ML13127A369

Memo from M. Murphy to G. Wilson, Subj: Safety Evaluation Regarding Vermont Yankee Nuclear Power Station Core Plate Hold Down Bolt Inspection Plan and Analysis (TAC ME6248)

ML13127A369
Person / Time
Site:	Vermont Yankee
Issue date:	03/15/2012
From:	Murphy M C Division of Engineering
To:	Wilson G A Plant Licensing Branch 1
References
FOIA/PA-2013-0030, TAC ME6248
Download: ML13127A369 (17)

Text

OFi-riA.rL USE ONLY rOREAYIFR'ITMarch 15, 2012MEMORANDUM TO: George A. Wilson, Jr. ChiefPlant Licensing Branch LPL 1-1Division of Operating Reactor LicensingOffice of Nuclear Reactor RegulationFROM: Martin C. Murphy, Chief IRA!Mechanical and Civil Engineering BranchDivision of EngineeringOffice of Nuclear Reactor RegulationSUBJECT: SAFETY EVALUATION REGARDING VERMONTYANKEE NUCLEAR POWER STATION CORE PLATEHOLD DOWN BOLT INSPECTION PLAN AND ANALYSIS(TAC ME6248)By letter dated March 18, 2011, Entergy Nuclear Operations (Entergy) submitted a plant-specificanalysis report of the core plate hold down bolts (ML1 10840068). In Amendment 11 of thelicense renewal application (LRA), Entergy committed to either install core plate wedges orcomplete a plant-specific analysis to determine the acceptance criteria for continued inspectionof the core plate hold down bolts in accordance with boiling water reactor (BWR) Vessel andInternals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25) andsubmit the inspection plan and analysis to the NRC two years prior to the period of extendedoperation (PEO). By letter dated December 30, 2011, Entergy updated the commitment toindicate the inspection plan and analysis would be provided one year prior to the PEO.The Mechanical and Civil Engineering Branch, and the Vessels and Internals Integrity Branchcompleted the review of the applicable portions of the subject request related to the core platehold down bolt inspection plan and analysis (ML1 10840069), and the responses to requests foradditional information (ML1 20100126, ML1 1353A407, ML1 2037A066, and ML1 20100126) Thesafety evaluation input for the core plate hold down bolt inspection plan and analysis is providedas stated in the enclosure.Docket Nos.: 50-271Enclosure:As statedCONTACT: Chakrapani Basavaraju, NRR/DE/EMCB, (301)-415-1221Jeffrey C. Poehler, NRR/DE/EVIB, (301)-415-8353--FF4GA6LUE-GflE--- ZPRTETARY_1NFO-RMvAT1ON_

OFFICIAL U~ VNLY -P~U~RIEIA~Y INFORMATIUNMarch 15, 2012MEMORANDUM TO:George A. Wilson, Jr. ChiefPlant Licensing Branch LPL 1-1Division of Operating Reactor LicensingOffice of Nuclear Reactor RegulationMartin C. Murphy, Chief IRA!Mechanical and Civil Engineering BranchDivision of EngineeringOffice of Nuclear Reactor RegulationFROM:SUBJECT:SAFETY EVALUATION REGARDING VERMONTYANKEE NUCLEAR POWER STATION CORE PLATEHOLD DOWN BOLT INSPECTION PLAN AND ANALYSIS(TAC ME6248)By letter dated March 18, 2011, Entergy Nuclear Operations (Entergy) submitted a plant-specificanalysis report of the core plate hold down bolts (ML1 10840068). In Amendment 11 of thelicense renewal application (LRA), Entergy committed to either install core plate wedges orcomplete a plant-specific analysis to determine the acceptance criteria for continued inspectionof the core plate hold down bolts in accordance with boiling water reactor (BWR) Vessel andInternals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25) andsubmit the inspection plan and analysis to the NRC two years prior to the period of extendedoperation (PEO). By letter dated December 30, 2011, Entergy updated the commitment toindicate the inspection plan and analysis would be provided one year prior to the PEO.The Mechanical and Civil Engineering Branch, and the Vessels and Internals Integrity Branchcompleted the review of the applicable portions of the subject request related to the core platehold down bolt inspection plan and analysis (ML1 10840069), and the responses to requests foradditional information (ML120100126, ML11353A407, ML12037A066, and ML120100126) Thesafety evaluation input for the core plate hold down bolt inspection plan and analysis is providedas stated in the enclosure.Docket Nos.:Enclosure:As statedDISTRIBUTION:EMCB R/F50-271JKimCBasavarajuJPoehlerCRoquecruzADAMS ACCESSION NO.: ML12074A274[OFFICE INRRIDE/EMCB NRR/DE/EMCBNAME cl~asavaraju M~urphyDATE 03/15/2012 03/15/2012NRR/DE/EVIB NRR/DE/EVIBJPoehler SRosenberg03/15/2012 03/15/2012IOFFICIAL RECORD COPY SAFETY EVALUATION INPUT BY THE EMCB & EVIBVERMONT YANKEE NUCLEAR POWER STATION (VYNPS)CORE PLATE HOLD DOWN BOLT INSPECTION PLAN AND ANALYSISENTERGY NUCLEAR OPERATIONS. INC. (ENTERGY)DOCKET NO. 50-271TAC NO. ME62481.0 INTRODUCTION1.1 ApplicationBy letter dated March 18, 2011, Entergy Nuclear Operations (Entergy) submitted a plant-specificanalysis report of the core plate hold down bolts (ML1 10840068) (Ref. 5). Vermont Yankee is aBWR type 4 with Mark I containment design. In Amendment 11 of the license renewalapplication (LRA), Entergy committed to either install core plate wedges or complete a plant-specific analysis to determine the acceptance criteria for continued inspection of the core platehold down bolts in accordance with Boiling Water Reactor (BWR) Vessel and Internals Project,BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25) (Ref. 1, 2, 3, and 4)and submit the inspection plan and analysis to the NRC two years prior to the period ofextended operation (PEO). By letter dated December 30, 2011, Entergy updated thecommitment to indicate the inspection plan and analysis would be provided one year prior to thePEO.The Mechanical and Civil Engineering Branch (EMCB), and the Vessels and Internals IntegrityBranch (EVIB) completed the review of the applicable portions of the subject request related tothe core plate hold down bolt inspection plan and analysis (ML110840068, and ML110840069)(Ref. 5 and 6), and the responses to requests for additional information (ML120100126,ML1 1353A407, ML12037A066, and ML120100126) This safety evaluation input is based onreview of the core plate hold down bolt inspection plan and analysis (Ref. 5) submittal byEntergy, Vermont Yankee core plate hold down bolt stress analysis report (Ref. 6) prepared byGE Hitachi (GEH) Nuclear Energy document NEDC-33618P, Rev. 0, and the responses to theRequests for additional information (RAIs) (Ref. 8, 12, and 14).1.2 Core Plate AssemblyThe core plate assembly, located inside the BWR reactor pressure vessel, consists of aperforated stainless steel plate reinforced by stiffener beams and supported on the perimeter bya circular rim. Stiffener beams are welded to the core plate to carry the pressure loads fromdesign basis loss of coolant accident (LOCA) events. The pressure loading from LOCA causescompressive stresses in the lower edges of the stiffener beams. Cross ties or stabilizer beamsare added between the stiffener beams to prevent flange buckling by providing lateral support.ENCLOSURE

