ML17333B036

Forwards Response to 970709 RAI Re 960711 5% Thermal Power Uprate AEP:NRC:1223 Submittal

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'1g INFORMATION DISTRIBUTZO1lgTEM (RIDE)ACCESSION'3NBR:9709170108 DOC.DATE:

97/09/09NOTARIZED:

YESDOCKETFACIL:50'-.316 DonaldC.CookNuclear PowerPlant,Unit2,IndianaM05000316AUTH.NAMEAUTHORAFFILIATION FITZPATRICK,E.

IndianaMichiganPowerCo.(formerly Indiana6MichiganEleRECIP.NAME RECIPIENT AFFILIATION DocumentControlBranch(Document ControlDesk)

SUBJECT:

Forwardsresponseto970709RAIre9607115%thermalpoweruprateAEP:NRC:1223 submittal.

DISTRIBUTION CODE:A001DCOPIESRECEIVED:LTR ENCLSIZE:TITLE:ORSubmittal:

GeneralDistribution NOTESRECIPIENT IDCODE/NAME PD3-3LAHICKMAN,J INTERNILECE1NRRDE/EMCBNRR/DSSA/SPLB NUDOCS-ABSTRACT EXTERNAL:

NOACCOPIESLTTRENCL11111111111111RECIPIENT IDCODE/NAME PD3-3PDNRR/DE/ECGB/A NRR/DRCH/HICB NRR/DSSA/SRXB OGC/HDS2NRCPDRCOPIESLTTRENCL111111111011E0DUNOTETOALL"RIDS"RECIPIENTS:

PLEASEHELPUSTOREDUCEWASTE.TOHAVEYOURNAMEORORGANIZATION REMOVEDFROMDISTRIBUTION LISTSORREDUCETHENUMBEROFCOPIESRECEIVEDBYYOUORYOURORGANIZATION, CONTACTTHEDOCUMENTCONTROLDESK(DCD)ONEXTENSION 41S-2083TOTALNUMBEROFCOPIESREQUIRED:

LTTR13ENCL12 i~

indianaMichiganPowerCompany~500CircleDriveBuchanan, Ml491071395 5~iIJrtINtIANSl SIICNIGAN PQWMSeptember 9,1997AEP:NRC:1223E DocketNo.:50-316U.S.NuclearRegulatory Commission ATTN:DocumentControlDeskWashington, D.C.20555Gentlemen:

DonaldC.CookNuclearPlantUnit2RESPONSETOREQUESTFORADDITIONAL INFORMATION REGARDING POWERUPRATEANDRELATEDCHANGESThisletteranditsattachment constitute aresponsetotheJuly9,1997,NRCrequestforadditional information regarding ourJuly11,1996,5%thermalpoweruprateAEP:NRC:1223 submittal.

Therequestforadditional information primarily involvesanalysisassumptions andmethodology.

Thisletterissubmitted pursuantto10CFR50.30(b)and,assuch,includesanoathstatement.

Sincerely, E.E.Fitzpatrick VicePresident SWORNTOANDSUBSCRIBED BEFOREMEmyrrhTHIS7DAYOFo~&~gP1997NotaryPublic~/-/-4/vlbAttachment UNDALBOEI.CKENotaryPublic,BerrienCounty,MlMyCommission ExpiresJanuary21,2001A.A.BlindA.B.BeachMDEQ-DW&RPDNRCResidentInspector J.R.Padgettrtr->a('.,~sAtsr%s'st70'sti70i08

'st70'st0'st PDRADOCK050003i6PPDRllllllllllllllllllllllllllllllllllllllll ttCFII.IIC~

ATTACHMENT TOAEP:NRC:1223EDonaldC.CookNuclearPlantUnit2RESPONSETOREQUESTFORADDITIONAL INFORMATION REGARDING POWERUPRATEANDRELATEDCHANGES Attachment toAEP:NRC:1223E Page1NRCUESTIONNO.1"InSection2.0ofReference 2,youindicated thatWCAP-11902 andSupplement wereusedasthebasisfortheevaluation oftheUnit2operation atcorepowerlevelof3588MWt.However,WCAP-11902 licensing reportwasreviewedandapprovedbythestaff,forD.C.CookUnit1operating at3250MWt.ClarifywhethertheSupplement toWCAP-11902,

entitled, "ReratedPowerandRevisedTemperature andPressureOperation forCookNuclearPlantUnits1and2Licensing Report,"wasreviewedandapprovedbythestaffforapplication attheCookNuclearPlant(CNP).Ifnot,statethebasisofapplyingthesetwopreviousevaluations forallperformance parameters betweentheproposedUnit2uprateandthepreviousreratedprogram."

RESPONSETOUESTIONNO.1Attachment 5toAEP:NRC:1223 submittal, fromE.E.Fitzpatrick totheUSNRCdocumentcontroldesk,datedJuly11,1996,is"Discussion ofPreviousRelatedSubmissions."

Theintroduction sectionofattachment 5addresses, inageneralway,thefactthattheanalysesthatsupporttheproposedupratinghavebeenperformed overaperiodofyearsasapartofothereffortswithmoreimmediate shortrangegoals.Thisattachment states:"TheanalysesthatsupporttheproposedupratingofDonaldC.CookNuclearPlantUnit2havebeenperformed overaperiodofyearsinseveralcontexts.

