlement (IE). Taking into considerationall of the material degradation mechanisms, including IE, the upper support column basecomponent was ranked in MRP-227-A as a "No Additional Measures" category component.Based on these results, the continued application of the MRP-227-A strategy for STP Unit 1meets the requirement for managing age-related degradation of the STP Unit 1 CASS reactorvessel internals components.STP Unit 2 has fifty upper internals assembly -upper support columns, column bases that arecomprised of CASS Grade CF8 material. All fifty column bases are not susceptible to thermalembrittlement since they have a ferrite content less than or equal to 20 percent based on CMTRdata. In MRP-1 91, Materials Reliability Program: Screening, Categorization, and Ranking ofReactor Internals Components for Westinghouse and Combustion Engineering PWR Design(MRP-19 1). EPRI, Palo Alto, CA: 2006, Table 5-1, the upper support column, column bases CF8are screened in for the material degradation effects of SCC, TE, and IE. Taking intoconsideration all of the material degradation mechanisms, including IE, the upper support columnbase component was ranked in MRP-227-A as a "No Additional Measures" component. Basedon these results, the continued application of the MRP-227-A strategy for STP Unit 2 meets therequirement for managing age-related degradation of the STP Unit 2 CASS reactor vesselinternals components.Enclosure 4 contains a copy of PWROG-1 5001-NP, Rev. 0, "South Texas Project Unit 1 and Unit2 Summary Report for Applicant/Licensee Action Items 1, 2, and 7," June 2015.3.0.3.3.6-9 -Response to AILAI #8, Item 5 -RVI Environmentally-Assisted FatigueBackgqround:In the applicant's letter of June 30, 2015, the applicant provides its response to A/LAI #8, Item 5,on the MRP-227-A report, which addresses the bases that will be used to manage or adequatelymanage environmentally-assisted fatigue in PWR RVI components. In the applicant's response,the applicant identifies that the metal fatigue TLAAs have been included and evaluated in LRASection 4.3.3 and that the PWR Reactor Internals Program will not be used as the basis formanaging cracking induced by environmentally-assisted fatigue during the period of extendedoperation.Issue:Although the scope of LRA AMP B3.1 includes activities to monitor the impacts ofenvironmentally-assisted fatigue on the CUE analyses for reactor coolant pressure boundarycomponents, it is not evident whether similar activities will be applied to the CUF analyses for theRVl components listed in the background section of this RAI, and if so, how such activities will beapplied to the cycle counting and CUE reanalysis criteria defined in the AMP.Request:Clarify whether the AMP's monitoring and trending activities for monitoring the impacts ofenvironmental effects of the adequacy of components with CUE analyses are being extended tothose RVI components with a CUE analysis. If not, identify the activities that will be performed toanalyze or manage environmentally-assisted fatigue in the RVI components. Justify theresponse to this RAI.

NOC-AE-1 5003320Enclosure 1Page 10 of 13STPNOC Response:LIRA AMP B3.1 has been revised to include activities to monitor the impacts of environmentally-assisted fatigue on the locations with fatigue usage calculations.Enclosure 2 provides LIRA Changes with Line-in/Line-out Annotations for LIRA section B3.1.Enclosure 4 provides LIRA Changes with Line-in/Line-out Annotations for LIRA Table A4.1"STPNOC Regulatory Commitments"RAI 3.0.3.3.6-10 -Adequacy of UFSAR Supplement Section A1.35Backqround:The current UFSAR supplement summary description for the PWR Reactor Vessel InternalsProgram is given in Section Al1.35 of the LIRA Appendix A. By letter dated June 30, 2015, theapplicant submitted its updated version of the plant-specific PWR Reactor Internals Program(LRA Section B2.1 .35) for staff review in order to respond to the staff's request in RAl B2.1 .35-1.The updated version of the AMP provided in the June 30, 2015, letter updates the programelement criteria for the AMP to be consistent with those provided in EPRRI Technical Report No.1022863, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection andEvaluation Guideline (MRP-227-A)," which was formally issued by the EPRI MRP in January2012.Issue:In the letter of June 30, 2015, the applicant did not administratively update LIRA Section Al1.35,PWR Reactor Internals, to be consistent with the updated version of the PWR Reactor InternalsProgram (LRA Section B2.1.35) provided in the letter of June 20, 2015. Thus, the currentversion of LIRA UFSAR Supplement Section Al1.35, "PWR Vessel Internals," is out of date andmust be updated to reflect the status of the AMP and reactor vessel internals inspection plan(RVIIP) that were submitted in the letter of June 30, 2015.Request:Justify why LIRA Section Al1.35 has not been updated to reflect that the current status of the AMPand RVIIP submitted in the letter of June 30, 2015. Specifically, justify why the USFARsupplement in Section Al1.35 has not been updated to reflect the follow aspects of the program:* Appropriate referenced ERPI Report for the AMP and UFSAR Supplement for the AMP isEPRI Technical Report No. 1022863, "Materials Reliability Program: Pressurized WaterReactor Internals Inspection and Evaluation Guideline (MRP-227-A)"*Protocols and activities for implementing the AMP and RVIIP in accordance withmethodology in MRP-227-A are appropriately adjusted to account for deviations from thegeneric design and inspection and evaluation criteria in MRP-227-A or for the applicant'sresponse bases for resolving specific Applicant/Licensee Action Items in theMRP-227-A report, as identified in the NRC safety evaluation for MRP-227-A datedDecember 16, 2011 NOC-AE-1 5003320Enclosure 1Page 11 of 13*Population of components in the AMP include "Primary," "Expansion," "Existing Program,"and "No Additional Measures" category components for the AMPSTPNOC Response:STP letter dated February 27, 2012 provided a revised Appendix A Section Al1.35 which updatedthe appropriate referenced ERPI Report for the AMP and UFSAR Supplement to EPRI TechnicalReport No. 1022863, "Materials Reliability Program: Pressurized Water Reactor InternalsInspection and Evaluation Guideline (MRP-227-A)" and identified the protocols and activities forimplementing the AMP and RVIIP in accordance with methodology in MRP-227-A.LRA Appendix C outlines STPNOC's response to Applicant/Licensee Action Items.Appendix B Section B2.1.35 outlines the population of components in the AMP include "Primary,""Expansion," "Existing Program," and "No Additional Measures" category components for theAMP.RAI B2.1.13-5a -LR-ISG-2013-O1 Inspection Frequency FollowupBackground:1. The applicant's response to RAl 3.0.3-2a Part (d) dated June 11, 2015, states thefollowing, in part:When visual inspections detect any blistering, cracking, erosion, cavitation erosion,flaking, peeling, delamination, rusting and physical damage the coating is considereddegraded. Degraded coatings are removed to sound material and replaced with newcoating. The as-found degraded condition is documented in the corrective action programfor trending. The NCS oversees the replacement of the degraded coatings assuring theextent of repaired or replaced coatings encompasses sound coating material. Review ofSTP's existing coating inspection program operating history demonstrates that theremediation of degraded coating conditions prior to returning the coating back in serviceis effective in managing the coating performance from one inspection to the next, with nochange in inspection interval.2. In regard to followup testing conducted to ensure that the extent of repaired or replacedcoatings encompasses sound coating material, the response to RAI 3.0.3-2a Part (d)states that the nuclear coatings specialist's oversight of the replacement of the degradedcoatings ensures that the extent of repaired or replaced coatings encompasses soundcoating material.3. Letters dated March 29, 2012, and May 10, 2012, state that the essential cooling water(ECW) pump internal coatings will be inspected on a nominal 10-year frequency. TheMay 10, 2012, letter states that ECW pumps are located upstream of self-cleaningstrainers and the strainer size is sufficient to preclude tube blockage of downstream heatexchangers.

