ML14206A837

Millstone Power Station, Unit 2 - Response to Request for Additional Information Regarding Aging Management Program Description: Inservice Inspection - Reactor Vessel Internals, License Renewal Commitment #13 (MF3402)

ML14206A837
Person / Time
Site:	Millstone
Issue date:	07/21/2014
From:	Sartain M D Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
14-286, TAC MF3402
Download: ML14206A837 (16)

Text

DominionDominion Nuclear Connecticut, Inc.5000 Dominion Boulevard, Glen Allen, VA 23060Web Address: www.dom.comU. S. Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555July 21, 2014Serial No.NSSL/MLCDocket No.License No.14-286RO50-336DPR-65DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 2RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING AGINGMANAGEMENT PROGRAM DESCRIPTION: INSERVICE INSPECTION -REACTORVESSEL INTERNALS, LICENSE RENEWAL COMMITMENT #13 (MF3402)By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the"Aging Management Program

Description:

Inservice Inspection -Reactor VesselInternals" to address License Renewal Commitment #13 for Millstone Power StationUnit 2 (MPS2). The submittal contains an updated Reactor Vessel Internals (RVI)Aging Management Program and RVI Inspection Plan in accordance with topical report,"Material Reliability Program: Pressurized Water Reactor Inspection and EvaluationGuidelines" (MRP-227-A).In an e-mail dated May 14, 2014, the Nuclear Regulatory Commission transmitted arequest for additional information (RAI) related to the submittal. Attachment 1 to thisletter contains DNC's response to the RAI.If you have any questions or require additional information, please contact Wanda Craftat (804) 273-4687.Sincerely,Mark D. SartainVice President -Nuclear Engineeringcommonweaft of VlkginilReg. # 140542My Commission Expires May 31, 2018COMMONWEALTH OF VIRGINIA)))COUNTY OF HENRICOThe foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, todayby Mark D. Sartain, who is Vice President -Nuclear Engineering of Dominion Nuclear Connecticut, Inc. He hasaffirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company,and that the statements in the document are true to the best of his knowledge and belief.Acknowledged before me this -l day of ",,i4V ,2014.My Commission Expires: n -__ I -'Notary Public1o~J Serial No. 14-286Docket No. 50-336Page 2 of 2Commitments made in this letter: None

Attachment:

1 .Response to Request for Additional Information Regarding License RenewalCommitment #13cc: U.S. Nuclear Regulatory CommissionRegion I2100 Renaissance BlvdSuite 100King of Prussia, PA 19406-2713Mohan C. ThadaniSenior Project ManagerU.S. Nuclear Regulatory CommissionOne White Flint North, Mail Stop 08 B111555 Rockville PikeRockville, MD 20852-2738NRC Senior Resident InspectorMillstone Power Station Serial No 14-286Docket No. 50-336Attachment 1Response to Request for Additional Information Regarding License RenewalCommitment #13Dominion Nuclear Connecticut, Inc.Millstone Power Station Unit 2 Serial No 14-286Docket No. 50-336Attachment 1, Page 1 of 13Response to Request for Additional Information Regarding License RenewalCommitment #13By letter dated July 31, 2013, Dominion Nuclear Connecticut, Inc. (DNC) submitted the"Aging Management Program

Description:

Inservice Inspection -Reactor Vessel Internals"to address License Renewal Commitment #13 for Millstone Power Station Unit 2 (MPS2).The submittal contains an updated Reactor Vessel Internals (RVI) Aging ManagementProgram (AMP) and RVI Inspection Plan in accordance with topical report "MaterialReliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines"(MRP-227-A).In an e-mail dated May 14, 2014, the Nuclear Regulatory Commission (NRC) transmitted arequest for additional information (RAI) related to the submittal. The response to this RAIis as follows:RAI IAs discussed in References 1 and 2, the staff has identified two additional questions thatall CE and Westinghouse design plants referencing topical report "Material ReliabilityProgram: Pressurized Water Reactor Inspection and Evaluation Guidelines" (MRP-227-A)must answer to close Applicant/Licensee Action Item (A/LAI) I related to plant-specificapplicability of the topical report. If the answer to either or both questions is yes, thenfurther evaluation will be necessary to demonstrate the applicability of MRP-227-A toMillstone Power Station, Unit 2 (MPS2). The staff therefore requests the followinginformation:1. Do the MPS2 RVI have non-weld or bolting austenitic stainless steel componentswith 20 percent cold work or greater, and if so do the affected components haveoperating stresses greater than 30 kilopounds per square inch?2. Has MPS2 ever utilized atypical design or fuel management that could make theassumptions of MRP-227-A regarding core loading/core design non-representativefor that plant, including power changes/uprates?DNC Response1. DNC is working with Westinghouse to provide this information. Based on vendorresource availability, DNC plans to respond to this question by December 19, 2014.2. MPS2 has not used atypical core design or fuel management. Average core powerdensity is based on licensed thermal power and the essentially fixed core geometry.For MPS2, the average core power density has been verified to be 83 Watts/cm3 forthe last six fuel cycles, and is less than the maximum value of 110 Watts/cm3recommended in the MRP-227-A applicability guidelines issued by Electric PowerResearch Institute (EPRI) (NRC Reference 2). For the peripheral fuel assembly Serial No 14-286Docket No. 50-336Attachment 1, Page 2 of 13power density, the calculated Figure of Merit (F) for the last six fuel cycles has beenF < 58 Watts/cm3, which is less than the maximum value of F = 68 Watts/cm3recommended in the EPRI applicability guidelines. The minimum height of theactive fuel to the fuel alignment plate in normal operation considering past andcurrent fuel designs is 13.16 inches, which is greater than the 12.4 inch minimumvalue recommended in the EPRI applicability guidelines. These core design andmanagement parameters are expected to remain representative for future plantoperations within the currently licensed thermal power limit.RAI 2In Attachment 2 to the submittal, in its response to A/LAI 7, the licensee described itsplant-specific evaluation of cast austenitic stainless steel (CASS) Reactor Vessel Internals(RVI). The licensee's evaluation used a screening approach using the criteria of U.S.Nuclear Regulatory Commission (NRC) Letter, "License Renewal Issue No. 98-0030,Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," (Ref. 3).The result of the screening was that 63 of 68 core support columns were determined to benon-susceptible (screened out) for thermal embrittlement (TE). For 5 of 68 columns, thecertified material test reports could not be located so the licensee conservatively assumedthese 5 columns are susceptible to TE.Recently, the staff has developed interim guidance (Ref. 4, 5) for screening of CASSmaterials that are susceptible to both TE and irradiation embrittlement (IE). Under theinterim guidance, low-molybdenum CASS materials (such as Type CF8) that receiveneutron fluences greater than 0.45 displacements per atom (dpa) (3x1020 n/cm2) areconsidered to be susceptible to TE and IE if the ferrite content is greater than 15 percent,while if the ferrite content is less than or equal to 15%, these materials are onlysusceptible to IE at neutron fluences greater than 1.5 dpa (lx1021 n/cm2). The staff notesthat all 63 MPS2 core support columns with CMTRs would screen out for TE based on thestaff's interim guidance.The columns screen in for IE based on the peak neutron fluence to the columns. Thestaff's understanding is that although the core support column welds are required to beinspected as a "Primary" category component by MRP-227-A, the weld inspection isconducted from above the core support plate, therefore no portion of the core supportcolumns would be viewed during the visual inspection of the core support column welds.1. Clarify the scope of the core support column weld inspection. Specifically, is anyportion of the columns below the core support plate viewed during the visualexamination of the core support column welds? If not, can any informationregarding the integrity of the columns be gained from the results of the core supportcolumn weld visual examination?2. If the core support columns (other than the welds) are not inspected, Serial No 14-286Docket No. 50-336Attachment 1, Page 3 of 13a. Provide a functionality evaluation of the lower support structure consideringthe effects of IE, and IE plus TE for those columns for which TE could not bescreened out; orb. Modify the RVI inspection program to include inspections of the core supportcolumns as a "Primary" or "Expansion" component. Provide the schedule,scope examination method and acceptance criteria for these inspections.Appropriate Primary link(s) to components with higher susceptibility to TEand IE would need to be identified if the columns are added as Expansioncomponents.DNC Response1. Due to the small diameter of most of the flow holes (less than 2%-inch except forone 3-inch flow hole per quadrant) and relative thick plate (approximately 2-inch),the portions of the core support columns below the core support plate are notaccessible for inspection in conjunction with the inspection of the core supportcolumn welds.During core support column weld visual examination, loss of integrity or materialdegradation would be detectable via evidence of changes to the surface at the weldlocation. Cracking (stress corrosion cracking (SCC), irradiation assisted stresscorrosion cracking (IASCC), and fatigue including damaged material) or structuraldistortion of the embrittled material at the weld would appear as a disruption of thenormally smooth machined surface of the lower core support plate.2. The question suggests that DNC conservatively assumed the five support columnsto be susceptible. To clarify, the response to A/LAI 7 stated that the five coresupport columns were "potentially susceptible" to TE based on the range of elementconcentrations permitted by the CASS material specification. The NRC statementrepresents the worst case possibility if no further information were availableregarding typical CASS composition. However, there is sufficient informationavailable to conclude with reasonable assurance that the remaining five supportcolumns were produced with a composition that would also result in ferrite contentswithin acceptable screening limits. A statistical analysis is presented to support thisconclusion:First, as noted in Table 1, note 3 of the July 31, 2013 submittal, the maximum ferritecontent of the available 63 column compositions is 9.84%1. The ferrite contents ofthe known 63 columns are based on 25 separate material heats. The average ofthe ferrite contents for the 25 heats is 5.53%, the minimum is 3.43%, and thestandard deviation is 1.43%. Thus, the maximum calculated ferrite content of9.84% is three standard deviations above the average. At five standard deviationsabove the average, the ferrite content would be 12.68%. For a normal distribution,Ferrite contents are weight percents based on Hull's equivalency using material element composition takenfrom available certified material test reports (CMTR)