Theanalysisofthenuclearsteamsupplysystem(NSSS)foranNSSSpowerof3600MWtwasperformed inconjunction withanalysestooperateunit1atreducedtemperature andpressure(the"Rerating Program").Mostofthecoreresponseanalyseswereperformed atanupratedcorethermalpowerof3588MWtasapartofthetransition fromAdvancedNuclearFueltoWestinghouse Vantage5fuel.Therecentlysubmitted

analyses, AEP:NRC:1207 (erroneously statedtobeAEP:NRC:1223 inthesubmittal),

tosupportanincreaseinthepermitted levelofsteamgenerator tubepluggingforunit1includesasteammassandenergyreleaseanalysistothecontainment whichboundsbothunitsat3600MWt.Forthissubmittal (i.e.,AEP:NRC:1223),

previousNSSSanalysesandcoreresponseanalyseshavebeenreviewed, newanalyseshavebeenperformed wherenecessary, andthebalanceofplantevaluated, asdescribed withinthissubmittal, tosupporttheproposaltoincreasethecoreratedthermalpowerto3588MWt."Inparticular, asindicated inattachment 5,thesupplement toWCAP11902wassubmitted inpartinsupportofanumberofproposedtechnical specification (T/S)changes.Itwassubmitted initsentiretyinsupportofourproposaltoreducetheboronconcentration intheboroninjection tanksofbothunitsto0ppm.Oursubmittal wasletterAEP:NRC:1140, "Technical Specification ChangeRequest,BoronInjection Tank(BIT),BoronConcentration Reduction",

fromM.P.AlexichtoT.E.Murley,datedMarch26,1991.TheproposalwasapprovedbyAmendment No.158toFacilityOperating LicenseNo.DPR-58'nd Amendment No.142toFacilityOperating licenseNo.DPR-74.

Attachment toAEP:NRC:1223E Page2NRCUESTZONNO.2"Clarifywhetherthereratinganalysesofthepressuretransients andthepostulated loss-of-coolant accident(LOCA)includetheproposedpressurizer safetyandreliefvalvetolerance

+/-3%,andthepreviously NRC-approved mainsteamsafetyandreliefvalvestolerance of+/-3%.Zfnot,statehowthereratinganalysesappliestotheproposedUnit2poweruprate."RESPONSETOUESTZONNO.2Theanalysesperformed forsubmittal AEP:NRC:1223, toincreasethethermalpowerofCookNuclearPlantunit2to3588MWt,assumedsetpointtolerances of3%forboththepressurizer safetyvalvesandthesteamgenerator safetyvalves.Thepressurizer safetyvalvesetpointtolerance isspecifically addressed fortheapplicable analysesinsection3.3,"Non-LOCA Analyses",

ofWCAP-14489,attachment 6tosubmittal AEP:NRC:1223.

Thisassumption iscalledoutspecifically fortheapplicable eventsbecausethisisanewassumption fortheunit2analyses.

Thepressurizer pressuresetpointdoesnotaffecttheLOCAeventbecausetheprimarysystemdepressurizes.

Theassumption ofa3%tolerance forsteamgenerator safetyvalvesetpoints wasnotspecifically calledoutforthenewanalysesbecauseitisanassumption thatwaspreviously submitted andreviewed.

Anassumption of3%setpointtolerance forsteamgenerator safetyvalvesetpoints isinputtotheapplicable analysesintheunit2upratesubmittal.

NRCUESTZONNO.3"Discusstheoperability ofthesafety-related mechanical components (i.e.,valvesandpumps)affectedbythepowerupratetoensurethattheperformance specifications andtechnical specification requirements (e.g.,flowrate,closeandopentimes)willbemetfortheproposedpoweruprate.Confirmthatthesafety-related motoroperatedvalves(MOVs)willbecapableofperforming theirintendedfunctions following thepoweruprateincluding suchaffectedparameters asfluidflow,temperature, pressureanddifferential

pressure, andambienttemperature conditions.

Zdentifymechanical components forwhichoperability attheupratedpowerlevelcouldnotbeconfirmed."

RESPONSETOUESTZONNO.3AFWCCWANDESWSYSTEMSThesafetysystemswereviewedforimpactfromupratedconditions aretheauxiliary feedwater (AFW),component coolingwater(CCW),andessential servicewater(ESW)systems.Ourreviewindicates thatthemechanical components (i.e.,valvesandpumps)inthesesystemsarenotsignificantly affectedbytheupratedpowerconditions.

Theperformance andT/Srequirements forthesesystemsremainunchanged.

Becausethesystemparameters havenotchanged,theassociated MOVoperability isnotimpacted.

Thefollowing summarizes ourreviewinsupportofthepreceding statement fortheindicated systems.TheAFWsystemprovideswatertothesteamgenerators whenthemainfeedwater,system isunavailable duetoalossoffeedwater, unit Attachment toAEP:NRC:1223E Page3trip,feedwater orsteamlinebreak,lossofoffsitepower,orloss-of-coolant accident(LOCA).TheAFWsystemisdesignedandanalyzedtoprovidesufficient flowtothesteamgenerators duringtheseeventsagainstasteamgenerator pressurecorresponding tothesetpressure, plusaccumulation ofthelowestsetsafetyvalves.TheAFWsystemisalsocapableofproviding reducedflowatthehighersteamgenerator pressures, plusaccumulation corresponding tothehighersetsafetyvalves.Theupratedconditions didnotaltertheAFWsystem'sflowrequirements orthesystem'sabilitytofulfilltheserequirements.

Theupratedconditions didnotaffectorrevisethesafetyvalve'ssetpressure, theAFWpump'soperating parameters (flowandhead),orthefluidparameters (temperature andpressure).

Theupratealsodidnotresultinanysignificant changesinambienttemperatures.

