ML17333B036
| ML17333B036 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/09/1997 |
| From: | FITZPATRICK E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1223E, NUDOCS 9709170108 | |
| Download: ML17333B036 (20) | |
Text
CATEGORY1IREQULA'1gINFORMATIONDISTRIBUTZO1lgTEM(RIDE)ACCESSION'3NBR:9709170108DOC.DATE:97/09/09NOTARIZED:YESDOCKETFACIL:50'-.316DonaldC.CookNuclearPowerPlant,Unit2,IndianaM05000316AUTH.NAMEAUTHORAFFILIATIONFITZPATRICK,E.IndianaMichiganPowerCo.(formerlyIndiana6MichiganEleRECIP.NAMERECIPIENTAFFILIATIONDocumentControlBranch(DocumentControlDesk)
SUBJECT:
Forwardsresponseto970709RAIre9607115%thermalpoweruprateAEP:NRC:1223submittal.DISTRIBUTIONCODE:A001DCOPIESRECEIVED:LTRENCLSIZE:TITLE:ORSubmittal:GeneralDistributionNOTESRECIPIENTIDCODE/NAMEPD3-3LAHICKMAN,JINTERNILECE1NRRDE/EMCBNRR/DSSA/SPLBNUDOCS-ABSTRACTEXTERNAL:NOACCOPIESLTTRENCL11111111111111RECIPIENTIDCODE/NAMEPD3-3PDNRR/DE/ECGB/ANRR/DRCH/HICBNRR/DSSA/SRXBOGC/HDS2NRCPDRCOPIESLTTRENCL111111111011E0DUNOTETOALL"RIDS"RECIPIENTS:PLEASEHELPUSTOREDUCEWASTE.TOHAVEYOURNAMEORORGANIZATIONREMOVEDFROMDISTRIBUTIONLISTSORREDUCETHENUMBEROFCOPIESRECEIVEDBYYOUORYOURORGANIZATION,CONTACTTHEDOCUMENTCONTROLDESK(DCD)ONEXTENSION41S-2083TOTALNUMBEROFCOPIESREQUIRED:LTTR13ENCL12 i~
indianaMichiganPowerCompany~500CircleDriveBuchanan,Ml4910713955~iIJrtINtIANSlSIICNIGANPQWMSeptember9,1997AEP:NRC:1223EDocketNo.:50-316U.S.NuclearRegulatoryCommissionATTN:DocumentControlDeskWashington,D.C.20555Gentlemen:DonaldC.CookNuclearPlantUnit2RESPONSETOREQUESTFORADDITIONALINFORMATIONREGARDINGPOWERUPRATEANDRELATEDCHANGESThisletteranditsattachmentconstitutearesponsetotheJuly9,1997,NRCrequestforadditionalinformationregardingourJuly11,1996,5%thermalpoweruprateAEP:NRC:1223submittal.Therequestforadditionalinformationprimarilyinvolvesanalysisassumptionsandmethodology.Thisletterissubmittedpursuantto10CFR50.30(b)and,assuch,includesanoathstatement.Sincerely,E.E.FitzpatrickVicePresidentSWORNTOANDSUBSCRIBEDBEFOREMEmyrrhTHIS7DAYOFo~&~gP1997NotaryPublic~/-/-4/vlbAttachmentUNDALBOEI.CKENotaryPublic,BerrienCounty,MlMyCommissionExpiresJanuary21,2001A.A.BlindA.B.BeachMDEQ-DW&RPDNRCResidentInspectorJ.R.Padgettrtr->a('.,~sAtsr%s'st70'sti70i08'st70'st0'stPDRADOCK050003i6PPDRllllllllllllllllllllllllllllllllllllllll ttCFII.IIC~
ATTACHMENTTOAEP:NRC:1223EDonaldC.CookNuclearPlantUnit2RESPONSETOREQUESTFORADDITIONALINFORMATIONREGARDINGPOWERUPRATEANDRELATEDCHANGES AttachmenttoAEP:NRC:1223EPage1NRCUESTIONNO.1"InSection2.0ofReference2,youindicatedthatWCAP-11902andSupplementwereusedasthebasisfortheevaluationoftheUnit2operationatcorepowerlevelof3588MWt.However,WCAP-11902licensingreportwasreviewedandapprovedbythestaff,forD.C.CookUnit1operatingat3250MWt.ClarifywhethertheSupplementtoWCAP-11902,entitled,"ReratedPowerandRevisedTemperatureandPressureOperationforCookNuclearPlantUnits1and2LicensingReport,"wasreviewedandapprovedbythestaffforapplicationattheCookNuclearPlant(CNP).Ifnot,statethebasisofapplyingthesetwopreviousevaluationsforallperformanceparametersbetweentheproposedUnit2uprateandthepreviousreratedprogram."RESPONSETOUESTIONNO.1Attachment5toAEP:NRC:1223submittal,fromE.E.FitzpatricktotheUSNRCdocumentcontroldesk,datedJuly11,1996,is"DiscussionofPreviousRelatedSubmissions."Theintroductionsectionofattachment5addresses,inageneralway,thefactthattheanalysesthatsupporttheproposedupratinghavebeenperformedoveraperiodofyearsasapartofothereffortswithmoreimmediateshortrangegoals.Thisattachmentstates:"TheanalysesthatsupporttheproposedupratingofDonaldC.CookNuclearPlantUnit2havebeenperformedoveraperiodofyearsinseveralcontexts.Theanalysisofthenuclearsteamsupplysystem(NSSS)foranNSSSpowerof3600MWtwasperformedinconjunctionwithanalysestooperateunit1atreducedtemperatureandpressure(the"ReratingProgram").Mostofthecoreresponseanalyseswereperformedatanupratedcorethermalpowerof3588MWtasapartofthetransitionfromAdvancedNuclearFueltoWestinghouseVantage5fuel.Therecentlysubmittedanalyses,AEP:NRC:1207(erroneouslystatedtobeAEP:NRC:1223inthesubmittal),tosupportanincreaseinthepermittedlevelofsteamgeneratortubepluggingforunit1includesasteammassandenergyreleaseanalysistothecontainmentwhichboundsbothunitsat3600MWt.Forthissubmittal(i.e.,AEP:NRC:1223),previousNSSSanalysesandcoreresponseanalyseshavebeenreviewed,newanalyseshavebeenperformedwherenecessary,andthebalanceofplantevaluated,asdescribedwithinthissubmittal,tosupporttheproposaltoincreasethecoreratedthermalpowerto3588MWt."Inparticular,asindicatedinattachment5,thesupplementtoWCAP11902wassubmittedinpartinsupportofanumberofproposedtechnicalspecification(T/S)changes.Itwassubmittedinitsentiretyinsupportofourproposaltoreducetheboronconcentrationintheboroninjectiontanksofbothunitsto0ppm.OursubmittalwasletterAEP:NRC:1140,"TechnicalSpecificationChangeRequest,BoronInjectionTank(BIT),BoronConcentrationReduction",fromM.P.AlexichtoT.E.Murley,datedMarch26,1991.TheproposalwasapprovedbyAmendmentNo.158toFacilityOperatingLicenseNo.DPR-58'ndAmendmentNo.142toFacilityOperatinglicenseNo.DPR-74.