Serial No 14-286Docket No. 50-336Attachment 1, Page 4 of 13only about six out of ten million measurements would exceed five standarddeviations above the mean. On this basis, DNC concludes that the CASS supportcolumns were consistently produced with acceptably low ferrite content, and it isstatistically unlikely that any of the five support columns with unavailablecomposition data would have a ferrite content greater than 12.68%, which is wellbelow the statically cast CF8 screening criteria of 20%2. Hence, DNC concludeswith reasonable assurance that all 68 CASS core support columns meet thescreening criteria and therefore, are not susceptible to TE.Since the core support columns screen out for TE, the CASS material in the upperportions of the columns may be considered equivalent to wrought austeniticmaterial with respect to IE. Each arm of the core support columns has a peg that isinserted into and welded to the core support plate. This location exposes the CASSmaterial to a bounding level of irradiation such that the Primary inspection of thisweld area required by MRP-227-A constitutes adequate management for potentialIE of the core support columns.RAI 3For the components with material differences identified in the A/LAI 2 evaluation, identifythe component, material type at MPS2, and the generic material from MRP-227-A.Describe how the difference in material was evaluated and explain why no changes to theaging management requirements were needed for those components.DNC ResponseThere were only two material differences identified in the A/LAI 2 evaluation. They areaddressed below." The in-core instrumentation (ICI) guide tubes are listed in "Materials ReliabilityProgram: Screening, Categorization, and Ranking of Reactor Internals Componentsfor Westinghouse and Combustion Engineering PWR Design" (MRP-191) (DNCReference 1) as Type 316 stainless steel, while for MPS2 they are fabricated fromType 304 stainless steel. The degradation mechanisms of concern listed for the ICIguide tubes are SCC and IE. Types 304 and 316 materials fall under the sameaustenitic stainless steel category and the two materials have equivalentdegradation susceptibility. Therefore, the same degradation mechanisms listed inMRP-191 are applicable to both materials and no change to aging managementrequirements for the ICI guide tubes is required." The control element assembly (CEA) shroud material is listed in MRP-191 as eitherType 304 stainless steel, or cast austenitic grade CPF8 or CF8 stainless steel.Although not listed separately in MRP-191, the shaft retention pin and retentionblock are sub-part items of the CEA shroud, and are Type 304L stainless steel2 Screening value for statically cast grade CF8 CASS as justified in DNC Reference 2 Serial No 14-286Docket No. 50-336Attachment 1, Page 5 of 13material for MPS2. The MRP-191 degradation mechanism of concern for the Type304 stainless steel CEA shroud material is SCC. Types 304 and 304L are in thesame austenitic stainless steel category and have equivalent degradationsusceptibility for SCC. Therefore, no additional degradation mechanisms areapplicable for the Type 304L components and no change to aging management isrequired for the CEA shroud.RAI 4Provide the MRP-191 equivalent component name for the following subcomponents thatare listed in Table 3.1.2-2 of the MPS2 LRA:" Expansion Compensating Ring" Fuel Alignment Plate Guide Lugs and Guide Lug InsertsDNC Response" The MRP-1 91 equivalent component name for the Expansion Compensating Ringis the "hold-down ring" listed in Table 4-5, page 4-11 of MRP-1 91 (DNC Reference1)." MRP-191 lists the Fuel Alignment Plate "Guide Lugs" and "Guide Lug Inserts" inTable 4-5, page 4-13.RAI 5The in-core instrumentation (ICI) flux thimble tubes are not included in MPS2 LicenseRenewal Application (LRA) Table 3.1.2-2, "Reactor Vessel, Internals, and Reactor CoolantSystem -Reactor Vessel Internals -Aging Management Evaluation." This component isalso not included in the Generic Aging Lessons Learned Report, Rev. 0. However,License Renewal Interim Staff Guidance (LR-ISG-2011-04): "Updated Aging ManagementCriteria for Reactor Vessel Internal Components for Pressurized Water Reactors, (Ref. 6),"Table IVB3, "Reactor Vessel, Internals, and Reactor Coolant System -Reactor VesselInternals -Combustion Engineering," does have line item IV.B3.RP-357 for ICI: /CIThimble Tubes -Lower. However, the applicable aging effect/mechanism is loss ofmaterial due to wear, not change in dimensions.Were the ICI Flux Thimble Tubes identified as a component subject to aging managementin the LRA for MPS2? If so, what aging effects and mechanisms were determined torequire aging management and which program(s) were credited for managing aging of the/C/ flux thimble tubes? If not identified in the LRA as subject to AMR, what is the basis forincluding the /C/ flux thimbles as a component required to be addressed under MRP-227-A A/LA/ 3?