'A' "'U' 'E- ONLY -FROPRi-ARY i',iNFORMATIOKN-2-The core plate rim is bolted to a ledge on the core shroud by stainless steel studs which preventvertical movement. The rim hold down bolts attach the core plate to the core shroud. Thestabilizer beams or rods also provide support for in-core housing monitors. Core plate assemblyprovides lateral support for the fuel bundles, control rod guide tubes, and in-coreinstrumentation during seismic events and provides vertical support for the peripheral fuelassemblies. The core plate is positioned on the shroud ledge by four vertical aligner pins. Theseismic and other dynamic loads are shared between the friction load of the shroud to rim boltconnection, and the shear resistance of the aligner pins. During seismic events the core plateprovides lateral support for the core to prevent misalignment that could affect the insertion of thecontrol rods. For plants such as VYNPS that do not have wedges and studs between core platerim and the shroud, the core plate may shift more than 0.75 inch if sufficient hold down boltfailures are assumed, According to BWRVIP-25 (Ref. 1), control rod insertion testing hasdemonstrated that a core plate horizontal misalignment of 0.75 inch would not significantlyincrease the scram time, and a displacement of 1.0 inch would inhibit insertion. The criticalnumber of intact hold down bolts required to prevent lateral displacement during a seismic eventis plant unique, and can be determined from a plant specific analysis. Even if hold down boltfailures resulted in significant core plate movement preventing the insertion of control rods, theplant could still be brought to a safe shutdown condition using the standby liquid control (SLC)system. Core plates experience tensile stresses and have stress concentrations due tothreaded regions. GEH has also determined that core plate bolt stress relaxation occurs due tothermal and irradiation effects.2.0 REGULATORY EVALUATIONTitle 10 Part 54 of the Code of Federal Regulations 10 CFR 54.21(a)(3) requires that for eachcomponent within the scope of license renewal as defined in 10 CFR 54.4 and subject to agingmanagement review according to the criteria of 10 CFR 54.21 (a)(1 )(typically described as long-lived, passive components), applicants for license renewal must demonstrate that the effects ofaging will be adequately managed so that the intended function(s) will be maintained consistentwith the current licensing basis (CLB) for the period of extended operation.10CFR54.21(c)(1) requires an evaluation of time-limited aging analyses (TLAAs), as defined in10 CFR 54.3, which states that [TLAAs], for the purposes of this part, are those licenseecalculations and analyses that:(1) Involve systems, structures, and components within the scope of license renewal, asdelineated in § 54.4(a);(2) Consider the effects of aging;(3) Involve time-limited assumptions defined by the current operating term, for example, 40years;(4) Were determined to be relevant by the licensee in making a safety determination;(5) Involve conclusions or provide the basis for conclusions related to the capability of thesystem, structure, and component to perform its intended functions, as delineated in §54.4(b); andOFICIAL' ON'LY -PROPRIETARY INFOPM,A.M/.TION-

@gFi-;IAL u~USFNLY -P-ROPRlETAFR', I.... ,,,r,_,1,A,-, O,,., N-3-(6) Are contained or incorporated by reference in the CLB.10 CFR 54.21(1)(c) requires for each TLAA that the applicant shall demonstrate that-(i) The analyses remain valid for the period of extended operation;(ii) The analyses have been projected to the end of the period of extended operation; or(iii) The effects of aging on the intended function(s) will be adequately managed for theperiod of extended operation.The initial version of "BWR [Boiling Water Reactor] Vessel and Internals Project, BWR CorePlate Inspection and Flaw Evaluation Guidelines (BWRVIP-25) (Reference 1) was approved bythe NRC staff for providing acceptable guidance for the inspection and evaluation of core platecomponents (including the core plate rim hold-down bolts also referred to as the core plate hold-down bolts, or simply core plate bolts) for the current operating period (plants in their initial 40years of operation) by letter dated December 19, 1999 (Reference 2). By letter dated July 17,1997 (Reference 3), the BWRVIP submitted "Appendix B, BWR Core Plate Demonstration ofCompliance with the Technical Information Requirements of the License Renewal Rule (10 CFR54.21)." The NRC staff transmitted its safety evaluation for referencing BWRVIP-25 in licenserenewal applications, as modified by Reference 3, via letter dated December 7, 2001(Reference 4). Reference 4 concluded that BWRVIP-25 provided an acceptable basis formanaging aging of the core plate bolt components, provided that applicants for license renewalmeet the limitations and conditions and the plant-specific action items of the enclosed SE.Plant-specific Applicant Action Items 4 and 5 are most relevant. Applicant Action Item 4 of theSE (Reference 7) stated that due to the susceptibility of the rim hold-down bolts to stressrelaxation, applicants referencing the BWRVIP-25 report for license renewal should identify andevaluate the projected stress relaxation as a potential TLAA issue. Applicant Action Item 5stated, that until such time as an expanded technical basis for not inspecting the rim hold-downbolts is approved by the staff, applicants referencing the BWRVIP-25 report for license renewalshould continue to perform inspections of the rim hold-down bolts.Since VYPNS did not have a plant-specific stress relaxation TLAA analysis for the core platebolts, Entergy provided Commitment No. 29 in Amendment 11 to the VYNPS License RenewalApplication to either install core plate wedges or complete a plant-specific analysis to determineacceptance for continued inspection of core plate bolts in accordance with BWRVIP-25.3.0 TECHNICAL EVALUATION3.1 Licensee EvaluationBy letter dated March 18, 2011 (Reference 5), the licensee submitted its plant-specific analysisof the core plate bolts intended to fulfill