AttachmenttoAEP:NRC:1223EPage2NRCUESTZONNO.2"Clarifywhetherthereratinganalysesofthepressuretransientsandthepostulatedloss-of-coolantaccident(LOCA)includetheproposedpressurizersafetyandreliefvalvetolerance+/-3%,andthepreviouslyNRC-approvedmainsteamsafetyandreliefvalvestoleranceof+/-3%.Zfnot,statehowthereratinganalysesappliestotheproposedUnit2poweruprate."RESPONSETOUESTZONNO.2TheanalysesperformedforsubmittalAEP:NRC:1223,toincreasethethermalpowerofCookNuclearPlantunit2to3588MWt,assumedsetpointtolerancesof3%forboththepressurizersafetyvalvesandthesteamgeneratorsafetyvalves.Thepressurizersafetyvalvesetpointtoleranceisspecificallyaddressedfortheapplicableanalysesinsection3.3,"Non-LOCAAnalyses",ofWCAP-14489,attachment6tosubmittalAEP:NRC:1223.Thisassumptioniscalledoutspecificallyfortheapplicableeventsbecausethisisanewassumptionfortheunit2analyses.ThepressurizerpressuresetpointdoesnotaffecttheLOCAeventbecausetheprimarysystemdepressurizes.Theassumptionofa3%toleranceforsteamgeneratorsafetyvalvesetpointswasnotspecificallycalledoutforthenewanalysesbecauseitisanassumptionthatwaspreviouslysubmittedandreviewed.Anassumptionof3%setpointtoleranceforsteamgeneratorsafetyvalvesetpointsisinputtotheapplicableanalysesintheunit2upratesubmittal.NRCUESTZONNO.3"Discusstheoperabilityofthesafety-relatedmechanicalcomponents(i.e.,valvesandpumps)affectedbythepowerupratetoensurethattheperformancespecificationsandtechnicalspecificationrequirements(e.g.,flowrate,closeandopentimes)willbemetfortheproposedpoweruprate.Confirmthatthesafety-relatedmotoroperatedvalves(MOVs)willbecapableofperformingtheirintendedfunctionsfollowingthepoweruprateincludingsuchaffectedparametersasfluidflow,temperature,pressureanddifferentialpressure,andambienttemperatureconditions.Zdentifymechanicalcomponentsforwhichoperabilityattheupratedpowerlevelcouldnotbeconfirmed."RESPONSETOUESTZONNO.3AFWCCWANDESWSYSTEMSThesafetysystemswereviewedforimpactfromupratedconditionsaretheauxiliaryfeedwater(AFW),componentcoolingwater(CCW),andessentialservicewater(ESW)systems.Ourreviewindicatesthatthemechanicalcomponents(i.e.,valvesandpumps)inthesesystemsarenotsignificantlyaffectedbytheupratedpowerconditions.TheperformanceandT/Srequirementsforthesesystemsremainunchanged.Becausethesystemparametershavenotchanged,theassociatedMOVoperabilityisnotimpacted.Thefollowingsummarizesourreviewinsupportoftheprecedingstatementfortheindicatedsystems.TheAFWsystemprovideswatertothesteamgeneratorswhenthemainfeedwater,systemisunavailableduetoalossoffeedwater,unit AttachmenttoAEP:NRC:1223EPage3trip,feedwaterorsteamlinebreak,lossofoffsitepower,orloss-of-coolantaccident(LOCA).TheAFWsystemisdesignedandanalyzedtoprovidesufficientflowtothesteamgeneratorsduringtheseeventsagainstasteamgeneratorpressurecorrespondingtothesetpressure,plusaccumulationofthelowestsetsafetyvalves.TheAFWsystemisalsocapableofprovidingreducedflowatthehighersteamgeneratorpressures,plusaccumulationcorrespondingtothehighersetsafetyvalves.TheupratedconditionsdidnotaltertheAFWsystem'sflowrequirementsorthesystem'sabilitytofulfilltheserequirements.Theupratedconditionsdidnotaffectorrevisethesafetyvalve'ssetpressure,theAFWpump'soperatingparameters(flowandhead),orthefluidparameters(temperatureandpressure).Theupratealsodidnotresultinanysignificantchangesinambienttemperatures.Therefore,theAFW'sMOVrequirementsareessentallyunchanged,andthemechanicalcomponentsinthesystemarenotsignificantlyaffected.TheCCWsystemisaclosedloopsystemthatservesasanintermediateloopbetweenpotentiallyradioactivesystemsandlakewatertoensurethatleakageofradioactivefluidiscontainedwithintheplant.TheCCWsystemisdesignedandanalyzedtosupplycoolingwaterflowduringtheinjectionandrecirculationphasesofaLOCAandduringunitoperation.TheLOCAlong-termmassandenergyreleaseandcontainmentintegrityanalysesperformedbyWestinghouseutilizedCCWsystemflowratesandheatexchangerUAsrepresentativeoftheupratedconditions.TheWestinghouseanalysesdeterminedtheresultswereacceptableforcontainmentintegritypressureandtemperatureresponse.ThesedetailswereprovidedinoursubmittalAEP:NRC:1223C,datedJune10,1997.Basedonthis,theupratedconditionsdidnotsignificantlyimpacttheCCWsystem'sheatremovalrequirements,orthesystem'scapabilitytomeettheserequirements.TheCCWpumps'peratingparameters(flowandhead)andfluidparameters(temperatureandpressure)werenotchangedasaresultoftheuprate.Theupratealsodidnotresultinanysignificantchangesinambienttemperatures.Therefore,theCCW'sMOVrequirementsareessentiallyunchangedandthemechanicalcomponentsinthesystemarenotsignificantlyaffected.TheESWsystemprovidescoolingwaterrequirementstotheCCWheatexchangers,emergencydieselgenerators,CTSheatexchangers,andthecontrolroomairconditioningcondensers.TheESWsystemisoperatedinconjunctionwiththeCCWandCTSsystems.TheESWpump'soperatingparameters(flowandhead)andfluidparameters(temperatureandpressure)werenotchangedasaresultoftheuprate.Theupratealsodidnotresultinanysignificantchangesinambienttemperatures.Therefore,theESW'sMOVrequirementsremainessentiallyunchangedandthemechanicalcomponentsinthesystemarenotsignificantlyaffected.RCSCVCSANDRHRSSYSTEMSThesafetysystemstobereviewedforimpactfromupratedconditionsarethereactorcoolantsystem(RCS),emergencycorecoolingsystem(ECCS),andchemicalvolumecontrolsystem(CVCS).Ourreviewindicatesthatthemechanicalcomponentsinthesesystemsarenotsignificantlyaffectedbytheupratedpowerconditions.TheperformanceandT/Srequirementsforthesesystemsremainunchanged.Becausethesystemparametershavenotchanged,theassociatedMOVsoperationisnotsignificantlyimpacted.
AttachmenttoAEP:NRC:1223EPage4TheRCSconsistsoffouridenticalheattransferloopsconnectedinparalleltothereactorvessel.Eachloopcontainsareactorcoolantpump(RCP)andasteamgenerator.Inaddition,thesystemincludesapressurizer,apressurizerrelieftank,inter-connectingpiping,andinstrumentationnecessaryforoperationalcontrol.Duringoperation,theRCPscirculatepressurizedwaterthroughthereactorvesselandthefourcoolantloops.Thewater,thatservesbothasacoolant,moderator,andsolventforboricacid(chemicalshimcontrol),isheatedasitpassesthroughthecore.Itthenflowstothesteamgeneratorswheretheheatistransferredtothesteamsystem,andreturnstotheRCPstorepeatthecycle.TheRCSpressureiscontrolledbytheuseofthepressurizerwherewaterandsteamaremaintainedinequilibriumbyelectricalheatersandwatersprays.Threespringloadedsafetyvalvesandthreepoweroperatedreliefvalvesareconnectedtothepressurizeranddischargetothepressurizerrelieftank,wherethesteamiscondensedandcooledbymixingwithwater.Fluidsystemscalculationswereperformed,evaluatingthecapabilityoftheRCStooperateattheuprateprogramconditions.TheupratedpowerconditionsdidnotaffectanyoftheRCSsafetyrelatedmechanicalcomponentsdesignbasis.TheMOVsfluidsystemdesignconditions(fluidflow,temperature,pressureanddifferentialpressure)werenotsignificantlyaffectedbytheupratedconditions.TheCVCSprovidesforboricacidaddition,chemicaladditionsforcorrosioncontrol,reactorcoolantclean-upanddegasification,reactorcoolantmake-up,reprocessingofwaterletdownfromtheRCS,andRCPsealwaterinjection.Duringplantoperation,reactorcoolantflowsthroughtheshellsideoftheregenerativeheatexchanger,thenthroughaletdownorifice.Theregenerativeheatexchangerreducesthetemperatureofthereactorcoolant,andtheletdownorificereducesthepressure.Thecooled,lowpressurewaterleavesthereactorcontainmentandenterstheauxiliarybuilding.Asecondtemperaturereductionoccursinthetubesideoftheletdownheatexchanger,followedbyasecondpressurereductionduetothelowpressureletdownvalve.Afterpassingthroughoneofthemixedbeddemineralizers,whereionicimpuritiesare,.removed,coolantflowsthroughthereactorcoolantfilterandentersthevolumecontroltank(VCT).TheregenerativeandletdownheatexchangersaredesignedtocoolletdownflowfromT,~to115'.ThevariationsinT,~consideredfortheuprateprogramareboundedbythedesigninlettemperatureof547'fortheregenerativeheatexchanger.Therefore,thecoolingrequirementsoftheletdownfunctionaremetwiththerevisedoperatingparameters.TheletdownfunctionisdesignedtoreducethestaticpressureofthereactorletdownstreamfromtheRCPsuctionpressuretoVCToperatingpressure,suchthatthedesignpressureofinterveningpipingandcomponentsisnotexceeded,andfluidismaintainedinasubcooledconditionthroughoutthesystem.Thepressurereductionxequirementsoftheletdownfunctionaremetwiththerevisedoperatingparameters.Thecentrifugalchargingpumpoperatingconditionshavenotbeenimpactedbytheupratingconditions.FluidsystemscalculationswereperformedevaluatingthecapabilityoftheCVCStooperateattheuprateprogramconditions.Theupratedpowerconditionsdonot