Therefore, theAFW'sMOVrequirements areessentally unchanged, andthemechanical components inthesystemarenotsignificantly affected.

TheCCWsystemisaclosedloopsystemthatservesasanintermediate loopbetweenpotentially radioactive systemsandlakewatertoensurethatleakageofradioactive fluidiscontained withintheplant.TheCCWsystemisdesignedandanalyzedtosupplycoolingwaterflowduringtheinjection andrecirculation phasesofaLOCAandduringunitoperation.

TheLOCAlong-term massandenergyreleaseandcontainment integrity analysesperformed byWestinghouse utilizedCCWsystemflowrates andheatexchanger UAsrepresentative oftheupratedconditions.

TheWestinghouse analysesdetermined theresultswereacceptable forcontainment integrity pressureandtemperature response.

Thesedetailswereprovidedinoursubmittal AEP:NRC:1223C, datedJune10,1997.Basedonthis,theupratedconditions didnotsignificantly impacttheCCWsystem'sheatremovalrequirements, orthesystem'scapability tomeettheserequirements.

TheCCWpumps'perating parameters (flowandhead)andfluidparameters (temperature andpressure) werenotchangedasaresultoftheuprate.Theupratealsodidnotresultinanysignificant changesinambienttemperatures.

Therefore, theCCW'sMOVrequirements areessentially unchanged andthemechanical components inthesystemarenotsignificantly affected.

TheESWsystemprovidescoolingwaterrequirements totheCCWheatexchangers, emergency dieselgenerators, CTSheatexchangers, andthecontrolroomairconditioning condensers.

TheESWsystemisoperatedinconjunction withtheCCWandCTSsystems.TheESWpump'soperating parameters (flowandhead)andfluidparameters (temperature andpressure) werenotchangedasaresultoftheuprate.Theupratealsodidnotresultinanysignificant changesinambienttemperatures.

Therefore, theESW'sMOVrequirements remainessentially unchanged andthemechanical components inthesystemarenotsignificantly affected.

RCSCVCSANDRHRSSYSTEMSThesafetysystemstobereviewedforimpactfromupratedconditions arethereactorcoolantsystem(RCS),emergency corecoolingsystem(ECCS),andchemicalvolumecontrolsystem(CVCS).Ourreviewindicates thatthemechanical components inthesesystemsarenotsignificantly affectedbytheupratedpowerconditions.

Theperformance andT/Srequirements forthesesystemsremainunchanged.

Becausethesystemparameters havenotchanged,theassociated MOVsoperation isnotsignificantly impacted.

Attachment toAEP:NRC:1223E Page4TheRCSconsistsoffouridentical heattransferloopsconnected inparalleltothereactorvessel.Eachloopcontainsareactorcoolantpump(RCP)andasteamgenerator.

Inaddition, thesystemincludesapressurizer, apressurizer relieftank,inter-connecting piping,andinstrumentation necessary foroperational control.Duringoperation, theRCPscirculate pressurized waterthroughthereactorvesselandthefourcoolantloops.Thewater,thatservesbothasacoolant,moderator, andsolventforboricacid(chemical shimcontrol),

isheatedasitpassesthroughthecore.Itthenflowstothesteamgenerators wheretheheatistransferred tothesteamsystem,andreturnstotheRCPstorepeatthecycle.TheRCSpressureiscontrolled bytheuseofthepressurizer wherewaterandsteamaremaintained inequilibrium byelectrical heatersandwatersprays.Threespringloadedsafetyvalvesandthreepoweroperatedreliefvalvesareconnected tothepressurizer anddischarge tothepressurizer relieftank,wherethesteamiscondensed andcooledbymixingwithwater.Fluidsystemscalculations wereperformed, evaluating thecapability oftheRCStooperateattheuprateprogramconditions.

Theupratedpowerconditions didnotaffectanyoftheRCSsafetyrelatedmechanical components designbasis.TheMOVsfluidsystemdesignconditions (fluidflow,temperature, pressureanddifferential pressure) werenotsignificantly affectedbytheupratedconditions.

TheCVCSprovidesforboricacidaddition, chemicaladditions forcorrosion control,reactorcoolantclean-upanddegasification, reactorcoolantmake-up,reprocessing ofwaterletdownfromtheRCS,andRCPsealwaterinjection.

Duringplantoperation, reactorcoolantflowsthroughtheshellsideoftheregenerative heatexchanger, thenthroughaletdownorifice.Theregenerative heatexchanger reducesthetemperature ofthereactorcoolant,andtheletdownorificereducesthepressure.

Thecooled,lowpressurewaterleavesthereactorcontainment andenterstheauxiliary building.

Asecondtemperature reduction occursinthetubesideoftheletdownheatexchanger, followedbyasecondpressurereduction duetothelowpressureletdownvalve.Afterpassingthroughoneofthemixedbeddemineralizers, whereionicimpurities are,.removed, coolantflowsthroughthereactorcoolantfilterandentersthevolumecontroltank(VCT).Theregenerative andletdownheatexchangers aredesignedtocoolletdownflowfromT,~to115'.Thevariations inT,~considered fortheuprateprogramareboundedbythedesigninlettemperature of547'fortheregenerative heatexchanger.

Therefore, thecoolingrequirements oftheletdownfunctionaremetwiththerevisedoperating parameters.

TheletdownfunctionisdesignedtoreducethestaticpressureofthereactorletdownstreamfromtheRCPsuctionpressuretoVCToperating

pressure, suchthatthedesignpressureofintervening pipingandcomponents isnotexceeded, andfluidismaintained inasubcooled condition throughout thesystem.Thepressurereduction xequirements oftheletdownfunctionaremetwiththerevisedoperating parameters.