Serial No 14-286Docket No. 50-336Attachment 1, Page 6 of 13DNC ResponseThe 1C1 flux thimble tubes were not identified as a component subject to agingmanagement in the license renewal application (LRA) for MPS2. Since the 101 design forMPS2 does not involve a pressure boundary function for the thimble tubes, the licenserenewal application did not identify any aging effects requiring management. The fluxthimble tubes are discussed in the submitted aging management program because MRP-227-A considers the flux thimble tubes to be subject to an Existing program and A/LAI 3requires that the evaluation of any future aging management program requirements bedescribed.The requirements of the applicable Existing program for MPS2 addressed dimensionalchanges of the flux thimble tubes which were completed upon their replacement in 2009.By design, the replacement flux thimble tubes have sufficient margin for future dimensionalchanges and do not require further monitoring.It is noted that loss of material due to wear of the 1C1 thimble tubes is listed as line itemIV.B3.RP-357 in LR-ISG-2011-04, and that the applicable aging management program isChapter XI.M16A in Revision 2 of the Generic Aging Lessons Learned report. However,the MPS2 design does not include the interface/features which caused significant fluxthimble wear at the fuel alignment plate in a different Combustion Engineering design.Therefore, there is no Existing program at MPS2 to manage wear of the flux thimble tubes.The limited operating experience related to wear of replacement flux thimble tubes notedin the submitted AMP is for information only. Since the incore monitoring system hasmultiple degrees of redundancy and the flux thimble tubes only guide the insertion of theinstrumentation, the wear discussed in the operating experience does not affect thefunctioning of the incore monitoring system. After the instrumentation is inserted,functioning of the system is not affected by flux thimble tube wear. The functionalcapability of the incore instrumentation system is monitored and maintained in accordancewith the requirements specified in Section 3.3.2 of the MPS2 Technical RequirementsManual.RAI 6The response to A/LAI 5 indicates for separation between the upper and lower coreshroud sections, a maximum gap of 1/8 inch is acceptable at the innermost comers and amaximum gap of 1/16 inch is acceptable at the corners furthest from the innermostcorners. The response also states the structural and functional effects associated with thepresence of these gaps have been evaluated (Westinghouse letter is referenced), and areacceptable.The staff requests the following additional information related to the evaluation thatdetermined the acceptable gap size:1. How are the acceptance criteria consistent with the licensing basis of MPS2?