AttachmenttoAEP:NRC:1223EPage5significantlyaffecttheCVCSsafetyrelatedmechanicalcomponents'esignbases.TheECCSinjectsboratedwaterintothereactorfollowingabreakineitherthereactororsteamsystemsinordertocoolthecoreandpreventanuncontrolledreturntocriticality.Twosafetyinjection(SI)pumpsandtworesidualheatremovalpumpstakesuctionfromtherefuelingwaterstoragetank(RWST)anddeliverboratedwatertofourcoldlegconnectionsviatheaccumulatordischargelines'.Inaddition,twocentrifugalchargingpumpstakesuctionfromtheRWSTonSIactuationandprovideflowtotheRCSviaseparateSIconnectionsoneachcoldleg.AtthecompletionoftheinjectionphasefromtheRWSTtheECCSisthenalignedtothecontainmentsump,asthesuctionsource,toprovidethecoldorhotlegrecirculationinjectionflows.Theprimarysystempressuresconsideredforthisprogramarelessthan,orequalto,theprimarysystempressureagainstwhichtheoriginalsystemwasdesignedtodeliver.Therefore,therevisedprimarysystemparametersdonotrequireanincreaseineitherthemotivepressureorcorecoolingcapacityoftheECCS.FluidsystemscalculationswereperformedevaluatingthecapabilityoftheECCStooperateattheuprateprogramconditions.TheupratedpowerconditionsdidnotsignificantlyaffecttheECCSsafetyrelatedmechanicalcomponents'esignbases.NRCUESTIONNO.4"InreferencetoSections3.11.2and3.11.3ofreference2(WCAP-14489),providethemaximumcalculatedstressesandcumulativeUsageFactorsatthemostlimitinglocationsandcomponentsofthereactorvesselandinternals,steamgenerator,reactorcoolantpump,pressurizer,andcontrolroddrivemechanism.Alsoprovidetheallowablecodelimits,thecode,andthecodeeditionusedintheevaluationforthepoweruprate.Ifdifferentfromthecodeofrecord,providethenecessaryjustification."RESPONSETOUESTIONNO.4ReactorVessel:Withrespecttosection3.11.2,theresultsofthereactorvesselanalysesandevaluationsaresummarizedbelow.Thestressintensityandfatigueusagelimits(withtheexceptionofthe3Smaximumrangeofprimaryplussecondarystressintensitylimitforthecontrolroddrivemechanism(CRDM)housingsandoutletnozzlesafeend)oftheASMEBoilerandPressureVesselCode,SectionI1I,1968Edition,withAddendathroughtheSummerof1968,aremet.Theexceedingofthe3SlimitfortheCRDMhousingsandoutletnozzlesafeendisreconciledbyusingtheASMEcodeacceptablemethodofelastic-plasticanalysesinaccordancewithASMEBoilerandPressureVesselCode,SectionIII,1971Edition.CRDMHousinThemaximumrangeofprimaryplussecondarystressintensityiscalculatedtobe77.76ksi,whichexceedsthe3Slimitof69.9ksi.However,asimplifiedelastic-plasticanalysiswasperformedinaccordancewithparagraphNB-3228.3oftheASMEBoilerandPressureVesselCode,SectionIII,1971Edition,andthehigherrangeofstressintensityisreconciled.Themaximumcumulative AttachmenttoAEP:NRC:1223EPage6fatigueusagefactoris0.1687,whichisbelowtheASMEcodelimitof1.0.MainClosureReionThemainclosureregionofthereactorvesselconsistsofthevesselflange,theclosureheadflange,andtheclosurestudassembliesthatcoupletheheadtothevessel.Themaximumrangesofstressintensityintheclosureheadflangeandthevesselflangeare65.26ksiand61.04ksi,respectively,comparedtotheASMEcode3Slimitof80.1ksi.Themaximumserviceintheclosurestudsis91.8ksi,whichcomparesfavorablytothe3Slimitof107.7ksi.Themaximumcumulativefatigueusagefactorfortheclosureheadflange,vesselflangeandclosurestudsare0.018,0.029and0.99,respectively.Theusagefactorsarealllessthanthe1.0ASMEcodelimit.However,itshouldbenotedthattheclosurestudusagefactorof0.99wascalculatedundertheassumptionthatthefirst25%ofthe11,680occurrencesofplantloadingandunloading,at5%offullpowerperminute(2,920occurrencesofeach),occurredduringthefirsttenyearsofoperationwhenthevesseloutlettemperature(T)was599.3'.OutletNozzleThemaximumrangeofprimaryplussecondarystressintensityintheoutletnozzleendiscalculatedtobe59.58ksicomparedtothe3Slimitforausteniticstainlesssteelmaterialof50.1k-i.Becausethemaximumrangeofstressintensityexceeds3S,asimplifiedelastic-plasticanalysisperparagraphNB-3228.3oftheASMEBoilerandPressureVesselCode,SectionIII,1971Edition,wasperformedthatjustifiedthehighermaximumrangeofstressintensity.Themaximumusagefactoratthesafeendis0.021,whichislessthan1.0.Themaximumrangeofstressintensityintheoutletnozzleandnozzletoshelljunctureis57.09ksi,comparedtothe3Sallowable80.1ksi.Themaximumcumulativeusagefactorinthenozzleandnozzletoshelljunctureis0.0631,whichisalsolessthan1.0.InletNozzleThemaximumrangeofstressintensityintheinletnozzlesafeendis49.65ksi,whichislessthan3S=50.1ksi.Themaximumrangeofstressintensityintheinletnozzleandnozzletoshelljunctureis49.86ksi,whichcomparesfavorablywitha3Slimitof80.1ksi.Themaximumcumulativeusagefactorsinthenozzlesafeendandnozzletoshelljunctureare0.0174and0.0977,respectively,whicharebothlessthan1.0.VesselWallTransitionThemaximumrangeofstressintensityandcumulativefatigueusagefactorforthevesselwalltransition,betweenthenozzleshellandthevesselbeltline,are33.57ksiand0.0066.ThesevaluesarelessthantheASMEcodelimitsof80.1ksiand1.0,respectively.