the requirements of the commitment described above.The analysis report (Reference 6) was included as Attachment 1 to Reference 5. The licenseedescribed the core bolt stress analysis, load cases, load combinations and results from the plantspecific analysis. The licensee described the method of evaluation of stress relaxation of thecore plate bolts in Section 5.0 of Reference 6. The licensee's evaluation is based on proprietarydata generated by General Electric-Hitachi (GEH). Figure 5-1 of Reference 6 shows a meanUt- 1-ICIAL U~ UNLY -flflOPRIETARY iNFORMATION O)FF-CALUS ONL'Y -uPRuHRETAHY i -4-design curve fit to the plotted data, designated the GEH design curve. The licensee alsopresented in Figure 5-2 of Reference 6 data from BWRVIP-99, "BWRVIP Vessel and InternalsProject Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components," forType 304/316/348 wedge loaded double cantilever beam specimens (DCBs) in a BWRenvironment. The data are for higher fluence levels (4-6 x1 020 n/cm2) than those experiencedby the core plate bolts. Figure 5-3 of the Reference 6 shows some additional test reactor datacompared to the mean design curve determined using GEH data only. This figure shows theGEH design curve is conservative compared to the test reactor data.The licensee provided the results of their evaluation of the potential for stress relaxation of thecore plate bolts in Section 6.7 of Reference 6. The licensee provided the percentage of preloadrelaxation due to the peak neutron fluence predicted for the core plate bolts. The licenseeindicated that the fluence was a conservative estimate based on a flux evaluation performed insupport of the extended power uprate (EPU) for VYNPS in 2003.3.2 Staff Evaluation3.2.1 Loss of Preload of Core Plate Bolts (EVIB)The staff used BWRVIP-25 as guidance for our review of the licensee's evaluation of stressrelaxation of the core plate hold-down bolts. Appendix B to BWRVIP-25 provides anevaluation of the potential loss of preload in BWR core plate bolts that is intended to bebounding for all BWRs. Additionally, in the "Safety Evaluation Report (SER) related to theLicense Renewal of Vermont Yankee Nuclear Power Station," (NUREG-1907, Reference 7), thestaff noted that [VYNPS] did not calculate a plant-specific value of the neutron fluence at thecore plate bolts. However, in NUREG-1907, the staff concluded the core plate bolt fluenceshould remain bounded by the fluence used for BWRVIP-25, based on VYNPS maximum EOLRV neutron fiuence being lower than that of most BWR's. However, because the staff has notpreviously approved a calculated or estimated plant-specific value for the core plate bolt neutronfluence, in RAI 1, the staff requested the applicant provide the details of the flux evaluation thatwas used to determine projected total fast neutron fluence for the core plate bolts for a 60-yearplant life.In its response to RAI 1 by letter dated December 9, 2011 (Reference 8), the licensee provideda discussion of the flux evaluation. The licensee indicated that the flux evaluation was based ona best-estimate flux evaluation performed in 2003 in support of an extended power uprate(EPU). Results from the EPU flux evaluation were used to estimate the flux and fluence for thecore plate bolts at VYNPS. In the EPU flux evaluations, best estimate fast flux values weredetermined at the RV inside surface, core shroud inside surface, and surveillance capsule. Todetermine the flux at the bolt location, the licensee first determined the core midplane fluxcorresponding to the radial location of the bolt. The licensee then divided the bolt into twentyevenly spaced axial sections. A synthesized flux was determined for each section bymultiplying the core midplane flux at the radius of the bolts (3.09x1011 n/cm2-s, E> 1 MeV) bythe axial flux factor (defined as the ratio of the flux at a particular axial location to the coremidplane flux), times a safety factor of 1.5 intended to account for uncertainties associated withflux calculation for regions beyond the core beltline. The licensee then averaged the* ONLYt -PROPRIETARY iNFORM~ATiOIN OFIIA S "N -INI-URMA I iN--5-synthesized fluxes for the 20 bolt sections to obtain the average flux for the bolt over the axiallength of 7.09x1 09 n/cm2-s (E > 1 MeV). For time periods prior to the implementation of the EPUin 2003, the licensee's analysis ratioed the flux based on the previous power levels inmegawatts thermal (MWt) to the post-EPU flux. VYNPS operated at two different thermal powerlevels including the previous thermal power and a transitional power level for the cycle prior tofull EPU implementation. The licensee thereby obtained peak and average fluxescorresponding to each power level at which VYNPS has operated.To determine the EOL fluences for the core plate hold down bolts, the licensee then multipliedthe EFPY for each power level by the flux for that power level (peak and average) to determinethe peak and average fluences for the bolts. A peak 60-year fluence of 5.2x1 0 9 n/cm2 for thebolt was thus obtained. The staff checked the licensee's calculation and obtained the sameresult.The staff finds the response to RAI 1 acceptable because it provides an adequate description ofhow the core plate hold-down bolt flux was extrapolated, and includes appropriateconservatisms to ensure the flux used to project the loss of preload is bounding. Specifically, 1)the peak azimuthal flux at the radius of the bolts was used as the starting point, 2) a factor of 1.5was applied to the synthesized flux for each bolt section, and 3) peak bolt flux rather than theaxial average was used as the basis for the loss of preload projection. Therefore, the staff findsRAI 1 is resolved.The staff verified that the percentage reduction in preload assumed by the licensee matches thepercentage reduction in preload that is indicated by the GEH design curve based on thepredicted peak neutron fluence .The staff compared the licensee's prediction of the reduction inpreload to other industry data for stress relaxation. Industry data relevant to BWRs can befound in BWRVIP-99-A, "BWR Vessel and Internals Project -Crack growth Rates in IrradiatedStainless Steels in BWR Internal Components" (Reference 9), and MRP-175, "MaterialsReliability Program: PWR Internals Material Aging Degradation Mechanism Screening andThreshold Values" (Reference 10). BWRVIP-99-A provided two figures showing fraction ofstress remaining for bent beams exposed at 60 and 3000C in the Chalk River