Thecentrifugal chargingpumpoperating conditions havenotbeenimpactedbytheupratingconditions.

Fluidsystemscalculations wereperformed evaluating thecapability oftheCVCStooperateattheuprateprogramconditions.

Theupratedpowerconditions donot Attachment toAEP:NRC:1223E Page5significantly affecttheCVCSsafetyrelatedmechanical components'esign bases.TheECCSinjectsboratedwaterintothereactorfollowing abreakineitherthereactororsteamsystemsinordertocoolthecoreandpreventanuncontrolled returntocriticality.

Twosafetyinjection(SI)pumpsandtworesidualheatremovalpumpstakesuctionfromtherefueling waterstoragetank(RWST)anddeliverboratedwatertofourcoldlegconnections viatheaccumulator discharge lines'.Inaddition, twocentrifugal chargingpumpstakesuctionfromtheRWSTonSIactuation andprovideflowtotheRCSviaseparateSIconnections oneachcoldleg.Atthecompletion oftheinjection phasefromtheRWSTtheECCSisthenalignedtothecontainment sump,asthesuctionsource,toprovidethecoldorhotlegrecirculation injection flows.Theprimarysystempressures considered forthisprogramarelessthan,orequalto,theprimarysystempressureagainstwhichtheoriginalsystemwasdesignedtodeliver.Therefore, therevisedprimarysystemparameters donotrequireanincreaseineitherthemotivepressureorcorecoolingcapacityoftheECCS.Fluidsystemscalculations wereperformed evaluating thecapability oftheECCStooperateattheuprateprogramconditions.

Theupratedpowerconditions didnotsignificantly affecttheECCSsafetyrelatedmechanical components'esign bases.NRCUESTIONNO.4"Inreference toSections3.11.2and3.11.3ofreference 2(WCAP-14489),providethemaximumcalculated stressesandcumulative UsageFactorsatthemostlimitinglocations andcomponents ofthereactorvesselandinternals, steamgenerator, reactorcoolantpump,pressurizer, andcontrolroddrivemechanism.

Alsoprovidetheallowable codelimits,thecode,andthecodeeditionusedintheevaluation forthepoweruprate.Ifdifferent fromthecodeofrecord,providethenecessary justification."

RESPONSETOUESTIONNO.4ReactorVessel:Withrespecttosection3.11.2,theresultsofthereactorvesselanalysesandevaluations aresummarized below.Thestressintensity andfatigueusagelimits(withtheexception ofthe3Smaximumrangeofprimaryplussecondary stressintensity limitforthecontrolroddrivemechanism (CRDM)housingsandoutletnozzlesafeend)oftheASMEBoilerandPressureVesselCode,SectionI1I,1968Edition,withAddendathroughtheSummerof1968,aremet.Theexceeding ofthe3SlimitfortheCRDMhousingsandoutletnozzlesafeendisreconciled byusingtheASMEcodeacceptable methodofelastic-plastic analysesinaccordance withASMEBoilerandPressureVesselCode,SectionIII,1971Edition.CRDMHousinThemaximumrangeofprimaryplussecondary stressintensity iscalculated tobe77.76ksi,whichexceedsthe3Slimitof69.9ksi.However,asimplified elastic-plastic analysiswasperformed inaccordance withparagraph NB-3228.3 oftheASMEBoilerandPressureVesselCode,SectionIII,1971Edition,andthehigherrangeofstressintensity isreconciled.

Themaximumcumulative Attachment toAEP:NRC:1223E Page6fatigueusagefactoris0.1687,whichisbelowtheASMEcodelimitof1.0.MainClosureReionThemainclosureregionofthereactorvesselconsistsofthevesselflange,theclosureheadflange,andtheclosurestudassemblies thatcoupletheheadtothevessel.Themaximumrangesofstressintensity intheclosureheadflangeandthevesselflangeare65.26ksiand61.04ksi,respectively, comparedtotheASMEcode3Slimitof80.1ksi.Themaximumserviceintheclosurestudsis91.8ksi,whichcomparesfavorably tothe3Slimitof107.7ksi.Themaximumcumulative fatigueusagefactorfortheclosureheadflange,vesselflangeandclosurestudsare0.018,0.029and0.99,respectively.

Theusagefactorsarealllessthanthe1.0ASMEcodelimit.However,itshouldbenotedthattheclosurestudusagefactorof0.99wascalculated undertheassumption thatthefirst25%ofthe11,680occurrences ofplantloadingandunloading, at5%offullpowerperminute(2,920occurrences ofeach),occurredduringthefirsttenyearsofoperation whenthevesseloutlettemperature (T)was599.3'.OutletNozzleThemaximumrangeofprimaryplussecondary stressintensity intheoutletnozzleendiscalculated tobe59.58ksicomparedtothe3Slimitforaustenitic stainless steelmaterialof50.1k-i.Becausethemaximumrangeofstressintensity exceeds3S,asimplified elastic-plastic analysisperparagraph NB-3228.3 oftheASMEBoilerandPressureVesselCode,SectionIII,1971Edition,wasperformed thatjustified thehighermaximumrangeofstressintensity.