AttachmenttoAEP:NRC:1223EPage7BottomHead-to-ShellJunctureThemaximumrangeofprimaryplussecondarystressintensityatthejuncture,betweenthevesselbottomhemisphericalheadandthevesselbeltlineshell,is34.53ksicomparedtoa3Sallowableof80.1ksi.Themaximumcumulativefatigueusagefactoratthejuncturewascalculatedtobe0.0182,whichislessthan1.0.BottomHeadInstrumentationPenetrationsThebottomheadinstrumentationpenetrationsareacceptableforuprating,baseduponamaximumrangeofprimaryplussecondarystressintensityof51.49ksi,andamaximumcumulativeusagefactorof0.1220.ThesevaluescomparefavorablywiththeASMEcodeallowablesof69.9ksi(3S)and1.0,respectively.CoreSuortPadsThecoresupportpadswereevaluatedtohaveamaximumrangeofstressintensityof69.7ksi,comparedtoa3Slimitof69.9ksi.Themaximumcumulativefatigueusagefactorwascalculatedtobe0.693,whichislessthanthe1.0ASMEcodelimit.ReactorVesselInternalsCookNuclearPlantunit2reactorinternalsarecomposedoftwosections,theupperinternalsandthelowerinternals.Evaluationswereperformedforthecriticalcomponentsforboththeupperinternalsandlowerinternals.Thefollowingisalistofthecriticalcomponentsfortheupperandlowerinternals.UerInternalsPerforatedsectionofthetophatsupportstructure.LowerInternalsLowerSupportAssemblyCoreBarrelandFlangeLowerRadialSupportClevisInsertsBaffle-FormerAssemblyUpperCorePlateAlignmentPinsThermalShieldThestructuralevaluationsperformedfortheaboveareasconfirmedthattheirstructuralintegrityandincreasedfatigueusagewasfoundtobewithinacceptablelimits,accordingtotheoriginaldesignbasis.SteamGenerator:Theunit2steamgeneratorswerereplacedin1987.Thediscussionbelowaddressesthereplacedcomponentsandremainingoriginaluppershellcomponentsseparately.
~AttachmenttoAEP:NRC:1223EPage8RelacementComonentsThecriteriausedtodetermineacceptablestressstatesareprovidedintheASNEBoilerandPressureVesselCode,SectionIII,1968Edition,andtheassociatedAddendathroughWinter1968.ComponentMaximumStressCalcu-latedMaximumStressAllow-ableFatigueUsageCalcu-latedFatigueUsageAllow-ablePrimaryChamber,Tube-sheet,StubBarrelPrimaryNozzles31.9ksi58.2ksi0.130.871.01.0PrimaryManways41.0ksi48.3ksi0.911.0Tubes47.96ksi79.80ksi0.591.0PrimaryChamberDividerPlate0.191.0TubetoTubesheetWeldLowerShell/Cone/UpperShellTrunnions79.2ksi58.8ksi80.1ksi80.1ksi0.750.120.011.01.01.0MinorBoltedOpenings93.9ksi94.3ksi0.741.0MinorNozzlesInternalsFeedwaterRingandJ-Nozzles29.3ksi(2)26.7ksi80.1ksi(2)27.0ksi0.880.060.561.01.01.0(1)Theprimary+secondarystressesexceedtheallowablestresslimitof3S.AplasticanalysiswasperformedperparagraphN-417.6(b)oftheASMEBoilerandPressureVesselCode,SectionIII,"NuclearVessels",1968EditionwithAddendatoandincludingWinter1968,codeofrecord,todemonstratestructuralintegrity.(2)Themaximumstressesinthesteamgeneratorinternalsoccurduringthefaultedconditions.Forthenormalandupsetconditions,theprimary+secondary+peakstressesinthesteamgeneratorinternalsarelow,andbelowtheendurancelimit.Therefore,themaximumfatigueusageforthesteamgeneratorinternalsis0.06.