Reactor, for purenickel and Alloy X-750. BWRVIP-99-A also included the data for wedge-loaded dual cantileverbeam (DCB) specimens for Type 304/316/348 that was shown in Figure 5-2 of Reference 6.This data was for higher fluence levels; the trend line extrapolated to fluence levels comparableto the core plate bolts indicates a much lower degree of relaxation (5% reduction or 95%remaining preload) than the applicant determined based on the GEH data. Even if an upperbound trend line were drawn on this figure, the reduction in preload would only be about 10%(90% preload remaining). MRP-175, Figure H-7, provides a lower bound curve for percentageof remaining stress versus displacements-per-atom (dpa) for various austenitic stainless steelsand nickel-based alloys at various temperatures. It should be noted that displacements-per-atom (dpa) are a measure of irradiation damage to a material that does not exactly convert tofluence in neutrons per square centimeter (n/cm2), but in light-water reactor neutron spectra, 1dpa _= 6.7x1 020 n/cm.2 A conservative lower bound curve was used by the MRP since the intentof the curve is to screen for the potential of stress relaxation. At 0.1 displacements-per-atom(dpa), the lower bound curve is at 50% remaining stress. However, if only the data points forannealed type 304 stainless steel are considered, a more realistic lower bound is around 75% ofOFFICIAL USE ..NL.P T PaPRIETAY ,

-6-remaining stress at 0.1 dpa. In addition, if a best estimate curve were fit to this data theremaining stress value would probably be between 85-90% which is consistent with thereduction in preload assumed in the licensee's analysis Based on the industry data, the stafffinds that the licensee's estimate of remaining preload is reasonably consistent with both lower-bound and best-estimate values that would be determined from other industry data, which wouldrange from about 75-95%.Section 4.7.3 of NUREG-1907 (Reference 7) indicates that, as stated in Appendix B toBWRVIP-25, a 5-19% reduction in core plate hold-down bolt stress due to thermal andirradiation effects should be expected over the 40-year life of a plant. However, Appendix B toBWRVIP-25 does not provide the neutron fluence value on which the preload relaxationevaluation was based. For comparison to the predicted loss of preload (14%) used in theVYNPS analysis, in RAI 2 the staff requested the neutron fluence value on which the 5-19%loss of preload is based. In its response to RAI 2 contained in the letter dated December 9,2011, the licensee stated that the GE evaluation of core plate relaxation determined that theBWRVIP-25 maximum reported stress relaxation value of 19% is valid to an average neutronfluence level of 8x1019 n/cm2 or less, and that this fluence is an average fluence over the entirelength of the core plate bolt, determined at the peak azimuthal flux location. The staff finds theresponse to RAI 2 is acceptable because it demonstrates the licensee's fluence value isbounded by the neutron fluence values analyzed in BWRVIP-25. Also, if ratio of the VYNPSpeak neutron fluence to the maximum BWRVIP-25 neutron fluence is multiplied by themaximum stress relaxation from BWRVIP-25, a similar percentage of stress relaxation to thatassumed by the licensee is obtained. Therefore, the staff finds the licensee's projected loss ofpreload as a-function.of neutron fluence is consistent with BWRVIP-25 and is thereforeacceptable. RAI 2 is resolved.The staff finds the licensee's evaluation of the projected loss of preload of the VYNPS core platehold-down bolts due to irradiation-assisted stress relaxation is acceptable because 1) thelicensee appropriately estimated the peak fluence for the bolts at EOL based on its EPU fluenceevaluation; 2) the licensee's projection of loss of preload based on the peak bolt fluence isconsistent with what would be expected based on the BWRVIP-25 generic analysis and otherindustry data.However, cracking of the core plate hold-down bolts due to intergranular stress corrosioncracking (IGSCC) could also result in loss of load carrying capacity and did not appear to havebeen considered in the stress analysis of Reference 2. The staff requested additionalinformation related to the possibility of cracked bolts due to IGSCC in RAI 3, discussed in detailin the next section, since this topic is related to the inspection plan for the core plate hold-downbolts.3.2.2 Inspection Plan for Core Plate Hold-Down Bolts (EVIB)Reference 5 indicates that the sample size of VYNPS core plate hold down bolts inspected hasbeen changed from 50 % to 25 %. The frequency and method of the inspections will remain thesame (visual VT-3 inspection from the top of the bolts every other refueling outage). Thisrepresents a deviation from the BWRVIP-25 requirements for ultrasonic inspection. This level ofOgFFICIAL USE ONLY -FROFRlETARY~ INFORMyA I IN-OFFiCiAL U~ LJNLY -PROi-RIETARY INFORMATiON-7-inspection would probably reveal if there was widespread failure of the bolts but could misspartially cracked bolts or a small number of failed bolts.Therefore, in RAI 3, the staff requested the following information:1. Given that VYNPS has reduced the sample size for VT-3 from that recommended byBWRVIP-25, justify that the sample size of core plate hold down bolts being inspected isadequate to ensure that there will be sufficient intact bolts to meet the load requirementsof the plant-specific stress analysis.2. Justify that performing the VT-3 inspection from above the core plate will provide asufficient level of assurance that cracked or broken bolts will be detected, given thatBWRVIP-25 recommends performing the VT-3 inspection from below the core plate.3. Does the core plate stress analysis account for some portion of the core plate boltsbeing either completely or partially cracked due to intergranular stress corrosion crackingor irradiation assisted stress corrosion cracking? If so, describe how the cracking wasaccounted for.4. If cracking was not accounted for in the stress analysis, provide a justification forcracking not being considered.In its response by letter dated December 9, 2011 (Reference 8), the licensee indicated thefollowing:With respect to RAI 3 Item 1, VYNPS performed inspection of 50% of the core plate hold-downbolts for four successive outages with no noted degradation. The licensee cited section 3.2.2.2of BWRVIP-25, which allows the re-inspection schedule for the core plate hold-down bolts to beadjusted based on good