NOC-AE-l15003320Enclosure 1Page 12 of 13Issue:1. LR-ISG-201 3-01, "Aging Management of Loss of Coating or Lining Integrity for InternalCoatings/Linings on In-Scope Piping, Piping Components, Heat Exchangers, and Tanks,"AMP XI.M42, "Internal Coatings/Linings for In-Scope Piping, Piping Components, HeatExchangers, and Tanks," recommends that when peeling, delamination, blisters, orrusting are observed during inspections or when cracking and flaking that does not meetacceptance criteria is observed during inspections, the subsequent inspection interval is 4years instead of 6 years. The responses to PAI 3.0.3-2a Part (d) state that the specificdegraded coatings will be replaced and therefore inspections will continue at a 6-yearinterval. However, the 4 year inspection interval is recommended regardless of whetherrepairs are conducted on the degraded coatings detected during an inspection. With aknown degradation mechanism potentially occurring in other locations with the samecoating and environment, the staff concluded that subsequent inspections should beconducted more frequently than if no degradation was noted in prior inspections. Thestaff lacks sufficient information to conclude that a 6-year inspection interval is adequatewhen the extent of coating degradation, similar to the observed degradation that wasrepaired, is not known.2. The "corrective actions" program element of LR-ISG-2013-01, "Aging Management ofLoss of Coating or Lining Integrity for Internal Coatings/Linings on In-Scope Piping,Piping Components, Heat Exchangers, and Tanks," AMP XI.M42, "InternalCoatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks,"recommends that testing or examination be conducted to ensure that the extent ofrepaired coatings/linings encompasses sound material. The extent of blistering, peeling,and delamination is not typically detectable by visual inspection alone. The staff lackssufficient information to conclude that follow-on testing or examination will be directed tobe performed by the NCS.3. Although the ECW pumps are located upstream of self-cleaning strainers, this in and ofitself is not a sufficient basis to justify a nominal 10-year inspection frequency. The stafflacks sufficient information to conclude that the strainers will provide an effective barrierto flow blockage of downstream heat exchangers. Plant-specific operating experience ofthe ECW coatings has revealed degraded coatings.Req uest:1. With the exception of the internal coatings for the fire water storage tanks, state andjustify the basis for how the extent of coatings that could be experiencing similardegradation to coated areas that were repaired will be determined in a reasonable timeframe.2. State whether testing and examination will be conducted during a coating repair to ensurethat replaced coatings encompasses sound coating material.3. In regard to the self-cleaning strainers downstream of the essential cooling water pumps,state:a. What backup indications are available to determine that fouling is not occurring onthe self-cleaning strainers.b. Please provide the inspection interval of the strainer elements on the self-cleaningstrainers.

NOC-AE-1 5003320Enclosure 1Page 13 of 13STPNOC Response:STP response was provided in Response to Request for Additional Information for the Review ofthe South Texas Project, Units 1 and 2, License Renewal Application -Set 32 (TAO Nos.ME4936 and ME4937), dated November 12, 2015, NOC-AE-1 5003303, as SupplementalInformation -B2.1.13-5a, LR-ISG-2013-01 inspection frequency follow-up.