AttachmenttoAEP:NRC:1223EPage9OriinalUerShellComonentsPrimarystressesandmaximumstressrangesarenotaffectedbytheupratingconditions,andthesecalculationswerenotrepeated.Whenconsideringtheupperandlowerboundprimarytemperatures,theupperboundtemperatureconditionsareveryclosetothetransientconditionsusedinthereferenceanalyses,andtheresultingfatigueusagesshowonlyslightvariationsfromthereferenceconditions.However,thelowerboundtemperatureconditionscanresultinincreasedfatigueusagesinsomecases.Asummaryofthefatigueusagesisprovidedbelow.ComponentReferencedFatigueUsageUpperBoundTemperatureFatigueUsageLowerBoundTemperatureMainFeedwaterNozzleSecondaryManwayShellPenetration0.530.170.7240.0510.9410.053SecondaryManwayBolts(3)0.4270.825SteamNozzle0.590.6160.616(3)Thereferencevalueforfatigueisnotprovided.Thestressesusedfortheanalysisoftheboltsaretakenfromanothermodelsteamgenerator,withscalefactorstoaccountforgeometryvariations.Aspartoftheupratingprogram,thesteamgeneratorstructuralintegritywasevaluatedtoaccountfortherevisedlossofloadandlossofoffsitepowertransients.Theevaluationshowedthatthecomponentmostaffectedbytheupratingprogramisthetubesheet-to-channelheadjunction.Thestressintensitiescontinuetosatisfythestresslimits.Thecalculatedvalueofthefatigueusage,0.34,remainswithinthemaximumallowablelimitof1.0.ReactorCoolantPumTheevaluationperformedfortheRCPsaddressedtheASMEcodestructuralconsiderationsfortheRCPcasing,mainflange,mainflangebolts,thermalbarrier,casingfoot,casingdischarge,andsuctionnozzles,casingweirplate,sealhousing,andauxiliarynozzles.Forunit2theASMECode,SectionZZI,1968Edition,withAddendathroughSummer1969,wasusedasaguide.TheRCPevaluationaddressedtherevisedNSSSparametersandNSSSdesigntransientsassociatedwiththeuprating,andcomparedtheseparametersandtransientstotheconditionsassumedintheoriginaldesignanalysesfortheRCPs.Thedifferences(i.e.,deltatemperatures[DTs]anddifferentialpressures[DPs])wereidentifiedandusedtoobtainstressand,fatigueresultsforpoweruprate.TheDPsassociatedwiththepowerupratedesigntransientswerereviewedtodetermineiftherewereanychangesthatwouldqualify AttachmenttoAEP:NRC:1223EPage10asa"significantfluctuation"inaccordancewiththeASMEcodedefinition,and,thus,requireconsiderationrelativetofatigue.Itwasconcludedduetothepowerupratedesigntransients,thatallDPswerelessthantheASMEcodedefinitionof"significantfluctuation"value,andthatnoratigueconsiderationisrequiredbecausethefatiguewaiverremainsunchanged.ThedesigntransientswerethenreviewedtoidentifythemaximumpressuretowhichtheRCPcouldbeexposed.Forunit2,thismaximumpressurewasdeterminedtobe2724.1psiaforthelossofloadtransient.AreviewofRCPanalysesperformedforotherplantsshowedthatincreasesto2725psiahavebeenanalyzed.indetailandshowntobeacceptable.Itwasconcludedthatthepressuretransientsareacceptable.TheeffectofpoweruprateonthevariousoriginalanalysesfortheRCPswasalsoassessedusingtheNSSSdesigntransientsandtheassociatedDTvalues.Forthemostpart,thecomparisonofNSSSdesigntransientsandassessmentsofassociatedDTvaluesweresufficienttoshowcontinuedapplicabilityoftheoriginalanalysestopoweruprateconditions.OneareawheretheincreaseinDTwassufficienttomeritanalysiswasforthecasingweirplate.Theevaluationshowedarangeofstressintensities=41,379psiforpoweruprateconditions.ComparisonothisvaluetotheASMEcodeprimaryplussecondarystresslimitof3S=50,700psishowedthattheASMEcodelimitissatisfied.Fatiguerequirementsfortheweirplateweresatisfiedbythefatiguewaiver(ASMEcode,NB-3222.4(d)).Insummary,theresultsofthepoweruprateassessmentsshowedthattheASMEcodecriteriaaresatisfiedatpoweruprateconditions.Pressurizer:Theexternalloadsarenotrevisedforthe3600MWtupratingconditions,andthechangesinthepressureloadsdonotaffectthepreviouslycompletedstresscalculations.Thus,theprimarystressescalculatedfortheoriginalanalysisremainvalidattheupratedconditions.Also,thechangesinthedesigntransients(lossofloadandlossofoffsitepower)didnothaveanysignificanteffectontheprimaryplussecondarystresses.However,forsomecomponents,thefatigueanalysisisaffected.Thenewcalculatedfatigueusagefactorsforeachofthepressurizercomponentsarelistedbelow.Becausethenewcalculatedfatigueusagefactorsarelessthan1.0,thepressurecomponentsmeetthestress/fatiguerequirementsoftheASMECode,SectionIII,1965Edition,includingAddendauptoWinter1966.PRESSURIZERFATIGUEUSAGEFACTORS~ComonentSurgeNozzleSprayNozzleSafetyandReliefNozzleLowerHead,HeaterWellLowerHead,PerforationUpperHeadandShellSupportSkirt/FlangeManwayPadManwayCoverManwayBoltsCalculatedFatiueUsae<0.340.991<0.15<0.07<0.020.973<0.020.00.00.0