Themaximumusagefactoratthesafeendis0.021,whichislessthan1.0.Themaximumrangeofstressintensity intheoutletnozzleandnozzletoshelljunctureis57.09ksi,comparedtothe3Sallowable 80.1ksi.Themaximumcumulative usagefactorinthenozzleandnozzletoshelljunctureis0.0631,whichisalsolessthan1.0.InletNozzleThemaximumrangeofstressintensity intheinletnozzlesafeendis49.65ksi,whichislessthan3S=50.1ksi.Themaximumrangeofstressintensity intheinletnozzleandnozzletoshelljunctureis49.86ksi,whichcomparesfavorably witha3Slimitof80.1ksi.Themaximumcumulative usagefactorsinthenozzlesafeendandnozzletoshelljunctureare0.0174and0.0977,respectively, whicharebothlessthan1.0.VesselWallTransition Themaximumrangeofstressintensity andcumulative fatigueusagefactorforthevesselwalltransition, betweenthenozzleshellandthevesselbeltline, are33.57ksiand0.0066.ThesevaluesarelessthantheASMEcodelimitsof80.1ksiand1.0,respectively.

Attachment toAEP:NRC:1223E Page7BottomHead-to-Shell JunctureThemaximumrangeofprimaryplussecondary stressintensity atthejuncture, betweenthevesselbottomhemispherical headandthevesselbeltlineshell,is34.53ksicomparedtoa3Sallowable of80.1ksi.Themaximumcumulative fatigueusagefactoratthejuncturewascalculated tobe0.0182,whichislessthan1.0.BottomHeadInstrumentation Penetrations Thebottomheadinstrumentation penetrations areacceptable foruprating, baseduponamaximumrangeofprimaryplussecondary stressintensity of51.49ksi,andamaximumcumulative usagefactorof0.1220.Thesevaluescomparefavorably withtheASMEcodeallowables of69.9ksi(3S)and1.0,respectively.

CoreSuortPadsThecoresupportpadswereevaluated tohaveamaximumrangeofstressintensity of69.7ksi,comparedtoa3Slimitof69.9ksi.Themaximumcumulative fatigueusagefactorwascalculated tobe0.693,whichislessthanthe1.0ASMEcodelimit.ReactorVesselInternals CookNuclearPlantunit2reactorinternals arecomposedoftwosections, theupperinternals andthelowerinternals.

Evaluations wereperformed forthecriticalcomponents forboththeupperinternals andlowerinternals.

Thefollowing isalistofthecriticalcomponents fortheupperandlowerinternals.

UerInternals Perforated sectionofthetophatsupportstructure.

LowerInternals LowerSupportAssemblyCoreBarrelandFlangeLowerRadialSupportClevisInsertsBaffle-Former AssemblyUpperCorePlateAlignment PinsThermalShieldThestructural evaluations performed fortheaboveareasconfirmed thattheirstructural integrity andincreased fatigueusagewasfoundtobewithinacceptable limits,according totheoriginaldesignbasis.SteamGenerator:

Theunit2steamgenerators werereplacedin1987.Thediscussion belowaddresses thereplacedcomponents andremaining originaluppershellcomponents separately.

~Attachment toAEP:NRC:1223E Page8RelacementComonentsThecriteriausedtodetermine acceptable stressstatesareprovidedintheASNEBoilerandPressureVesselCode,SectionIII,1968Edition,andtheassociated AddendathroughWinter1968.Component MaximumStressCalcu-latedMaximumStressAllow-ableFatigueUsageCalcu-latedFatigueUsageAllow-ablePrimaryChamber,Tube-sheet,StubBarrelPrimaryNozzles31.9ksi58.2ksi0.130.871.01.0PrimaryManways41.0ksi48.3ksi0.911.0Tubes47.96ksi79.80ksi0.591.0PrimaryChamberDividerPlate0.191.0TubetoTubesheet WeldLowerShell/Cone/Upper ShellTrunnions 79.2ksi58.8ksi80.1ksi80.1ksi0.750.120.011.01.01.0MinorBoltedOpenings93.9ksi94.3ksi0.741.0MinorNozzlesInternals Feedwater RingandJ-Nozzles 29.3ksi(2)26.7ksi80.1ksi(2)27.0ksi0.880.060.561.01.01.0(1)Theprimary+secondary stressesexceedtheallowable stresslimitof3S.Aplasticanalysiswasperformed perparagraph N-417.6(b) oftheASMEBoilerandPressureVesselCode,SectionIII,"NuclearVessels",

1968EditionwithAddendatoandincluding Winter1968,codeofrecord,todemonstrate structural integrity.

(2)Themaximumstressesinthesteamgenerator internals occurduringthefaultedconditions.

Forthenormalandupsetconditions, theprimary+secondary

+peakstressesinthesteamgenerator internals arelow,andbelowtheendurance limit.Therefore, themaximumfatigueusageforthesteamgenerator internals is0.06.

Attachment toAEP:NRC:1223E Page9OriinalUerShellComonentsPrimarystressesandmaximumstressrangesarenotaffectedbytheupratingconditions, andthesecalculations werenotrepeated.

Whenconsidering theupperandlowerboundprimarytemperatures, theupperboundtemperature conditions areveryclosetothetransient conditions usedinthereference

analyses, andtheresulting fatigueusagesshowonlyslightvariations fromthereference conditions.

However,thelowerboundtemperature conditions canresultinincreased fatigueusagesinsomecases.Asummaryofthefatigueusagesisprovidedbelow.Component Referenced FatigueUsageUpperBoundTemperature FatigueUsageLowerBoundTemperature MainFeedwater NozzleSecondary ManwayShellPenetration 0.530.170.7240.0510.9410.053Secondary ManwayBolts(3)0.4270.825SteamNozzle0.590.6160.616(3)Thereference valueforfatigueisnotprovided.

Thestressesusedfortheanalysisoftheboltsaretakenfromanothermodelsteamgenerator, withscalefactorstoaccountforgeometryvariations.