NOC-AE-1 5003320Enclosure 2Enclosure 2STPNOC LRA Changes with Line-in/Line-out AnnotationsList of Revised LRA SectionsRAI3.0.3.3.6-13.0.3.3.6-23.0.3.3.6-9Affected LRA SectionB2.1 .35B2.1 .35B3. 1I NOC-AE-1 5003320Enclosure 2Page 1 of 13B2.1.35 PWR Reactor InternalsProgram DescriptionThe PWR Reactor Internals program manages cracking, loss of material, loss of fracturetoughness, dimensional changes, and loss of preload for reactor vessel components that providea core structural support intended function. The program implements the guidance ofEPRI 1022863, PWR Internals Inspection and Evaluation Guideline (MRP-227-A, Rev .0) andEPRI 1016609, Inspection Standard for PWR Internals (MRP-228, Rev. 0). The programmanages aging consistent with the inspection guidance for Westinghouse designated primarycomponents in Table 4-3 of MRP-227-A and Westinghouse designated expansion componentsin Table 4-6 of MRP-227-A, and the Westinghouse designated existing components in Table 4-9of MRP-227-A. Primary components are expected to show the leading indications of thedegradation effects. The expansion components are specified to expand the primary componentsample should the indications of the sample be more severe than anticipated. The aging effectsof a third set of MRP-227-A internals locations are deemed to be adequately managed byexisting program components whose aging is managed consistent with ASME Section Xl TableIWB-2500-1, Examination Category B-N-3. A fourth ,qroup consistingq of those PWR internalscomponents for which the effects of all eight agqingq mechanisms are below the screening criteriawere placed in the No Additional Measures gqroup. No further action is required for managingq theagqingq of the No Additional Measures components.Program examination methods include visual examination (VT-3), enhanced visual examination(EVT-1), volumetric examination, and physical measurements. Bolting ultrasonic examinationtechnical justifications in MRP-228 have demonstrated the indication detection capability todetect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting.For some components, the MRP-227-A methodology specifies a focused visual (VT-3)examination, similar to the current ASME Code, Section Xl, Examination Category B-N-3examinations, in order to determine the general mechanical and structural condition of theinternals by (a) verifying parameters, such as clearances, settings, and physical displacements;and (b) detecting discontinuities and imperfections, such as loss of integrity at bolted or weldedconnections, loose or missing parts, debris, corrosion, wear, or erosion. In some cases, VT-3visual methods are used for the detection of surface cracking when the component material hasbeen shown to be tolerant of easily detected large flaws. In some cases, where even morestringent examinations are required, enhanced visual (EVT-1) examinations or ultrasonicmethods of volumetric inspection, are specified for certain selected components and locations.The program provides both examination acceptance criteria for conditions detected as a result ofmonitoring the primary components, as well as criteria for expanding examinations to theexpansion components when warranted by the level of degradation detected in the primarycomponents. Based on the identified aging effect, and supplemental examinations if required,the disposition process results in an evaluation and determination of whether to accept thecondition until the next examination or implement corrective actions. Any detected conditionsthat do not satisfy the examination acceptance criteria are required to be dispositioned throughthe corrective action program, which may require repair, replacement, or analytical evaluation forcontinued service until the next inspection.The PWR Vessel Internals program is a new program that has been implemented. The programwill include future industry operating experience, as it is incorporated into the future revisions ofMRP-227-A, to provide reasonable assurance for long-term integrity of the reactor internals. The NOC-AE-1 5003320Enclosure 2Page 2 of 13reactor vessel internals included in the scope of the PWR Reactor Internals program areidentified in Element 1. The scope of the program does not include welded attachments to theinternal surface of the reactor vessel because these components are managed by the ASMESection Xl Inservice Inspection, Subsections IWB, IWO, and IWO program (B2.1.1) (examcategory B-N-2) and/or the Nickel-Alloy Aging Management Program (B2. 1.34). The scope ofthe program also does not include BMI flux thimble tubes which are managed by the FluxThimble Tube Inspection program (B2.1.21).Aging Management Program ElementsThe results of an evaluation of each element against the 10 elements described in Appendix A ofNUREG-1 800, Standard Review Plan for Review of License Renewal Applications for NuclearPower Plants are provided below.Scope of Program -Element 1The scope of the program applies the guidance in MRP-227-A which provides augmentedinspection and flaw evaluation methodology for assuring the functional integrity of Westinghousereactor vessel internals. The scope of the PWR Reactor Internals program includescomponents that provide a core structural support intended function and are managed by theWestinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghousedesignated expansion components in Table 4-6 of MRP-227-A and applicable MRP-227-Amethodology license renewal applicant action items. MRP-227-A Table 4-9 also identifiesexisting program components whose aging is managed consistent with ASME Section Xl TableIWB-2500-1, Examination Category B-N-3.Primary components are expected to show the leading indications of the degradation effects.The expansion components are specified to expand the primary component sample should theindications of the sample be more severe than anticipated. The aging effects of a third set ofMRP-227-A internals locations are deemed to be adequately managed by existing programcomponents whose aging is managed consistent with ASME Section XI Table IWB-2500-1,Examination Category B-N-3. A fourth .qroup consistinq of those PWR internals components forwhich the effects of all eigqht agqingq mechanisms are below the screening criteria were placed inthe No Additional Measures gqroup. No further action is required for managingq the agqingq of the NoAdditional Measures components.The STP reactor vessel internals are divided into the following major component groups: thelower core support assembly (including the entire core barrel assembly, baffle-former assembly,neutron shield panel, core support plate, and energy absorber assembly), the upper core support(UCS) assembly (including the upper support plate, support column, control rod guide tubeassembly, upper core plate, and protective skirt), the incore instrumentation support structures(including the instrumentation columns (exit thermocouples), upper/lower tie plates, andinstrumentation columns (BMI)), and miscellaneous alignment/interface components (includinginternals hold-down spring, upper core plate guide pins, and radial support keys including clevisinserts).The following reactor vessel internals are included in the scope of the PWR Reactor Internalsprogram:1. Control rod guide tube assembly-Guide plate (cards) [Primary component]