~AttachmenttoAEP:NRC:1223EPage11SupportLugInstrumentNozzleZmmersionHeaterValveSupportBracket<0.05<0.11<0.010.01ControlRodDriveMechanism:TheevaluationperformedfortheCRDMsaddressedtheASMEcodestructuralconsiderationsforthepressureboundarycomponentsofboththepart-lengthCRDMs,whicharenotinuse,butthepressureboundarycomponentsremainpresent,andthefull-lengthCRDMs.Theunit2CRDMsweredesignedandfabricatedtotherequirementsofthe1968EditionoftheASMECode,SectionIZI.Theanalysiswasbasedonthecriteriacontainedinthe1971editionoftheASMECode,SectionZII.InlatereditionsofSectionZZI(NCA-1140),itisanacceptedpracticetousealaterASMEcodeeditionforanalysisofcomponents.TheCRDMevaluationaddressedtherevisedNSSSparametersandNSSSdesigntransientsassociatedwiththeupratingandcomparedtheseparametersandtransientstotheconditionsassumedintheoriginaldesignanalysisfortheCRDMs.Thedifferenceswereidentifiedandusedtoobtainstressandfatigueresultsforpoweruprate.Intheoriginalanalyses,thecomponentofthepressurehousingthatexperiencesthegreateststressrangeandhasthehighestfatigueusageistheuppercanopy.TheDTsandDPsduetoupratingwereidentifiedandusedtoestablishstresslevelsusingtheratiomethodbasedontheoriginalanalysis.Thethermalandpressurestressesoftheoriginalanalysiswereseparatedsothattheincrementalchangesfromeitherpressureortemperaturecouldbedetermined.Theresultsoftheevaluationare:Themaximumstressintensityrangeis109,960psi,whichislessthanthemaximumallowablerangeofthermalstressof127,105psideterminedusingthethermalratchettingrequirementsoftheASMECode,SectionIII,NB-3228.2.Thetotalfatigueusagefactoris0.672,whichislessthantheusagefactorcalculatedintheoriginalconservativeanalysis(0.858)andislessthantheallowablelimitof1.0(ASMECode,SectionIII,1971Edition).Inconclusion,basedonthenumericalevaluationofthestressatthelocationoftheCRDMhavingthegreatestfatigueusage,theCRDMpressurehousingmeetstherequirementsoftheASMEcodeatpoweruprateconditions.NRCUESTZONNO.5"InTable2.1-1ofReference1,thecurrentcorepowerlimitis3391MWtthermal.Onpage2ofAppendix1toReference1,thegrouponeproposedchangeshavethecurrentratedcorepowerlevelof3411MWt.Clarifythedifference."RESPONSETOUESTZONNO.5Table2.1-1ispartofWCAP-14489thatisattachment6toourAEP:NRC:1223submittal.WCAP-14489waspreparedbyourcontractor,WestinghouseElectricCorporation.TheentryindicatestheoriginallicensedcorepowerofCookNuclearPlantunit2was3391MWt.Thisiscorrect.However,CookNuclearPlant'sunit2was AttachmenttoAEP:NRC:1223EPage12upratedfromaratedthermalpowerof3391MWttoaratedthermalpowerof3411MWtforcycle4byAmendmentNo.48toLicenseNo.DPR-74.Thiseffortwassupportedbyourcontractor,ExxonNuclearCompany,Incorporated.SinceWestinghousedidnotplayamajorroleintheuprateto3411MWt,theauthorsofWCAP-14489decidedtoreferenceonlytheoriginalratedthermalpowerinWCAP-14489.NRCUESTIONNO.6"Discusstheanalyticalmethodology'ndassumptionsusedinevaluatingpipesupports,nozzles,penetration,guides,valves,pumps,heatexchangers,andsupportanchorsattheuprateconditions.Weretheanalyticalcomputercodesusedintheevaluationdifferentfromthoseusedintheoriginaldesignbasisanalysis2Ifso,identifythenewcodesandprovidejustificationforusingthenewcodesandstatehowthecodeswerequalifiedforsuchapplications."RESPONSETOUESTIONNO.6Theupratingprogramwillhaveaninsignificantimpactonpipesupports,guides,andanchors.Thatis,theresultantprimaryandsecondarysidetemperaturesareonlyslightlyhigherthantheoriginaldesignbasistemperatures.Thissmalltemperaturerisewillresultinminimalincreasesintheforcesthatthesupports,guides,andanchorswillexperience.Theseincreasesarewellwithinthesubstantialdesignmarginsforthecomponents.Thus,theslightincreaseintemperaturewillnotresultinadeviationfromtheoriginaldesignbasesofthesupports,guides,andanchors.Nonewcomputercodeswereusedforthisreview.Asdetailedinourresponsetoquestionno.3,thesafetysystemsreviewedforimpactfromtheuprateconditionsweretheAFW,CCW,andESWsystems.ThisreviewindicatedthatthepumpsandvalvesarenotsignificantlyaffectedbytheupratedpowerconditionsbecausetheoriginaldesignbasisperformanceandT/Srequirementsremainunchanged.TheESWandCCWsystemswereanalyzed,utilizingtheProto-FlocomputercodeinordertodeterminethesysteminputsusedbyWestinghouse.Theuseofthesysteminputswasdetailedinour'EP:NRC:1223Csubmittal,datedJune10,1997.DetailsoftheProto-FlocomputercodewerediscussedinourAEP:NRC:1238F1submittal,datedApril10,1997,whichwasourreplytoarequestforadditionalinformationoncalculationsprovidedtotheNRCduringaSOPIinspection.TheWestinghousesystemsevaluatedarethe:1)reactorcoolantsystem(RCS);2)chemicalandvolumecontrolsystem(CVCS);3)emergencycorecoolingsystem(ECCS);and4)residualheatremovalsystem(RHRS).Thefluidsystemscomputercodesusedinthisevaluationwerethe:RHRCOOLCodeusedtoevaluatetheRHRScooldowncapabilities,andTSHXBheatexchangercodeusedtoevaluatetheheatexchangerperformance.Theanalyticalmethodologyinthecomputercodesisnotdifferentthantheoriginaldesignbasiscode.Thesecomputercodesarein