Aspartoftheupratingprogram,thesteamgenerator structural integrity wasevaluated toaccountfortherevisedlossofloadandlossofoffsitepowertransients.

Theevaluation showedthatthecomponent mostaffectedbytheupratingprogramisthetubesheet-to-channel headjunction.

Thestressintensities continuetosatisfythestresslimits.Thecalculated valueofthefatigueusage,0.34,remainswithinthemaximumallowable limitof1.0.ReactorCoolantPumTheevaluation performed fortheRCPsaddressed theASMEcodestructural considerations fortheRCPcasing,mainflange,mainflangebolts,thermalbarrier,casingfoot,casingdischarge, andsuctionnozzles,casingweirplate,sealhousing,andauxiliary nozzles.Forunit2theASMECode,SectionZZI,1968Edition,withAddendathroughSummer1969,wasusedasaguide.TheRCPevaluation addressed therevisedNSSSparameters andNSSSdesigntransients associated withtheuprating, andcomparedtheseparameters andtransients totheconditions assumedintheoriginaldesignanalysesfortheRCPs.Thedifferences (i.e.,deltatemperatures

[DTs]anddifferential pressures

[DPs])wereidentified andusedtoobtainstressand,fatigue resultsforpoweruprate.TheDPsassociated withthepowerupratedesigntransients werereviewedtodetermine iftherewereanychangesthatwouldqualify Attachment toAEP:NRC:1223E Page10asa"significant fluctuation" inaccordance withtheASMEcodedefinition, and,thus,requireconsideration relativetofatigue.Itwasconcluded duetothepowerupratedesigntransients, thatallDPswerelessthantheASMEcodedefinition of"significant fluctuation" value,andthatnoratigueconsideration isrequiredbecausethefatiguewaiverremainsunchanged.

Thedesigntransients werethenreviewedtoidentifythemaximumpressuretowhichtheRCPcouldbeexposed.Forunit2,thismaximumpressurewasdetermined tobe2724.1psiaforthelossofloadtransient.

AreviewofRCPanalysesperformed forotherplantsshowedthatincreases to2725psiahavebeenanalyzed.

indetailandshowntobeacceptable.

Itwasconcluded thatthepressuretransients areacceptable.

TheeffectofpoweruprateonthevariousoriginalanalysesfortheRCPswasalsoassessedusingtheNSSSdesigntransients andtheassociated DTvalues.Forthemostpart,thecomparison ofNSSSdesigntransients andassessments ofassociated DTvaluesweresufficient toshowcontinued applicability oftheoriginalanalysestopoweruprateconditions.

OneareawheretheincreaseinDTwassufficient tomeritanalysiswasforthecasingweirplate.Theevaluation showedarangeofstressintensities

=41,379psiforpoweruprateconditions.

Comparison othisvaluetotheASMEcodeprimaryplussecondary stresslimitof3S=50,700psishowedthattheASMEcodelimitissatisfied.

Fatiguerequirements fortheweirplateweresatisfied bythefatiguewaiver(ASMEcode,NB-3222.4(d)).Insummary,theresultsofthepoweruprateassessments showedthattheASMEcodecriteriaaresatisfied atpoweruprateconditions.

Pressurizer:

Theexternalloadsarenotrevisedforthe3600MWtupratingconditions, andthechangesinthepressureloadsdonotaffectthepreviously completed stresscalculations.

Thus,theprimarystressescalculated fortheoriginalanalysisremainvalidattheupratedconditions.

Also,thechangesinthedesigntransients (lossofloadandlossofoffsitepower)didnothaveanysignificant effectontheprimaryplussecondary stresses.

However,forsomecomponents, thefatigueanalysisisaffected.

Thenewcalculated fatigueusagefactorsforeachofthepressurizer components arelistedbelow.Becausethenewcalculated fatigueusagefactorsarelessthan1.0,thepressurecomponents meetthestress/fatigue requirements oftheASMECode,SectionIII,1965Edition,including AddendauptoWinter1966.PRESSURIZER FATIGUEUSAGEFACTORS~ComonentSurgeNozzleSprayNozzleSafetyandReliefNozzleLowerHead,HeaterWellLowerHead,Perforation UpperHeadandShellSupportSkirt/Flange ManwayPadManwayCoverManwayBoltsCalculated FatiueUsae<0.340.991<0.15<0.07<0.020.973<0.020.00.00.0

~Attachment toAEP:NRC:1223E Page11SupportLugInstrument NozzleZmmersion HeaterValveSupportBracket<0.05<0.11<0.010.01ControlRodDriveMechanism:

Theevaluation performed fortheCRDMsaddressed theASMEcodestructural considerations forthepressureboundarycomponents ofboththepart-length CRDMs,whicharenotinuse,butthepressureboundarycomponents remainpresent,andthefull-length CRDMs.Theunit2CRDMsweredesignedandfabricated totherequirements ofthe1968EditionoftheASMECode,SectionIZI.Theanalysiswasbasedonthecriteriacontained inthe1971editionoftheASMECode,SectionZII.InlatereditionsofSectionZZI(NCA-1140),

itisanacceptedpracticetousealaterASMEcodeeditionforanalysisofcomponents.

TheCRDMevaluation addressed therevisedNSSSparameters andNSSSdesigntransients associated withtheupratingandcomparedtheseparameters andtransients totheconditions assumedintheoriginaldesignanalysisfortheCRDMs.Thedifferences wereidentified andusedtoobtainstressandfatigueresultsforpoweruprate.Intheoriginalanalyses, thecomponent ofthepressurehousingthatexperiences thegreateststressrangeandhasthehighestfatigueusageistheuppercanopy.TheDTsandDPsduetoupratingwereidentified andusedtoestablish stresslevelsusingtheratiomethodbasedontheoriginalanalysis.