inspection results combined with good operating experience. Based onperformance, the licensee adjusted the inspection frequency and sample size to 25% of thebolts every other outage beginning in 2007 and has performed these inspections since that timewith no noted degradation. The staff notes that the inspections performed were VT-3 visualexaminations performed from above the core plate rather than VT-1 visual examinationsperformed from below the core plate as prescribed by BWRVIP-25.With respect to RAI 3 Item 2, VYNPS stated that it is currently industry practice only to performVT-3 inspections from above the core plate, because performing VT-1 examination from belowthe core plate requires extensive disassembly and a UT technique has yet to be developed.The licensee also referenced its March 18, 2011 letter (Reference 11) documenting its deviationfrom the BWRVIP-25 inspection requirements. Reference 11 provides a summary of thelicensee's justification for the deviation, which cites the following factors supporting thedeviation:" Low susceptibility to cracking and high flaw tolerance of the bolting,* Postulated flaws would not grow to a size that significantly reduces the bolt preload overthe life of the plant* Redundancy of structural components that would prevent adverse displacement of thecore plate even if significant cracking occurs in the bolts..FF.IA UE ONL'," -FRuJ<IlTARY IN-UTMATION

• OFFICIAL 'C ,ONL,, vrROPR;I__TAR", "l-"" "."A-l-,N-8-* Even if all the core plate hold-down bolts and the redundant hardware failed, preventinginsertion of the control blades, the standby liquid control system could be used to bringthe reactor to a safe shutdown.In response to RAI 3 Item 4, the licensee stated that the core plate stress analysis did notaccount for some portion of the core plate [hold-down] bolts either completely or partiallycracked due to IGSCC or irradiation assisted stress corrosion cracking (IASCC). In response toRAI 3 Item 3, the licensee provided its justification for not assuming that some portion of thecore plate bolts were either completely or partially cracked due to IGSCC or IASCC. In itsjustification, the licensee cited Section 2.2.9 of BWRVIP-25, which notes that thecore platehold-down bolts are not sensitized, which reduces the IGSCC susceptibility, and that there havebeen no instances of IGSCC in the field of these bolts.The staff agrees that the IASCC susceptibility of these bolts is low, because the peak fluencelevel of the bolts is below the range at which IASCC can typically begin to be a factor in BWRs(5x1 02o n/cm2). However, although bolts are not sensitized, the staff was concerned they couldpotentially be cold worked which can increase the susceptibility to IGSCC.The licensee did not account for the possibility of some cracked or broken bolts in their analysis.Since the licensee is inspecting only a sample of the bolts, and the inspection method used isvisual VT-3 examination, which only allows the ends of the bolts and nuts to be examined, thestaff had concerns that the current inspection plan is not capable of detecting cracked or brokenbolts. Only the top end of the bolt and the nut can be viewed from above the core plate. Thenut is fillet welded to the bolt to prevent loosening. To address these issues, the staff requestedthe following additional information:1. Provide a justification that the VT-3 visual examinations would be effective atdetecting failed core-plate hold-down bolts.2. What percentage of core plate bolts for VYNPS must be intact to avoidexceeding the allowable stresses on the bolts as given by Table 8-1 of theanalysis (Reference 6)?3. Considering the effectiveness of the VT-3 examination at detecting cracked orbroken bolts, does the percentage of the bolts being sampled supportdemonstration that the required number of bolts are intact, assuming no failedbolts are found in the sample? Provide a statistical argument or analysis similarto that provided in BWRVIP-25, Section 3.2.2.2.4. If a statistical argument cannot be made, provide a more detailed basissupporting a very low probability of significant loss of load bearing capability dueto IGSCC of the bolts, and/or revise the analysis to account for the possibility ofsome bolt failures due to SCC.In response to the follow-up RAI 1 by letter dated February 1, 2012 (Reference 12), the licenseejustified the effectiveness of the VT-3 visual examinations by citing a portion of General ElectricServices Information Letter (SIL) No. 588R1. The information indicates that the core plate holddown bolts for older BWRs have low susceptibility to SCC because they were procured to aCFFkiiAL WLUI: NLY -PR~OPRIETARY liINFORMvATiOiN, SO,,FCIAL..,U,- ONLY rIPRO'IFTARY' iNFORMvATION-9-specification prohibiting cold forming operations after solution heat treatment, and have a lowpreload (10-15 ksi). Therefore, the SIL 588 R1 recommended inspection is to show the boltshave not loosened and rotated due to a combination of vibration and failure of the welds on thelocking device, which should be obvious by visual VT-3 examination. The staff finds thelicensee's response to follow-up RAI 1 acceptable because the information provideddemonstrates the core plate hold-down bolts should have low IGSCC susceptibility.In its response to follow-up RAI 1, the licensee also cited Section 3.2.5 of BWRVIP-47-A, "BWRVessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and FlawEvaluation Guidelines," which states that"The BWRVIP has determined that removing or dismantling of internalcomponents for the purpose of performing inspections is not warranted to assuresafe operation. However, on occasion, utilities may have access to the lowerplenum due to maintenance activities not part of normal refueling outageactivities. In such cases, utilities will perform a visual inspection to the extentpractical. Results of the inspection will be reported to the BWRVIP and will beforwarded by the BWRVIP to the NRC."The licensee further stated that the VYNPS Reactor Vessel Internals (RVI) Program contains aprovision for performing inspections when access to the lower plenum is available due tomaintenance activities.Although the specification of no cold forming and low preload for the bolts would not completelypreclude IGSCC, these factors combined with operating experience for core plate bolts acrossthe BWR fleet, which has noted no failures of these bolts, provides reasonable assurance thatwidespread IGSCC failure of these bolts is unlikely. Further, the