NOC-AE-1 5003320Enclosure 2Page 3 of 13-Lower flange welds and adjacent base metal [Primary component]-Guide Tube Support Pins [Existing programs component]2. Core barrel assembly-Upper core barrel flange weld and adjacent base metal [Primary component]-Core barrel flange [Expansion component and Existing programs component]-Core barrel vertical axial welds and adjacent base metal [Expansion component]-Core barrel circumferential girth welds and adjacent base metal [Primary component]-Core barrel outlet nozzle welds and adjacent base metal [Expansion component]-Lower core barrel flange weld and adjacent base metal Addressed in AMR byComponent Types of "RVI Core Barrel Assembly") [Primary component]3. Baffle-former assembly-Baffle-former bolting [Primary component]-Baffle-former assembly baffle and former plates [Primary component]4. Alignment and interfacing components-Internals hold-down spring [Primary component]-Clevis insert bolts [Existing programs component]-Upper core plate alignment pins [Existing programs component]5. Bottom Mounted Instrumentation (BMI) Column assembly-BMI columns bodies [Expansion component]6. Upper Internals assembly-Upper core support skirt [Existing programs component]-Upper Core Plate [Expansion component]7. Lower internal assembly-XL lower core-plate [pans~ln Existing component]The scope of the program also does not include welded attachments to the internal surface ofthe reactor vessel because these components are managed by the ASME Section XI InserviceInspection, Subsections IWB, IWC, and IWD program (B2.1.1) (exam category B-N-2) and/orthe Nickel-Alloy Aging Management Program (B2.1.34). The scope of the program also does not NOC-AE-1 5003320Enclosure 2Page 4 of 13include BMI flux thimble tubes which are managed by the Flux Thimble Tube Inspection program(B2.1 .21).The STP reactor vessel internals configuration does not include the lower internals assembly(lower support column bodies and lower core plate) noted in MRP-227-A.The PWR Reactor Internals program is consistent with the following MRP-227-A assumptions(determination of applicability) which are based on PWR representative internals configurationsand operational histories.(1) STP has operated for less than 30 years of operation with high leakage core loading patterns.Operation with high leakage core loading was followed by implementation of a low-leakagefuel management pattern for the remaining operating life.(2) STP operates at fixed power levels and does not usually vary power based on calendar orload demand schedule.(3) STP has not implemented any design changes beyond those identified in industry guidanceor recommended by Westinghouse.Preventive Actions -Element 2The PWR Reactor Internals program does not prevent degradation due to aging effects, butprovides measures for monitoring to detect the degradation prior to loss of intended function.Preventive measures to mitigate aging effects such as loss of material and cracking includemonitoring and maintaining reactor coolant water chemistry consistent with the guidelines ofEPRI TR 1014986, PWR Primary Water Chemistry Guidelines, Volume 1. The primary waterchemistry program is described separately in the Water Chemistry program (B2.1 .2).Parameters Monitored or Inspected -Element 3The PWR Reactor Internals program monitors the following aging effects by inspection inaccordance with the guidance of MRP-227-A or ASME Section Xl Category B-N-3:(1). CrackingCracking is due to stress corrosion cracking (SCC), primary water stress corrosion cracking(PWSCC), irradiation assisted stress corrosion cracking (IASCC), or fatigue /cyclical loading.Cracking is monitored with a visual inspection for evidence of surface breaking lineardiscontinuities or a volumetric examination. Surface examinations may also be used tosupplement visual examinations for detection and sizing of surface-breaking discontinuities.(2). Loss of MaterialLoss of Material is due to wear. Loss of material is monitored with a visual inspection for grossor abnormal surface conditions.(3). Loss of Fracture ToughnessLoss of fracture toughness is due to thermal aging or neutron irradiation embrittlement. Theimpact of loss of fracture toughness is indirectly monitored by using visual or volumetricexamination techniques to monitor for cracking and by applying applicable reduced fracturetoughness properties in the flaw evaluations if cracking is detected and is extensive enough towarrant a supplemental flaw growth or flaw tolerance evaluation.