AttachmenttoAEP:NRC:1223EPage13theWestinghousequalityprogramdescribedintheenergysystemsbusinessunitpolicyandprocedures.SentFuelPoolDecaHeatAnalsisMethodAllspentfuelpooldecayheatcalculationswereperformedusingimplementationsoftheORIGEN2computercodedevelopedatOakRidgeNationalLaboratory.Thisprogramhasalonghistoryofuseinthecommercialnuclearpowerindustryforbothisotopeproductionandthermalpowercalculations.TheORIGEN2codeisarigorousisotopegenerationanddepletioncodethataccuratelypredictstheproductsandby-productsoffissionandtheresultingheatgenerationrates.Thedecayheatgenerationrateinthepoolconsistsoftwocomponents:thedecayheatgeneratedbypreviouslydischargedfuelassemblies,andthedecayheatgeneratedbyfreshly(recently)dischargedassemblies.Thedecayheatcontributionofpreviouslydischargedfuelassemblieschangesverylittleovershortperiodsoftime,andis,therefore,heldconstantintheanalyses.Becauseofthenatureofexponentialdecay,thissimplificationisconservative.TheHoltecQAValidatedLONGORcomputerprogram,whichincorporatestheORIGEN2code,wasusedtocalculatethisdecayheatcomponent.Thedecayheatcontributionofthefreshlydischargedfuelassemblieschangessubstantiallyoverevenveryshortperiodsoftime.Thisdecayheatcontributionisthereforeevaluatedastime-varying.TheHoltecQAValidatedBULKTEMcomputerprogram,thatincorporatestheORIGEN2code,wasusedtocalculatethisdecayheatcomponent.BulkSentFuelPitSFPTemeratureAnalsisMethodDuetothetime-varyingdecayheatcomponent,thetotaldecayheatisalsotime-varying.ThebulkSFPtemperatureisthereforecalculatedasafunctionoftime..Thefollowingenergybalanceissolvedtoobtainthetemperatureateachinstantintime:where:CistheSFPthermalcapacity,Btu/oFTisthebulkSFPtemperature,~F7isthetimeafterreactorshutdown,hrQ~~(r)isthedecayheatgeneration,Btu/hrQ~(T)istheSFPCSheatrejection,Btu/hrQ~>>(T)istheevaporativeheatloss,Btu/hrTheevaporativeheatlosstermincludesbothevaporativeandsensibleheattransferfromthesurfaceoftheSFP.Theimplementationofthistermhasbeenbenchmarkedagainstactualin-planttestdata.Thesolutionofthisfirst-orderordinarydifferentialequationisperformedusingtheBULKTEMprogram.Time-to-BoilAnalsisMethodFollowingalossofforcedcooling,thecontinuingdecayheatloadintheSFPwillcausethebulkSFPtemperaturetorise.Theequationenergy=balancethatdefinesthistransientphenomenais AttachmenttoAEP:NRC:1223EPage14similartotheordinarydifferentialequationpresentedabove,butdoesnotincludetheQ~termanddoesincludeatime-varyingSFPthermalcapacity,toaccountfortheevaporativewaterlosses.ThetimeavailableforcorrectiveactionbeforebulkSFPboilingoccursisdeterminedusingtheHoltecQAvalidatedTBOILcomputerprogram.Thedecayheatgenerationandevaporativeheatlosstermsinthisformulationareidenticaltothosedefinedabove,exceptforthefollowingtwodifferences:Thedecayheat.iscalculatedzsingthecorrelationsofUSNRCBranchTechnicalPositionASB9-2insteadofORIGEN2.NoincrementalcreditisgivenforevaporativeheatlossatSFPbulktemperaturesgreaterthan170'.LocalTemeraturesAnalsisMethodThedecayheatgeneratedbythefuelassembliesstoredintheSFPinducedabuoyancydrivenflowfieldupwardthroughthefuelrackcells.Coolerwaterissuppliedtothebottomoftherackscellsthroughtherack-to-wallgapsandrack-to-floorplenum.TheHoltecQAValidatedTHERPOOLcomputerprogramwasusedtoperformthisanalysis.NRCUESTIONNO.7"DiscusstheeffectofflowinducedvibrationonthesteamgeneratorU-bendtubesandtheheatexchangerinconsiderationofhighflowraterequiredforthepoweruprate."RESPONSETOUESTIONNO.7ThesteamgeneratorsevaluatedforCookNuclearPlant'sunit2upratingprogramarethereplacementmodel51Fseries.AcompleteU-bendfatigueevaluationwasnotnecessarybecauseoftheadvanceddesignfeaturesincorporatedintothereplacementsteamgenerators.OneoftheprerequisitesforexcessiveU-bendtubefatigueisdentinginthetoptubesupportplate.Thequatrefoilstainlesssteeldesignisexpectedtoinhibitfuturedenting.Inaddition,theanti-vibrationbars(AVBs)incorporatedintothereplacementsteamgeneratorswereinsertedtoauniformdepththreerowsdeeperthanconventionalsteamgenerators.Uniforminsertioninhibitslocalflowpeaking,anddeeperinsertionaddsmargintocalculatedtubestabilityratiosforthelargestradiustubenotsupportedbyAVBs.Boththesefactorsreducetheriskoffluidelastictubevibration,whichcouldleadtoexcessiveU-bendtubefatigue.FlowinducedtubevibrationandwearanalysisforCookNuclearPlant'sunit2model51Freplacementsteamgeneratorsreferencesnormaldesignloadsforoperationat852.75MWtpersteamgeneratorplusconsiderationofarangeofoperatingconditionsforwhichoperationisapprovedat900MWtpersteamgenerator.Themainimpactoftherangeofoperatingconditionswastherangeofoperatingpressuresconsidered,soexplicitcalculationsprimarilyaddresspressureloadingeffectsthataddtothe852.75MWtbase.Calculatedresultsfortheadvancedmodel51Fdesignyieldlargemarginsrelativetofluidelasticinstabilitylimits:themaximumstabilityratiois0.36versusalimitof1.00.Upratingfrom852.75to900MWtwouldincreasethelimitingstabilityratiotoonly0.38;aresultthatisstillmorethan2.5timesbelowthe AttachmenttoAEP:NRC:1223EPage15limit.Correspondingdisplacementsduetoturbulenceintheflowarewellbelow0.001inch.Basedontheseconsiderations,thereplacementsteamgeneratorsatCook'uclearPlant'sunit2areconsideredtobeeffectivelydesignedforthehighflowratesrequiredforthepower,uprateto3600MWt.