staff agrees that the VT-3examination should detect loosening of the bolts due to vibration combined with failure of thelocking device welds. Finally, in accordance with BWRVIP-47-A, inspections of opportunitywhen access to the lower plenum is possible due to maintenance should provide additionalassurance that core plate bolts are intact since it should be possible to view the threadedportion of the bolts from below the lower plenum region. Therefore, follow-up RAI 1 is resolved.In response to follow-up RAI 2, the licensee indicated that the VYNPS core-plate stress analysisdid not assume any of the bolts were initially failed or cracked, and that this is consistent withthe methodology of BWRVIP-25, Appendix A. Therefore, the staff could not determine from thelicensee's response if there is an acceptable number of bolts that could be failed that would notresult in the allowable stresses being exceeded in one of the design-basis scenarios analyzed inthe stress analysis.In response to follow-up RAI 3, the licensee indicated that they had performed a statisticalevaluationusing ANSI-ASQ Standard Z1.4 Table 1. This table indicated a sample size of 13 fora nonconformance value of 1% -i.e., the finding of no failures in the sample of 13 bolts indicatesthat less than 1% of the bolts in the overall population of 30 bolts would be defective. Based onthis statistical evaluation, the licensee determined that their previous sample size of 25% for theVT-3 examination is inadequate, and stated that they would increase the sample size to 50% or-OFFICIAL USE ONLY -PROPRETARY1-' IFORIvIA I IUN OgF-FICIAL USE ONLY- rROFR',ETARY ,l,--NFOR -10-15 bolts, beginning with RFO 31. The licensee also included this change in sample size as acommitment in Attachment 2 to the February 1, 2012 letter. The licensee stated that noresponse to follow-up RAI 4 is required because a statistical argument was made in response toItem 3.The staff notes that the licensee's statistical evaluation is based on a standard used todetermine the acceptance quality limit (AQL), which is defined as the quality level that is theworst tolerable process average when a continuing series of lots is submitted for acceptancesampling. This standard is typically used for quality assurance of manufactured products. Thestandard does not describe the statistical analysis behind the determination of the proportion ofthe population that is defective. Therefore, the staff performed an independent statisticalevaluation of the probable number of cracked bolts in the overall population given that nocracked bolts are found in the 50% sample. The staff used a hypergeometric distribution, whichcan be used as the basis for a sampling scheme (a hypergeometric experiment) that samples apopulation for attributes without replacement and which satisfies the following conditions(Reference 13):" The sampled population is finite;* Once an item is selected, it cannot be selected again;* The size of the population is known;* The number of items with the attribute of interest is known;* -Each item in the sample is drawn at random.The staff determined that if no cracked bolts are present in the 50% sample, the probability thatthe number of cracked bolts in the overall population would result in the ASME Code allowablestresses being exceeded, based on the margins given in Table 8-1 of Reference 6, is less than5%.The staff also notes there are several conservatisms in the VYNPS stress analysis that make iteven less likely the ASME Code allowable stresses would be exceeded. First, as noted in theresponse to RAI 4 via letter dated January 5, 2012 (Reference 14), a conservative coefficient offriction was used in determining the reduction in the applied horizontal loading due to frictionalresistance. Second, in Scenarios 1 and 3, no credit was taken for load being borne by thealigner pins.Based on the staff's independent statistical evaluation, and considering the conservatisms in theVYNPS core plate hold-down bolt structural analysis, follow-up RAI's 2 and 3 are resolvedbecause there is reasonable assurance that the number of bolts that could possibly be cracked,given the finding no cracked bolts in the proposed sample inspection, would not result in theallowable stresses being exceeded in the event of a design basis accidentBased on the information submitted by the licensee supporting low IGSCC susceptibility for theVYNPS core plate hold-down bolts, and the margins present in the VYNPS core plate boltstress analysis as supported by the staff's statistical evaluation, the staff finds the licensee's..OFFICIAL U~L UI'.JLY -PROPRIETAi-'y INFORMATION~

-11 -proposal to visually inspect a 50% sample of the bolts every other refueling outage to beacceptable until the BWRVIP revises its guidance for core plate hold-down bolt inspection andevaluation.3.2.3 Stress Analysis of Vermont Yankee Core Plate Hold-Down Bolts (EMCB)The licensee performed stress calculations to demonstrate the structural adequacy of theVYNPS core plate bolts and aligner pins. The methodology and assumptions utilized areconsistent with BWRVIP-25. The results of the stress evaluations for three different scenariosin accordance with BWRVIP-25 Appendix A are summarized. The three scenarios consideredby VYNPS are as follows.i) Loads on the core plate bolts taking no credit for the aligner pins. In this case, thebolts take all of the horizontal and vertical loads.ii) Shear load on the aligner pins with no credit for horizontal restraint from bolts. In thiscase, the bolts take vertical loads and the aligner pins take all of the horizontal loads.iii) Loads on the core plate bolts with no credit for aligner pins. This case also assumesthe stiffener beam to rim weld cracked. In this case, the core plate bolts take all ofthe horizontal and vertical loads.The staff's review of the three scenarios considered in VYNPS core plate bolts analysisindicates that the scenarios considered are acceptable because they are consistent with thethe scenarios discussed in Appendix A of BWRVIP-25 topical report that was previouslyreviewed by the staff. These scenarios represent the most limiting conditions for the coreplate bolts.3.2.3.1 LoadsThe stress evaluation of the core plate bolts included the effects of dead weight (DW), Fluiddrag load due to reactor internal pressure difference (RIPD) across core plate for normal andfaulted conditions, Seismic loads from operating basis earthquake and safe shutdownearthquake (OBE, and SSE), Fuel