NOC-AE-1 5003320Enclosure 2Page 5 of 13(4). Dimensional ChangesDimensional Changes are due to void swelling and irradiation growth, distortion or deflection.The program supplements visual inspection with physical measurements to monitor for anydimensional changes due to void swelling, irradiation growth, distortion, or deflection.(5). Loss of PreloadLoss of preload is caused by thermal and irradiation-enhanced stress relaxation or creep. Lossof preload is monitored with a visual inspection for gross surface conditions that may beindicative of loosening in applicable bolted, fastened, keyed, or pinned connections.The PWR Reactor Internals program manages the aging effects noted above consistent with theinspection guidance for Westinghouse designated primary components in Table 4-3 of MRP-227-A and Westinghouse designated expansion components in Table 4-6 of MRP-227-A. MRP-227-A also identifies Existing Program components whose aging is managed consistent with ASMESection Xl Table IWB-2500-1, Examination Category B-N-3. See the component list in element1 to identify Primary, Expansion, and Existing components. A fourth group consisting of thosePWR internals components for which the effects of all eight aging mechanisms are below thescreening criteria were placed in the No Additional Measures for aging management arespecified.Detection of Aging Effects -Element 4The PWR Reactor Internals program detects aging effects through the implementation of theparameters monitored or inspected criteria and bases for Westinghouse designated PrimaryComponents in Table 4-3 of MRP-227-A and for Westinghouse designated ExpansionComponents in Table 4-6 of MRP-227-A. The aging effects of a third set of MRP-227-A internalslocations identified in Table 4-9 of MRP-227-A are deemed to be adequately managed byexisting program components whose aging is managed consistent with ASME Section Xl TableIWB-2500-1, Examination Category B-N-3.One hundred percent of the accessible volume/area of each component will be examined for thePrimary and Expansion components inspection category components. The minimumexamination coverage for primary and expansion inspection categories is 75 percent of thecomponent's total (accessible plus inaccessible) inspection area/volume be examined. Whenaddressing a set of like components (e.g. bolting), the minimum examination coverage forprimary and expansion inspection categories is 75 percent of the component's total population oflike components (accessible plus inaccessible).If defects are discovered during the examination, STP enters the information into the STPcorrective action program and evaluates whether the results of the examination ensure that thecomponent (or set of components) will continue to meet its intended function under all licensingbasis conditions of operation until the next scheduled examination. Engineering evaluations thatdemonstrate the acceptability of a detected condition will be performed consistent with WCAP-17096-NP.Monitoring and Trending -Element 5The program provides both examination acceptance criteria (See Element 6) for conditionsdetected as a result of monitoring the primary components as described in Element 4, as well ascriteria for expanding examinations to the expansion components when warranted by the level ofdegradation detected in the primary components. Based on the identified aging effect, and NOC-AE-1 5003320Enclosure 2Page 6 of 13supplemental examinations if required, the disposition process results in an evaluation anddetermination of whether to accept the condition until the next examination or implementcorrective actions. Any detected conditions that do not satisfy the examination acceptancecriteria (See Element 6) are required to be dispositioned through the corrective action program(See Element 7), which may require repair, replacement, or analytical evaluation for continuedservice until the next inspection.Acceptance Criteria -Element 6Examination acceptance for the Primary and Expansion component examinations are consistentwith Section 5 of MRP-227-A. ASME Section Xl section IWB-3500 acceptance criteria apply toExisting Programs components. The following examination acceptance criteria apply to the STPreactor vessel internals:Visual examination (VT-3) and enhanced visual examination (EVT-1)For existing program components, the ASME Code Section XI, Examination Category B-N-3provides the following general relevant conditions for the visual (VT-3) examination of removablecore support structures.(1) Structural distortion or displacement of parts to the extent that component function may beimpaired,(2) Loose, missing, cracked, or fractured parts, bolting, or fasteners,(3) Corrosion or erosion that reduces the nominal section thickness by more than 5 percent,(4) Wear of mating surfaces that may lead to loss of function; and(5) Structural degradation of interior attachments such that the original cross-sectional area isreduced more than 5 percent.In addition, for the visual examinations (VT-3) of Primary and Expansion components, the PWRReactor Internals program is consistent with the more specific descriptions of relevant conditionsprovided in Table 5-3 of MRP-227-A. EVT-1 examinations are used for detecting small surfacebreaking cracks and surface crack length sizing when used in conjunction with sizing aids. EVT-1 examination has been selected to be the appropriate NDE method for detection of cracking inplates or their welded joints. The relevant condition applied for EV-I'- examination is the sameas found for cracking in ASME Section XI section 3500 which is crack-like surface breakingindications.Volumetric examinationIndividual bolts are accepted (pass/fail acceptance) based on the detection of relevantindications established as part of the examination technical justification. When a relevantindication is detected in the cross-sectional area of the bolt, it is assumed to be non-functionaland the indication is recorded. Bolted assemblies are evaluated for acceptance based onmeeting a specified number and distribution of functional bolts. Acceptance criteria forvolumetric examination of STP reactor internals bolting are consistent with Table 5-3 of MRP-227-A.