Thethermalandpressurestressesoftheoriginalanalysiswereseparated sothattheincremental changesfromeitherpressureortemperature couldbedetermined.

Theresultsoftheevaluation are:Themaximumstressintensity rangeis109,960psi,whichislessthanthemaximumallowable rangeofthermalstressof127,105psidetermined usingthethermalratchetting requirements oftheASMECode,SectionIII,NB-3228.2.Thetotalfatigueusagefactoris0.672,whichislessthantheusagefactorcalculated intheoriginalconservative analysis(0.858)andislessthantheallowable limitof1.0(ASMECode,SectionIII,1971Edition).

Inconclusion, basedonthenumerical evaluation ofthestressatthelocationoftheCRDMhavingthegreatestfatigueusage,theCRDMpressurehousingmeetstherequirements oftheASMEcodeatpoweruprateconditions.

NRCUESTZONNO.5"InTable2.1-1ofReference 1,thecurrentcorepowerlimitis3391MWtthermal.Onpage2ofAppendix1toReference 1,thegrouponeproposedchangeshavethecurrentratedcorepowerlevelof3411MWt.Clarifythedifference."

RESPONSETOUESTZONNO.5Table2.1-1ispartofWCAP-14489 thatisattachment 6toourAEP:NRC:1223 submittal.

WCAP-14489 waspreparedbyourcontractor, Westinghouse ElectricCorporation.

Theentryindicates theoriginallicensedcorepowerofCookNuclearPlantunit2was3391MWt.Thisiscorrect.However,CookNuclearPlant'sunit2was Attachment toAEP:NRC:1223E Page12upratedfromaratedthermalpowerof3391MWttoaratedthermalpowerof3411MWtforcycle4byAmendment No.48toLicenseNo.DPR-74.Thiseffortwassupported byourcontractor, ExxonNuclearCompany,Incorporated.

SinceWestinghouse didnotplayamajorroleintheuprateto3411MWt,theauthorsofWCAP-14489 decidedtoreference onlytheoriginalratedthermalpowerinWCAP-14489.

NRCUESTIONNO.6"Discusstheanalytical methodology'nd assumptions usedinevaluating pipesupports, nozzles,penetration, guides,valves,pumps,heatexchangers, andsupportanchorsattheuprateconditions.

Weretheanalytical computercodesusedintheevaluation different fromthoseusedintheoriginaldesignbasisanalysis2 Ifso,identifythenewcodesandprovidejustification forusingthenewcodesandstatehowthecodeswerequalified forsuchapplications."

RESPONSETOUESTIONNO.6Theupratingprogramwillhaveaninsignificant impactonpipesupports, guides,andanchors.Thatis,theresultant primaryandsecondary sidetemperatures areonlyslightlyhigherthantheoriginaldesignbasistemperatures.

Thissmalltemperature risewillresultinminimalincreases intheforcesthatthesupports, guides,andanchorswillexperience.

Theseincreases arewellwithinthesubstantial designmarginsforthecomponents.

Thus,theslightincreaseintemperature willnotresultinadeviation fromtheoriginaldesignbasesofthesupports, guides,andanchors.Nonewcomputercodeswereusedforthisreview.Asdetailedinourresponsetoquestionno.3,thesafetysystemsreviewedforimpactfromtheuprateconditions weretheAFW,CCW,andESWsystems.Thisreviewindicated thatthepumpsandvalvesarenotsignificantly affectedbytheupratedpowerconditions becausetheoriginaldesignbasisperformance andT/Srequirements remainunchanged.

TheESWandCCWsystemswereanalyzed, utilizing theProto-Flo computercodeinordertodetermine thesysteminputsusedbyWestinghouse.

Theuseofthesysteminputswasdetailedinour'EP:NRC:1223C submittal, datedJune10,1997.DetailsoftheProto-Flo computercodewerediscussed inourAEP:NRC:1238F1 submittal, datedApril10,1997,whichwasourreplytoarequestforadditional information oncalculations providedtotheNRCduringaSOPIinspection.

TheWestinghouse systemsevaluated arethe:1)reactorcoolantsystem(RCS);2)chemicalandvolumecontrolsystem(CVCS);3)emergency corecoolingsystem(ECCS);and4)residualheatremovalsystem(RHRS).Thefluidsystemscomputercodesusedinthisevaluation werethe:RHRCOOLCodeusedtoevaluatetheRHRScooldowncapabilities, andTSHXBheatexchanger codeusedtoevaluatetheheatexchanger performance.

Theanalytical methodology inthecomputercodesisnotdifferent thantheoriginaldesignbasiscode.Thesecomputercodesarein Attachment toAEP:NRC:1223E Page13theWestinghouse qualityprogramdescribed intheenergysystemsbusinessunitpolicyandprocedures.

SentFuelPoolDecaHeatAnalsisMethodAllspentfuelpooldecayheatcalculations wereperformed usingimplementations oftheORIGEN2computercodedeveloped atOakRidgeNationalLaboratory.

Thisprogramhasalonghistoryofuseinthecommercial nuclearpowerindustryforbothisotopeproduction andthermalpowercalculations.

TheORIGEN2codeisarigorousisotopegeneration anddepletion codethataccurately predictstheproductsandby-products offissionandtheresulting heatgeneration rates.Thedecayheatgeneration rateinthepoolconsistsoftwocomponents:

thedecayheatgenerated bypreviously discharged fuelassemblies, andthedecayheatgenerated byfreshly(recently) discharged assemblies.