Serial No 14-286Docket No. 50-336Attachment 1, Page 7 of 132. Other than distortion, what structural effects are expected to occur (for example,increased stresses), and how were these determined to be acceptable?3. How is the function of the core shroud affected, if at all, by the maximum allowableswelling?4. How were the effects on the core shroud functions determined to be acceptable andwhat is the source of the functionality acceptance criteria for the core shroud?DNC Response1. Gaps between the upper and lower core shroud (CS) subassemblies may resultfrom (postulated) irradiation-induced void swelling. The acceptance criteria for thevisual examinations of the CS consist of maximum allowable values for these gaps.The potential adverse structural effects of these maximum gaps were identified andwere determined to satisfy the stress criteria defined in the MPS2 final safetyanalysis report (FSAR) (see item 2 below). The potential adverse effects of thesemaximum gaps on the functions of the CS were identified and were determined tobe within the limits prescribed in the MPS2 FSAR (see items 3 and 4 below). Thus,the acceptance criteria for the visual examinations of the CS are consistent with theMPS2 licensing basis.2. Potential adverse structural effects of the maximum CS gaps are identified andevaluated below:Identification of Potential Adverse Structural Effects of CS Gapsa. structural effect on the interfacing horizontal plates of the upper and lower CSsubassembliesb. structural effect on the tie rods joining the upper and lower CS subassembliesEvaluation of Potential Adverse Structural Effects of CS Gapsa. Stresses due to irradiation-induced void swelling have not been calculated forany of the CE plants. Based on the stress classifications defined in Section III,Subsection NG of the ASME B&PV Code (prescribed in the MPS2 FSAR), voidswelling stresses are classified as secondary stresses since they are self-limiting. In a structural evaluation, the main impact of secondary stresses is onthe fatigue analysis. The inclusion of void swelling stresses would increase themaximum primary plus secondary stress range. However, because voidswelling stresses are not cyclical, they would not contribute to an increase in thecumulative usage factor. Therefore, it is reasonable to conclude that voidswelling stresses need not be included in the structural evaluation of the CS,and thus would not adversely impact the satisfaction of the MPS2 licensingbasis acceptance criteria.