Lift load (FL), and bolt preload. DW of the core plateassembly is a vertical downward load. The seismic loads OBE & SSE are calculated based onVermont Yankee seismic accelerations and act in both horizontal and vertical directions. Thefluid drag load RIPD is an upward load on core plate bolts. The fuel lift load FL is an upwardload considered for the faulted condition. Friction at the interface of core shroud ledge and coreplate rim is also considered. Safety relief valve (SRV) actuation loads and torus induced loss ofcoolant (LOCA) accident loads are not significant for Vermont Yankee because the torus anddrywell are not substantially coupled for Mark I type containment. The annulus pressurization(AP) load is not part of VYNPS design basis, and is not a significant. The acoustic load (AC)resulting from the initial transient phase from a double ended guillotine break of the recirculationsuction line (RSL) is very abrupt relative to the shroud inertia and frequencies and therefore hasinsignificant effect on the shroud. The steady state portion of the load from RSL break affectsthe shroud and components external to the shroud. The core plate being inside of the shroud isessentially unaffected by the RSL break steady state load. The staff's review finds that thelicensee appropriately considered the applicable loadings in the structural evaluation of the coreplate bolts.-OFflCIAL USE ONLY -FROFRi~i AI~Y INhU~iv1ATlON

-12-3.2.3.2 Load Combinations and Acceptance criteriaThe VYNPS core plate bolt stress analysis utilized the criteria for allowables in accordance withthe Updated Final Safety Analysis Report (UFSAR, appendix section C.2, Ref.15), and ASMEBoiler and Pressure Vessel (ASME B&PV) Code,Section III (Ref. 16). The material propertiesfor the core plate bolts and the aligner pins are based on type 304 austenitic stainless steel ofRef. 16. The staff notes and accepts that ASME B&PV Code is not mandatory for the design ofthe VYNPS reactor vessel internals due to the vintage of the plant. However, the licenseecommitted to meet the intent of the ASME B&PV Code as described in UFSAR (Ref. 15).The staff's review determined that the licensee utilized for Normal & Upset, emergency, andfaulted condition general membrane stress allowables of 1SI, 1.5S=, and 2Sm respectively,where Sm is the allowable stress intensity of the material. The licensee utilized for Normal &Upset, emergency, and faulted condition, membrane plus bending stress allowables of 1.5S=,2.25Sm, and 3Sm respectively The licensee utilized for Normal & Upset, emergency, and faultedcondition, shear stress allowables of 0.6 Sm, 0.9Sm, and 1.2Sm respectively. Based on thereview of the licensee's stress evaluations for the core plate hold down bolts and aligner pin, thestaff concludes that the acceptance criteria are in accordance with the ASME B&PV Code, andUFSAR commitment3.2.3.3 Stress EvaluationsThe Vermont Yankee core plate design contains 30 core plate bolts of 2 inch diameter each andfour vertically oriented aligner pins of 2.625 inch diameter The finite element (FE) model usedfor the core plate assembly is not exactly VYNPS plant specific but is based on FE model inAppendix-A of BWRVIP-25. In response to an RAI for not having Vermont Yankee plantspecific FE model, the licensee provided justification that the analysis is linear and the resultsare appropriately scaled to account for the plant specific items. The staff reviewed the VYNPSplant specific items provided in the licensee's response that the licensee considered for scalingthe results. The stress evaluations for the VYNPS core plate bolts and aligner pins consideredappropriately for scaling the BWRVIP-25 analysis results based on Vermont Yankee geometryitems that include the number of bolts, size of core plate components, bolts and aligner pins,Vermont Yankee loadings, and Vermont Yankee specific bolt relaxation due to fluence andthermal effects.In response to an RAI (Ref. 14) on the justification of friction in Vermont Yankee calculations,the licensee stated that ignoring friction is overly conservative. The staff reviewed and agreeswith the licensee's justification that not considering friction at the interface of the core plate rimand shroud ledge because (i) the frictional resistance in a clamped connection of this type with alarge clamping force has significant friction, and (ii) the licensee used a smaller frictionalcoefficient of 0.2 to be conservative compared with GEH tests (Ref. 6) that determined africtional coefficient close to 0.5 for 304 (stainless steel (SS) sliding on 304 SS withdeoxygenated water as a lubricant.In its core plate bolt evaluations, the licensee appropriately accounted for bolt preload relaxationof 14 percent from neutron fluence due to 60 year plant life (see SE Section 3.2.1), and 6.2percent relaxation from modulus of elasticity decrease due to temperature effect between 700 F~FFIOIML USE ONLY -t-'I~<UPl-<lL I AI-<Y IN~U~MATION Q¥,FFCIAL USE ONLY -2'ROFRIETAR'1 INFORMATION-13-and 550' F. The preload loss from fluence is based on conservative fluence that uses peakfluence at azimuthal location for all bolts, and the use of the highest axial fluence at the bottomof active fuel for all bolts. The preload on core plate bolts is accounted for by adding themembrane stress due to preload to the calculated membrane stress, which is consistent withBWRVIP-25 Appendix-A.The licensee performed evaluations for core plate bolts for the Normal & Upset (DW+NormalRIPD+OBE), Emergency DW+Normal RIPD+SSE), and Faulted load combinations DW+FaultedRIPD+SSE+FL), and summarized the results for the bounding faulted load combinations for thethree scenarios described above. The licensee considered the applicable loads anddemonstrated that the membrane and membrane plus bending stresses in core plate bolts andthe shear stresses in the aligner pins satisfy the corresponding allowable criteria in the ASMEB&PV Code. The results show that the computed mean membrane stress is 12200 pounds persquare inch (psi) compared to its allowable of 32000 psi, and computed mean membrane plusbending stress of 41700 psi compared to its allowable of 48000 psi for the faulted conditioncases (i) when all the vertical and horizontal loads are taken by the core plate bolts with nocredit for aligner pins, and (ii).when all the vertical and horizontal loads are taken by the coreplate bolts with no credit for aligner pins, and the stiffener beam to rim weld cracked. Theresults also show