Thedecayheatcontribution ofpreviously discharged fuelassemblies changesverylittleovershortperiodsoftime,andis,therefore, heldconstantintheanalyses.

Becauseofthenatureofexponential decay,thissimplification isconservative.

TheHoltecQAValidated LONGORcomputerprogram,whichincorporates theORIGEN2code,wasusedtocalculate thisdecayheatcomponent.

Thedecayheatcontribution ofthefreshlydischarged fuelassemblies changessubstantially overevenveryshortperiodsoftime.Thisdecayheatcontribution istherefore evaluated astime-varying.TheHoltecQAValidated BULKTEMcomputerprogram,thatincorporates theORIGEN2code,wasusedtocalculate thisdecayheatcomponent.

BulkSentFuelPitSFPTemeratureAnalsisMethodDuetothetime-varying decayheatcomponent, thetotaldecayheatisalsotime-varying.

ThebulkSFPtemperature istherefore calculated asafunctionoftime..Thefollowing energybalanceissolvedtoobtainthetemperature ateachinstantintime:where:CistheSFPthermalcapacity, Btu/oFTisthebulkSFPtemperature,

~F7isthetimeafterreactorshutdown, hrQ~~(r)isthedecayheatgeneration, Btu/hrQ~(T)istheSFPCSheatrejection, Btu/hrQ~>>(T)istheevaporative heatloss,Btu/hrTheevaporative heatlosstermincludesbothevaporative andsensibleheattransferfromthesurfaceoftheSFP.Theimplementation ofthistermhasbeenbenchmarked againstactualin-planttestdata.Thesolutionofthisfirst-order ordinarydifferential equationisperformed usingtheBULKTEMprogram.Time-to-Boil AnalsisMethodFollowing alossofforcedcooling,thecontinuing decayheatloadintheSFPwillcausethebulkSFPtemperature torise.Theequationenergy=balancethatdefinesthistransient phenomena is Attachment toAEP:NRC:1223E Page14similartotheordinarydifferential equationpresented above,butdoesnotincludetheQ~termanddoesincludeatime-varying SFPthermalcapacity, toaccountfortheevaporative waterlosses.Thetimeavailable forcorrective actionbeforebulkSFPboilingoccursisdetermined usingtheHoltecQAvalidated TBOILcomputerprogram.Thedecayheatgeneration andevaporative heatlosstermsinthisformulation areidentical tothosedefinedabove,exceptforthefollowing twodifferences:

Thedecayheat.iscalculated zsingthecorrelations ofUSNRCBranchTechnical PositionASB9-2insteadofORIGEN2.Noincremental creditisgivenforevaporative heatlossatSFPbulktemperatures greaterthan170'.LocalTemeraturesAnalsisMethodThedecayheatgenerated bythefuelassemblies storedintheSFPinducedabuoyancydrivenflowfieldupwardthroughthefuelrackcells.Coolerwaterissuppliedtothebottomoftherackscellsthroughtherack-to-wall gapsandrack-to-floor plenum.TheHoltecQAValidated THERPOOLcomputerprogramwasusedtoperformthisanalysis.

NRCUESTIONNO.7"Discusstheeffectofflowinducedvibration onthesteamgenerator U-bendtubesandtheheatexchanger inconsideration ofhighflowraterequiredforthepoweruprate."RESPONSETOUESTIONNO.7Thesteamgenerators evaluated forCookNuclearPlant'sunit2upratingprogramarethereplacement model51Fseries.AcompleteU-bendfatigueevaluation wasnotnecessary becauseoftheadvanceddesignfeaturesincorporated intothereplacement steamgenerators.

Oneoftheprerequisites forexcessive U-bendtubefatigueisdentinginthetoptubesupportplate.Thequatrefoil stainless steeldesignisexpectedtoinhibitfuturedenting.Inaddition, theanti-vibration bars(AVBs)incorporated intothereplacement steamgenerators wereinsertedtoauniformdepththreerowsdeeperthanconventional steamgenerators.

Uniforminsertion inhibitslocalflowpeaking,anddeeperinsertion addsmargintocalculated tubestability ratiosforthelargestradiustubenotsupported byAVBs.Boththesefactorsreducetheriskoffluidelastictubevibration, whichcouldleadtoexcessive U-bendtubefatigue.Flowinducedtubevibration andwearanalysisforCookNuclearPlant'sunit2model51Freplacement steamgenerators references normaldesignloadsforoperation at852.75MWtpersteamgenerator plusconsideration ofarangeofoperating conditions forwhichoperation isapprovedat900MWtpersteamgenerator.

Themainimpactoftherangeofoperating conditions wastherangeofoperating pressures considered, soexplicitcalculations primarily addresspressureloadingeffectsthataddtothe852.75MWtbase.Calculated resultsfortheadvancedmodel51Fdesignyieldlargemarginsrelativetofluidelasticinstability limits:themaximumstability ratiois0.36versusalimitof1.00.Upratingfrom852.75to900MWtwouldincreasethelimitingstability ratiotoonly0.38;aresultthatisstillmorethan2.5timesbelowthe Attachment toAEP:NRC:1223E Page15limit.Corresponding displacements duetoturbulence intheflowarewellbelow0.001inch.Basedontheseconsiderations, thereplacement steamgenerators atCook'uclear Plant'sunit2areconsidered tobeeffectively designedforthehighflowratesrequiredforthepower,uprate to3600MWt.