NOC-AE-1 5003320Enclosure 2Page 7 of 13Physical MeasurementsPhysical measurement of the internals hold down spring is not required because STP internalshold down spring are fabricated from 403 stainless steel.Corrective Actions -Element 7The following corrective actions are available for the disposition of detected conditions thatexceed the examination acceptance criteria:(1) Supplemental examinations to further characterize and potentially dispose of a detectedcondition consistent with Section 5.0 of MRP-227-A;(2) Engineering evaluation that demonstrates the acceptability of a detected condition consistentwith WCAP-1 7096-NP;(3) Repair, in order to restore a component with a detected condition to acceptable status (ASMESection Xl); or(4) Replacement of a component with an unacceptable detected condition (ASME Section Xl)(5) Other alternative corrective action bases if previously approved or endorsed by the NRC.Relevant indications failing to meet applicable acceptance criteria are repaired or replaced inaccordance with plant procedures. Appropriate codes and standards are specified in both the"ASME Section Xl Repair, Replacement, and Post-Maintenance Pressure Testing" procedureand in design drawings. Quality assurance requirements for repair and replacement activitiesare also included in the STP Operations Quality Assurance Plan.STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptablein addressing corrective actions. The QA program includes elements of corrective action,confirmation process and administrative controls, and is applicable to the safety-related and non-safety related systems, structures, and components that are subject to aging managementreview.Confirmation Process -Element 8STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptablein addressing the confirmation process. The QA program includes elements of corrective action,confirmation process and administrative controls and is applicable to the safety-related and non-safety related systems, structures and components that are subject to aging managementreview.Administrative Controls -Element 9STP site QA procedures, review and approval process, and administrative controls areimplemented in accordance with the requirements of 10 CFR 50 Appendix B and are acceptablein addressing administrative controls. The QA program includes elements of corrective action,confirmation process and administrative controls and is applicable to the safety-related and non-NOC-AE-1 5003320Enclosure 2Page 8 of 13safety related systems, structures and components that are subject to aging managementreview.Operating Experience -Element 10Relatively few incidents of PWR internals aging degradation have been reported in operatingU.S. commercial PWR plants. However, a considerable amount of PWR internals agingdegradation has been observed in European PWRs, with emphasis on cracking of baffle-formerbolting. The experience reviewed includes NRC Information Notice 84-18, Stress CorrosionCracking in PWR Systems and NRC Information Notice 98-1 1, Cracking of Reactor VesselInternal Baffle Former Bolts in Foreign Plants. Most of the industry operating experiencereviewed has involved cracking of austenitic stainless steel baffle-former bolts, or SCC of high-strength internals bolting. SCC of control rod guide tube split pins has also been reported.Several other items with existing or suspected material degradation concerns that have beenidentified for PWR components are wear in thimble tubes and potentially in control guide cardsand observed cracking in some high-strength bolting and in control rod guide tube alignment(split) pins. The latter are conditions that have been corrected primarily through bolt replacementwith less susceptible material and improved control of pre-load.Based on industry operating experience, STP replaced the Alloy-750 guide tube support pins(split pins) with strained hardened (cold worked) 316 stainless steel pins during RefuelingOutage 1RE12 (Spring 2005) for Unit 1 and Refueling Outage 2RE1 1 (Fall 2005) for Unit 2. Thereplacement was conducted to reduce the susceptibility for stress corrosion cracking in the splitpins. There were no cracked Alloy X-750 pins discovered during the replacement process.The ASME Code, Section Xl, Examination Category B-N-3 examinations of core supportstructures conducted during Refueling Outage 1 RE1 5 (Fall 2009) for Unit 1, and RefuelingOutage 2RE14 (Spring 2010) for Unit 2, did not identify any conditions that required repair,replacement or evaluation.The ISI Program portion of the PWR Reactor Internals program at STP is updated to account forindustry operating experience. ASME Section Xl is also revised every three years and addendaissued in the interim, which allows the code to be updated to reflect operating experience. Therequirement to update the ISI Program to reference more recent editions of ASME Section XI atthe end of each inspection interval ensures the ISI Program reflects enhancements due tooperating experience that have been incorporated into ASME Section Xl.With exception of the ASME Section ISI portions, the PWR Reactor Internals program will be anew program and has no direct programmatic history. A key element of the MRP-227-A programis the reporting of aging of reactor vessel components. STP, through its participation in PWROwners Group and EPRI-MRP activities, will continue to benefit from the reporting of inspectioninformation and will share its own operating experience with the industry through those groups orINPO, as appropriate.As additional Industry and applicable plant-specific operating experience become available, theOE will be evaluated and appropriately incorporated into the program through the STP CorrectiveAction and Operating Experience Programs. This ongoing review of OE will continue throughoutthe period of extended operation, and the results will be maintained on site. This process willconfirm the effectiveness of this new license renewal aging management program byincorporating applicable QE and performing self assessments of the program.

NOC-AE-1 5003320Enclosure 2Page 9 of 13ConclusionThe implementation of the PWR Reactor Internals program provides reasonable assurance thataging effects will be adequately managed such that the systems and components within thescope of this program will continue to perform their intended functions consistent with the currentlicensing basis for the period of extended operation.