that the shear stress in the aligner pin is 7700 psi compared to its allowable of19200 psi for the faulted condition case when the aligner pins take all of the horizontal loadswith no credit for horizontal restraint from bolts. The core plate stresses and aligner pin stressesare acceptable because they meet the respective allowables with some conservativeassumptions regarding friction, and preload relaxation.The staff requested the licensee to provide the cumulative usage factor (CUF) for the core platebolts for 60 year plant life. In response to an RAI (Ref. 14), the licensee demonstrated based ona simplified analysis that the alternating stress for the core plate bolts is only 1150 psi fromnormal & upset loadings and is well below the endurance limit of 25000 psi. The normal andupset cycles are less than 10000 cycles and the number of cycles for endurance limit is overone million. Thus, the CUF is negligible. Based on a review of this information, the conclusionthat the CUF is negligible for the core plate bolts, is acceptable to the staff.4.0 CONCLUSIONSWith respect to the effects of neutron irradiation on the core plate bolt properties, specifically theloss of preload determined by the licensee, the staff found the licensee's evaluation to beacceptable.With respect to the inspection plan propose by the licensee for the core plate bolts, the stafffinds the inspection plan as modified by the commitment contained in Attachment 2 to thelicensee's February 1, 2012 letter, to be-acceptable. Specifically, the licensee committed toinspect of 50% of the VYNPS core plate hold down bolts every other refueling outage,commencing with RFO 31, using the VT-3 [visual examination] method in accordance with theVYNPS Reactor Vessel Internals Inspection Program until BWRVIP-25 is revised. The licenseefurther committed to implement the revised BWRVIP-25 guidance for the core plate bolts....FFICIAL USE ONLYi'- UPUHI-I-TARY INFORMAl I U'N U ,, O vL PROPRIETaRy lN'O-- .......-14-With respect to the stress analysis of the core plate bolt, including the preload relaxation due tothermal effects and fluence for a 60 year life, the staff finds the licensee's evaluation acceptablebecause the core plate bolts satisfy the ASME B&PV Code criteria for the applicable loads andload combinations The methodology and assumptions utilized in stress analysis are reasonableand consistent with BWRVIP-25, and therefore are acceptable. The NRC staff concludes thatthere is reasonable assurance that the VYNPS core plate bolts are structurally acceptable for 60year plant life.References1. BWR Vessel and Internals Project BWR Core Plate Inspection and Flaw EvaluationGuidelines (BWRVIP-25), EPRI Report TR-1 07284, December 1996 (ProprietaryInformation. Not Publicly Available)2. Letter from Jack Strosnider to Carl Terry dated December 19, 1999, Subject: Final SafetyEvaluation of BWRVIP Vessel and Internals Project, "BWR Vessel and Internals Project,BWR Core Plate Inspection and Flaw Evaluation Guideline (BWRVIP-25)," EPRI Report TR-107284, December 1996 (TAC No. M97802) (ADAMS Accession No. ML993620267)(Proprietary Information. Not Publicly Available)3. Letter from Vaughn Wagoner to NRC dated July 17, 1997, Subject: License RenewalAppendix B to BWR Vessel and Internals Project BWR Core Plate Inspection and FlawEvaluation Guidelines (BWRVIP-25), EPRI Report TR-1 07284, December 19964. Letter from Christopher Grimes to Carl Terry dated December 7, 2000, Subject: SafetyEvaluation for Referencing of BWR Vessel and Internals Project, BWR Core PlateInspection And Flaw Evaluation Guidelines (BWRVIP-25) Report for Compliance With theLicense Renewal Rule (10 CFR Part 54) and Appendix 0, BWR Core Plate Demonstration ofCompliance with the Technical information Requirements of the License Renewal Rule (10CFR 54.21)5. Letter from Michael J. Colomb to NRC dated March 18, 2011, Subject: "Core Plate HoldDown Bolt Inspection Plan and Analysis," Vermont Yankee Nuclear Power Station DocketNo. 50-271 License No. DPR-28 (BVY 11-021) (ADAMS Accession No. ML110840068)6. NEDC-33618P -Revision 0, Vermont Yankee Core Plate Bolt Stress Analysis, March 2011,(ADAMS Accession No. ML1 10840070 -Proprietary Version, ML1 10840069 -Non-Proprietary Version)7. Safety Evaluation Report Related to the License Renewal of Vermont Yankee NuclearPower Station (NUREG-1907 Vol. 2), May 2008 (ADAMS Accession No. ML081430109)8. Letter from Christopher J. Wamser to NRC dated December 9, 2011, Subject: "Response toRequest for Additional Information for Core Plate Hold Down Bolt Inspection Plan andAnalysis, Vermont Yankee Nuclear Power Station Docket No. 50-271, License No. DPR-28(BVY 11-078) (ADAMS Accession No. ML11353A407)-O,,FICIAL- USE, ,,," ... -FPROFRiE. IAINY INhI-OMAT'IUN

,.OFiFrIClIL A I I I N PROPCI R rr r I 1! 11-15-9. BWRVIP-99-A, BWR Vessel and Internals Project -Crack Growth Rates in IrradiatedStainless Steels in BWR Internal Components 1016566, Final Report, October 2008 -Proprietary (ADAMS Accession No. ML091620165); Non-proprietary version BWRVIP-99NP-A (ADAMS Accession No. ML091620164)10. Materials Reliability Program: PWR Internals Material Aging Degradation MechanismScreening and threshold Values (MRP-175) 1012081, Topical Report, December 2005 -Proprietary (ADAMS Accession No. ML063470637); Non-proprietary version (ADAMSAccession No. ML061880278)11. Letter from Michael J. Colomb to NRC dated March 18, 2011, Subject: "Deviation fromBWRVIP-25 Inspection Requirements, Vermont Yankee Nuclear Power Station," Docket No.50-271, License No. DPR-28 (BVY 11-024) (ADAMS Accession No. ML1 10840044)12. Letter from Christopher J. Wamser to NRC dated February 1, 2012, Subject: "Response toRequest for Additional Information Regarding Core Plate Hold-down Bolt Inspection Planand Analysis, Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28 (BVY 12-008)(ADAMS Accession No. ML12037A066)13. NUREG-1475, Rev. 1,"Applying Statistics", March 201114. Letter from Christopher J. Wamser to NRC dated January 5, 2012, Subject: "Response toRequest for Additional Information Regarding Core Plate Hold-down Bolt Inspection Planand Analysis, Vermont Yankee Nuclear Power Station Docket No. 50-271 License No. DPR-28 (BVY 12-008)(ADAMS Accession No. ML120100126)15. Vermont Yankee Nuclear Power Station (VYNPS) Updated Final Safety Analysis Report(UFSAR), Appendix Section C.2.5.1, Rev. 24.16. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels, 1965.O.FFI ,I.L USE, ONLY -PROPRIETARY INFORMATION