Serial No 14-286Docket No. 50-336Attachment 1, Page 8 of 13b. Irradiation-induced void swelling would increase the thicknesses of theinterfacing horizontal plates of the upper and lower CS subassemblies. Thisincrease in plate thickness would produce an increase in the overall height ofthe CS assembly, and thus would increase loads in the tie rods which join theupper and lower CS subassemblies. However, the increase in plate thickness isvery small compared with the overall height of the CS assembly, and theresulting increase in tie rod loading is negligible. Stresses in the tie rods wouldcontinue to satisfy stress limits defined in the MPS2 FSAR.3. As described in the MPS2 FSAR, the CS provides an envelope for the core andlimits the amount of coolant bypass flow. The potential adverse effects of themaximum CS gaps on these functions are identified:a. increase in CS-to-core support barrel (CSB) bypass coolant flow,b. inward deflection of CS plates encroaching on fuel space.4. The potential adverse effects of the maximum CS gaps on the functions of the CS,identified in item 3 above, are evaluated below:a. Coolant flow jetting through the gaps between the interfacing horizontal platesof the upper and lower CS subassemblies would increase the CS-to-CSBbypass coolant flow. This increase in CS-to-CSB bypass flow wasconservatively estimated and added to the existing total bypass flow. Theresulting increased total bypass flow was less than the allowable valueprescribed in the MPS2 FSAR. Therefore, the effect of the CS gaps on corebypass flow was determined to be acceptable and bounded with regard to theMPS2 licensing basis.b. The maximum gaps between the interfacing plates of the CS upper and lowersubassemblies result from the vertical deflection of one plate relative to theother. There is also a horizontal (inward) component to this deflection. Themaximum inward deflection is less than the dimensional tolerance on thelateral position of the CS plates relative to the core centerline. In addition,this inward deflection, which would occur at core mid-height (approximately),could be accommodated by the lateral flexibility of the fuel assemblies.Therefore, the inward deflection of the CS due to irradiation-induced voidswelling would not have a significant adverse effect on the fuel assemblies.RAI 7With respect to the MPS2 plant-specific item "core support barrel assembly -crack stopholes at areas of prior fatigue cracking near thermal shield support bracket assemblies,"Table 4-2 of the RVI Program Description lists cracking due to fatigue as the onlyapplicable aging effect/mechanism. The staff requests the licensee provide the followingadditional information:

Serial No 14-286Docket No. 50-336Attachment 1, Page 9 of 131. Describe how the applicable aging mechanisms and effects for the period ofextended operation were determined. Provide a justification for the process used ifdifferent than the process described in MRP-22 7-A, Section 2.2. Specifically, provide details of the evaluation that determined that fatigue crackingwas the only aging effect requiring management for these locations. Explain whythe other aging effects/mechanisms generically evaluated in MRP-227-A are notapplicable to these locations.DNC Response1 .DNC used the guidance of MRP-191 to determine the applicable aging mechanismsand effects. The repair areas, including crack stop holes at the two former locationsof thermal shield support brackets, was considered a small scale feature of the coresupport barrel lower cylinder that was encompassed by the MRP-191 review of thelarger structure and included other surface discontinuities, such as the remainingseven abandoned thermal shield support brackets. The applicable agingmechanisms for the repair area were considered the same as those applicable to thecore support barrel lower cylinder. MRP-191 Table 5-2 lists the screened in agingmechanisms as SCC of welds, IASCC and IE. Since the welded support brackets atthe two locations were removed, the SCC aging mechanism no longer screens in forthe repair area.As noted in the AMP submittal, the repair was developed by Combustion Engineeringand was within the scope of the applicability of MRP-227-A. DNC has confirmed withWestinghouse that the repair area was known and evaluated as part of the coresupport barrel during the original MRP-191 screening and failure modes, effects, andcriticality analysis (FMECA) evaluations that were performed during the developmentof MRP-227-A. Therefore, the inclusion of the repair area in the Millstone AMP is anenhancement exceeding the requirements of MRP-227-A.2. The remaining aging mechanisms evaluated for applicability but which did not screenin for the core support barrel lower cylinder (and by inclusion the crack stop holes)are (1) wear, (2) TE, (3) void swelling (VS), and (4) irradiation stress relief/irradiationcreep (ISR/IC). Wear is not applicable to this area because there are no mating partsin contact with each other. Thermal embrittlement is not applicable because thecomponent is not cast austenitic stainless steel. Void swelling is not applicable forthe core support barrel lower cylinder because, as shown in Table 4-7 of MRP-191,the core support barrel lower cylinder is considered a T-cold component. Irradiationstress relief may occur to a slight degree at this location; however, the componentfunction is not dependent upon maintaining a controlled preload. Therefore, theremaining aging mechanisms are not applicable to the repaired area of the coresupport barrel lower cylinder and do not create a need for additional agingmanagement.