NOC-AE-1 5003320Enclosure 2Page 10 of 13B3.1 METAL FATIGUE OF REACTOR COOLANT PRESSURE BOUNDARYProgram DescriptionThe Metal Fatigue of Reactor Coolant Pressure Boundary program manages fatigue crackingcaused by anticipated cyclic strains in metal components of the RCPB. The program ensuresthat actual plant experience remains bounded by the transients assumed in the designcalculations, or that appropriate corrective actions maintain the design and licensing basis byother acceptable means.The Metal Fatigue of Reactor Coolant Pressure Boundary program consists of cycle countingactivities. The program will be enhanced to monitor and trend fatigue usage at selectedlocations in the reactor coolant pressure boundary and reactor vessel internals. The program willbe enhanced to include additional transients and locations identified by the evaluation of ASMESection III fatigue analyses, locations necessary to ensure accurate calculations of fatigue, andthe NUREG/CR-6260 locations for a newer-vintage Westinghouse Plant. The set includesfatigue monitorinq of the NUREG/CR-6260 sample locations for a newer-vintage WestinghousePlant, plant-specific bounding environmentally assisted fatigue (EAF) locations in the reactorcoolant pressure boundary, and reactor vessel internals locations with fatigue usagqecalculations. The supporting environmental life correction factor calculations were performed withNUREG/CR-6583 for carbon and low alloy steels and with NUREG/CR-5704 for austeniticstainless steels.The Metal Fatigue of Reactor Coolant Pressure Boundary program tracks the occurrences ofselected transients and will be enhanced to monitor the cumulative usage factors (CU~s) atselected locations using one of the following methods:1) The Cycle Counting (CC) method does not periodically calculate CUE; however, transientevent cycles affecting the location (e.g. plant heatup and plant cooldown) are counted to ensurethat the numbers of transient events assumed by the design calculations are not exceeded.2) The Cycle Based Fatigue (CBF) management method utilizes the CC results and stressintensity ranges generated with the ASME Ill methods that use six stress-tensors to performperiodic CUE calculations, consistent with RIS 2008-30, Fatigue Analysis of Nuclear Power PlantComponents for a selected location. The fatigue accumulation is tracked to determine approachto the ASME allowable fatigue limit of 1.0.The Metal Fatigue of Reactor Coolant Pressure Boundary program continuously monitors plantdata, and maintains a record of the data collected. The collected data are analyzed to identifyoperational transients and events, calculate usage factors for selected monitored locations, andcompare the calculated usage factors to allowable limits. Periodic review of the calculationsensures that usage factors will not exceed the allowable value of 1.0 without an appropriateevaluation and any further necessary actions. If a cycle count or CUE value increases to aprogram action limit, corrective actions will be initiated to evaluate the design limits anddetermine appropriate specific corrective actions. Action limits permit completion of correctiveactions before an assumed number of events in a fatigue analysis is exceeded.

NOC-AE-1 5003320Enclosure 2Page 11 ofl13NUREG-1 801 ConsistencyThe Metal Fatigue of Reactor Coolant Pressure Boundary program is an existing program that,following enhancement, will be consistent with NUREG-1 801,Section X.Ml, Metal Fatigue ofReactor Coolant Pressure Boundary.Exceptions to NUREG-1 801NoneEnhancementsPrior to the period of extended operation, the following enhancements will be implemented in thefollowing program elements:Scope of Program (Element 1) and Monitoring and Trending (Element 5)Procedures will be enhanced to include locations identified by the evaluation of ASME Section IIIfatigue analyses, locations necessary to ensure accurate calculations of fatigue, and theNUREGICR-6260 locations for a newer-vintage Westinghouse Plant. The set includes fatiquemonitorinq of the NUREG/CR-6260 sample locations for a newer-vintage Westinghouse Plant,plant-specific bounding environmentally assisted fatigue (EAF) locations in the reactor coolantpressure boundary, and reactor vessel internals locations with fatigue usagqe calculations.Scope of the Program (Element 1), and Parameters Monitored or Inspected (Element 3)Procedures will be enhanced to include additional transients that contribute significantly tofatigue usage identified by the evaluation of ASME Section III fatigue analyses.Scope of the Program (Element 1)Procedures will be enhanced to ensure the fatigue crack growth analyses, which support theleak-before-break analyses and ASME Section Xl evaluations, remain valid by counting thetransients used in the analyses.Detection of Aging Effects (Element 4)The procedures will be enhanced to 1) include additional transients necessary to ensureaccurate calculations of fatigue, 2) fatigue usage monitoring at specified locations, and 3) specifythe frequency and process of periodic reviews of the results of the monitored cycle count andCUF data at least once per fuel cycle. This review will compare the results against the correctiveaction limits to determine any approach to action limits and any necessary revisions to thefatigue analyses will be included in the corrective actions.Monitoring and Trending (Element 5)STP will perform a review of design basis ASME Class 1 component fatigue evaluations todetermine whether the NUREG/CR-6260-based components that have been evaluated for theeffects of the reactor coolant environment and reactor vessel internals on fatigue usage are thelimiting components for the STP configuration. If more limiting components are identified, themost limiting component will be evaluated for the effects of the reactor coolant environment on NOC-AE-1 5003320Enclosure 2Page 12 of 13fatigue usage. If the limiting location consists of nickel alloy, the methodology for nickel alloy inNUREG/CR-6909 will be used to perform the environmentally-assisted fatigue calculation.Preventive Actions (Element 2) and Acceptance Criteria (Element 6)The procedures will be enhanced to include additional cycle count and fatigue usage actionlimits, which will invoke appropriate corrective actions if a component approaches a cycle countaction limit or a fatigue usage action limit. Action limits permit completion of corrective actionsbefore the design limits are exceeded. The acceptance criteria associated with the NUREG/OR-6260 sample locations for a newer vintage Westinghouse plant will account for environmentaleffects on fatigue.Cycle Count Action Limits:Cycle count action limits are selected to initiate corrective action when the cycle count for any ofthe critical thermal or pressure transients is projected to reach the design limit within the nextthree fuel cycles.CUE Action Limits:CUF action limits require corrective action when the calculated CUF for any monitored location isprojected to reach 1 .0 within the next three fuel cycles.Corrective Actions (Element 7)Procedures will be enhanced to include appropriate corrective actions to be invoked if acomponent approaches a cycle count or CUF action limit.If a cycle count action limit is reached, acceptable corrective actions include:1 ) Review of fatigue usage calculations:a) To identify the components and analyses affected by the transient in question.b) To determine whether the transient in question contributes significantly to CUE.c) To ensure that the analytical bases of the high energy line break (HELB) locations aremaintained.2) Evaluation of remaining margins on CUE.3) Review of fatigue crack growth and stability analyses which support the leak before breakexemptions and relief from the ASME Section Xl flaw removal or inspection requirements toensure that the analytical bases remain valid. Re-analysis of a fatigue crack growth analysismust be consistent with or reconciled to the originally submitted analysis and receive the samelevel of regulatory review as the original analysis.4) Redefinition of the specified number of cycles (e.g., by reducing specified numbers of cyclesfor other transients and using the margin to increase the allowed number of cycles for thetransient that is approaching its specified number of cycles).5) Redefinition of the transient to remove conservatism in the pressure and temperature ranges.