Serial No 14-286Docket No. 50-336Attachment 1, Page 10 of 13RAI 8For two of the "Primary" inspection category components applicable to MPS2, MRP-227-Apermits a demonstration of fatigue life versus a time-limited aging analysis (TLAA) insteadof inspection. These components are the Core Support Barrel Assembly -Lower FlangeWeld and the Lower Support Structure -Core Support Plate. For each of thesecomponents, Note 6 to Table 2 of the RVI Program Description applies, which states thisitem originally screened in for fatigue by MRP-227-A, but as permitted by MRP-227-A forthis item a plant specific fatigue evaluation has been performed and demonstratedacceptable fatigue life. Therefore, the listed enhanced visual examination (EVT-1) is notrequired and this item is subject to the normal American Society of Mechanical EngineersBoiler and Pressure Vessel Code, Section X1, Examination Category B-N-3 inspectionrequirements.Details of these plant-specific fatigue evaluations were not provided in the RVI ProgramDescription. Further, these analyses appear to be new TLAAs since the analyses are notidentified as such in the MPS2 LRA. Additionally, Title 10 of the Code of FederalRegulations (10 CFR) 54.37, "Additional records and recordkeeping requirements," (b)states that:After the renewed license is issued, the Final Safety Analysis Report (FSAR)update required by 10 CFR 50.71(e) must include any systems, structures, andcomponents newly identified that would have been subject to an agingmanagement review or evaluation of time-limited aging analyses in accordance with§ 54.21. This FSAR update must describe how the effects of aging will bemanaged such that the intended function(s) in § 54.4(b) will be effectivelymaintained during the period of extended operation.The staff therefore requests:1. Provide the plant-specific fatigue evaluations for the RVI components for whichfatigue evaluations are being credited in lieu of inspections.2. Discuss the need to update the MPS2 FSAR to reflect the fatigue analyses for thethree components.DNC Response1. The vendor did not provide a plant-specific evaluation for this MRP-227-A activity forMPS2. The evaluation for MPS2 was developed based on comparisons and scalingof analyses performed by the vendor for another plant. The methodology of theevaluation is described briefly in the following:

Serial No 14-286Docket No. 50-336Attachment 1, Page 11 of 131. Identify a reference plant that is similar to MPS2 in terms of RVI design andoperational parameters.2. Review the RVI design drawings to confirm that there are no significantdifferences between MPS2 and the reference plant.3. Review the analysis for MPS2 to obtain design loads applied to the coresupport plate (CSP) and the core support barrel/lower support structure(CSB/LSS) flexure weld. Similarly, review the analysis for the reference plantto obtain the same information for that plant.4. Confirm that the design loads applied to the MPS2 CSP and CSB/LSS flexureweld are either bounded by, or very similar to, the equivalent loads for thereference plant.5. Review the RVI design transients to confirm that there are no significantdifferences between MPS2 and the reference plant with regard to both thetypes of transient events and the numbers of lifetime occurrences in theoriginal design bases.6. Identify those design transients that have a significant impact on fatigue usagein RVI components.7. Use the stress calculation methodology defined in the analysis for thereference plant to calculate stresses in the MPS2 CSP and CSB/LSS flexureweld.8. Use the fatigue analysis methodology defined in the analysis for the referenceplant (consistent with ASME Code practice for that era) to calculate fatiguecumulative usage factors for the MPS2 CSP and CSB/LSS flexure weld.Consider significant design transients with numbers of cycles from the originaldesign bases, and also the projected 60-year cycle counts for thecorresponding transients based on MPS2 operating history.The results of the evaluation identified that the analyzed components had acumulative usage factor that did not exceed the screening value allowed by MRP-191. Therefore the MPS2 CSP and CSB/LSS flexure welds screened out for fatiguecracking concerns under MRP-227-A. However, existing ASME Section Xl programrequirements remain applicable.2. The MPS2 CSP and CSB/LSS flexure weld are not newly identified components foraging management review, but were considered in the development of the licenserenewal application. The fatigue evaluation described above was performed as apermitted option in the implementation of MRP-227-A in lieu of inspection. BecauseMRP-191 conservatively screened in certain components based on a lack of plantspecific fatigue susceptibility information and operational history, MRP-227-Apermitted the option to re-evaluate the fatigue screening of those components on aplant-specific basis. Since the evaluation is not required to demonstrate compliancewith a MPS2 licensing basis requirement, no FSAR description of the evaluation isrequired to be included in the FSAR.