NOC-AE-l15003320Enclosure 2Page 13 of 13These preliminary actions are designed to determine how close the approach is to the 1.0 limit,and from those determinations, set new action limits. If the CUF has approached 1.0 then furtheractions described below for cumulative fatigue usage action limits may be invoked.If a CUF action limit is reached acceptable corrective actions include:1) Repair the component.2) Replace the component. If a limiting component is replaced, assess the effect on locationsmonitored by the program. If a limiting component is replaced, resetting its cumulative fatigueusage factor to zero, a component which was previously bounded by the replaced componentwill become the limiting component and may need to be monitored.3) Perform a more rigorous analysis of the component to demonstrate that the design code limitwill not be exceeded.Operating ExperienceThe STP industry operating experience program reviews industry experience, includingexperience that may affect fatigue management, to ensure that applicable experience isevaluated and incorporated in plant analyses and procedures. Any necessary evaluations areconducted under the plant corrective action program.The Metal Fatigue of Reactor Coolant Pressure Boundary program was implemented inresponse to industry experience that indicated that the design basis set of transients used forfatigue analyses of the reactor coolant pressure boundary did not include some significanttransients, and therefore might not be limiting for components affected by them. Examples:Thermal stratification of pressurizer surge line piping:In response to NRC Bulletin 88-1 1, Westinghouse performed a plant-specific evaluation of STPpressurizer surge lines. The surge line stratification analysis was based on STP designtransients. It was concluded that thermal stratification does not affect the integrity of thepressurizer surge lines. STP responses to NRC Bulletin 88-11 describe the inspections,analyses, and procedural revisions made to ensure that thermal stratification does not affect theintegrity of the pressurizer surge lines. In addition, the responses noted that fatigue analyseswere updated to ensure compliance with applicable codes and license commitments.Thermal fatigue cracking in normally-isolated piping:In 1988, as identified in NRC Bulletin 88-08, there were several instances of thermal fatiguecracking in normally stagnant lines attached to reactor coolant system (RCS) piping. This issuewas addressed by utilities by conducting evaluations and monitoring to ensure that furtherleakage would not occur. STP performed a complete analysis of systems connected to the RCS.The review concluded that the potential for the described thermal conditions existed only in thenormal charging, alternate charging, and auxiliary spray lines. However, these systems areseparated and only hot water can leak through the charging and auxiliary spray lines, reducingthe potential for thermal cycling.

NOC-AE-1 5003320Enclosure 3Enclosure 3Additional Supporting DocumentsIRAI I Additional Information3.0.3.3.6-3 PWROG-1 4072-NP, Rev. 0PWROG-1 4072-NP, Rev. 0, "South Texas Project Units 1 and 2 Summary Report for the FuelDesign / FueJ Management Assessments to Demonstrate MRP-227-A Applicability," June 3,2015.

NOC-AE-1 5003320Enclosure 5Enclosure 5STPNOC Regulatory Commitment

Enclosure

5NOC-AE-1 5003320Page 1 of 1Table A4-1 License Renewal CommitmentsCommitment... 7 iLRA :: = Imaplementation ,30 Enhance the Metal Fatigue of Reactor Coolant Pressure Boundary program procedures to: B3.1 Complete no later* include additional locations necessary to ensure accurate calculations of fatigue, than six months* inlud aditinal raniens tat ontrbut sinifcanly t faigu usgeprior to the period* include counting of the transients used in the fatigue crack growth analyses, which of extendedsupport the leak-before-break analyses and ASME Section Xl evaluations to ensure the operationanalyses remain valid, Inspections to be* include additional transients necessary to ensure accurate calculations of fatigue, fatigue complete no laterusage monitoring at specified locations, and specify the frequency and process of than six monthsperiodic reviews of the results of the monitored cycle count and CUF data at least once prior to the PEOper fuel cycle, or the end of the* include additional cycle count and fatigue usage action limits, which will invoke last refuelingappropriate corrective actions if a component approaches a cycle count action limit or a outage prior to thefatigue usage action limit. The acceptance criteria associated with the NUREG/CR-6260 PEO, whicheversample locations for a newer vintage Westinghouse plant will account for environmental occurs later.effects on fatigque locations in the reactor coolant pressure boundary, and reactor vesselCR1-30internals locations with fatigque usagqe calculations, andCR1-30* include appropriate corrective actions to be invoked if a component approaches a cyclecount action limit or a fatigue usage action limit. Acceptable corrective actions includefatigue reanalysis, repair, or replacement. Re-analysis of a fatigue crack growthanalysis must be consistent with or reconciled to the originally submitted analysis andreceive the same level of regulatory review as the oniginal analysis.__________