Serial No 14-286Docket No. 50-336Attachment 1, Page 12 of 13RAI 9Explain or correct the following inconsistencies between the MPS2 "Primary" and"Expansion" inspection category components identified in Table 2 and 3 of the RVIProgram Description, and Tables 4-2 and 4-5 of MRP-227-A:1. Table 3 lists Core Shroud Assembly (Welded) -Remaining Axial Welds withapplicability to plant designs with core shrouds assembled in two vertical sections.MRP-227-A Table 4-5 has Core Shroud Assembly (welded) -Remaining axialwelds, ribs and rings, but this component is only applicable to plant designs withcore shrouds assembled with full-height shroud plates.2. Core Support Barrel Assembly -Core Barrel Assembly Axial Welds,"Expansion" component in Table 4-5 of MRP-227-A, is not included in Table 3 ofthe RVI Program Description.DNC Response1. The MPS2 AMP submittal Table 2 entry for the expansion examination of theRemaining Axial Welds is correct for welded core shrouds assembled in two verticalsections, which is the applicable design of MPS2. The published version of MRP-227-A Table 4-5 had an error that inadvertently omitted the correct entry for designslike MPS2. EPRI has been notified of the publishing error.2. The MPS2 AMP submittal Table 3 entry for the expansion examination of the CoreSupport Barrel Assembly Axial Welds was inadvertently omitted from the submittedprogram for MPS2 but was included in internal plant documents. It is reproducedbelow. This entry follows the last entry in Table 3 of Enclosure 1, page 25 of 44.Core Support All plants Cracking Core barrel Enhanced visual (EVT-1) 100% of one side of theBarrel (SCC) assembly girth examination, with initial and accessible weld andAssembly welds subsequent examinations adjacent base metalCore barrel dependent on the results of surfaces for the weldassembly axial core barrel assembly girth with the highestwelds weld examinations, calculated operatingstress.See Figure 4-15.

Serial No 14-286Docket No. 50-336Attachment 1, Page 13 of 13NRC References1. 2/25/2013 Summary of Telecon with EPRI and Westinghouse Electric Company,March 15, 2013 (ADAMS Accession No. ML13067A262)2. MRP-227-A Applicability Guidelines for Combustion Engineering and WestinghousePressurized Water Reactor Designs, Enclosure to MRP Letter 2013-025, October14, 20133. "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of CastAustenitic Stainless Steel Components," May 19, 2000 (NRC ADAMS AccessionNo. ML003717179)4. Summary Tables for CASS Position, March 12, 2014 (ADAMS Accession No.ML14072A012)5. Email re: Summary Tables for CASS Position, from Joseph Holonich (NRC) to KyleAmberge (EPRI) dated March 12, 2014 (ADAMS Accession No. ML14071A411)6. LR Interim Staff Guidance LR-ISG-2011-04: Updated Aging Management Criteriafor Reactor Vessel Internal Components for Pressurized Water Reactors. May 28,2013 (ADAMS Accession No. ML12270A251)DNC References1. Materials Reliability Program: Screening, Categorization, and Ranking of ReactorInternals Components for Westinghouse and Combustion Engineering PWR Design(MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.2. "Project No. 704 -BWRVIP Response to NRC Request for Additional Informationon BWRVIP-234", BWRVIP Letter 2014-086 dated May 23, 1014 from A. McGeheeand D Madison to Joseph Holonich (ADAMS Accession No. ML14174A841).