ML17255A590

Proposed Tech Spec Allowing Insertion of 14x14 Optimized Fuel Assembly Design for Cycle 14 Reload

ML17255A590
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Site:	Ginna
Issue date:	12/20/1983
From:	ROCHESTER GAS & ELECTRIC CORP.
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TECHNICAL SPECIFICATIONS

1.0 DEFINITIONS

1.2Thefollowing termsaredefinedforuniforminterpretation ofthespecifications.

ThermalPowerTheratethatthethermalenergygenerated bythefuelisaccumulated bythecoolantasitpassesthroughthereactorvessel.Reactor0eratinModesModeReactivity akk'oolantTemperature 1.3Refueling ColdShutdownHotShutdownOperating WQ5avg.avgavgavg140200540580Anyoperation withinthecontainment involving movementoffueland/orcontrolrodswhenthevesselheadisunbolted.

1.4~OerableCapableofperforming allintendedfunctions intheintendedmanner.S4OiOaO35 OGOOOa4+-----pBS<mo,'POPAQQCpDRPAmendment No.Proposed i'

regimeistermeddeparture fromnucleateboiling(DNB)andatthispointthereisasharpreduction oftheheattransfercoefficient whichwouldresultinhighcladtemperatures andthepossibility ofcladfailure.DNBisnot,however,anobservable parameter duringreactoroperation.

Therefore, theobservable parameters, thermalpower,reactorcoolanttemperature andpressurehavebeenrelatedtoDNBthroughtheW-3and/orWRB-1,DNBcorrelation.

TheseDNBcorrelations havebeendeveloped topredicttheDNBfluxandthelocationofDNBforaxiallyuniformandnon-uniform heatfluxdistributions.

ThelocalDNBheatfluxratio,definedastheratiooftheheatfluxthatwouldcauseDNBataparticular corelocationtothelocalheatflux,isindi-cativeofthemargintoDNB.AminimumvalueoftheDNBratio,MDNBR,isspecified sothatduringsteadystateoperation, normaloperational transients andanticipated transients, thereisa95%probability ata95%confidence levelthatDNBwillnotoccur.(1)ThecurvesofFigure2.1-1represent thelociofpointsofthermalpower,coolantsystempressureandaveragetemperature forwhichthisminimumDNBvalueissatisfied.

Theareaofsafeoperation isbelowtheselines.Proposed Sinceitispossibletohavesomewhatgreaterenthalpyrisehotchannelfactorsatpartpowerthanatfullpowerduetothedeepercontrolbankinsertion whichispermitted atpartpower,aconservative allowance hasbeenmadeinobtaining thecurvesinFigure2.1-1foranincreaseinFHwithdecreasing powerlevels.Rodwithdrawal blockandloadrunbackoccursbeforereactortripsetpointsarereached.TheReactorControlandProtective Systemisdesignedtopreventanyanticipated combination oftransient conditions forreactorcoolantsystemtemperature, pressureandthermalpowerlevelthatwouldresultintherebeinglessthana95%probability ata95%confidence levelthatDNBwouldnotoccur.(3)Amendment No.March30,1976Proposed

(1)FSAR,Section3.2.2(2)FSAR,Section3.2.1(3)FSAR,Section14.1.1Proposed

FIGURE2.1-1COREDNBSAFETYLIMITS2LOOPOPERATION 668'658645-635:638625628o6l56l86852400PSIA~00gPgg~l))~Sp~UNACCEPTABLE OPERATION 595598585588575578ACCEPTABLE OPERATION 8..l.2~3~45.6.7.8.91.l.llPOVER(traction of'ominal)

Amendment No.Proposed d.Overtemperature hT-DT[Kl+K2(PP)-K3(TT)1+x2SwherehT=indicated hTatratedpower,'F0T=averagetemperature,

'FT=5735FP=pressurizer

pressure, psigP=2235psig1K=1.201K2=.000900K=.02093vl=25sect2=5secandf(DI)isafunctionoftheindicated differ-encebetweentopandbottomdetectors ofthepower-range nuclearionchambers; withgainstobeselectedbasedonmeasuredinstrument responseduringplantstartuptestswhereqtandqbarethepercentpowerinthetopandbottomhalvesofthecorerespectively, andq+qbisthetotalcorepowerinpercentofratedpowersuchthat:(i)forqt-qblessthan+21percent,f(hI)=02~32Amendment No.March30,1976Proposed 0

(ii)foreachpercentthatthemagnitude ofqt-qbismorepositvethan+21percent,thebTtripsetpointshallbeautomatically reducedbyanequivalent of1.6percentofratedpower.Overpower bTDT[K4Kg(TT)K63S+1]+f(6?)whereTTlK~K5K6T3f(zI)indicated hTatratedpower,'Faveragetemperature,

'Findicated Tavgatnominalconditions atratedpower,'F1.077.0.0forT<T0.0011forT>T0.0262forincreasing T0.0fordecreasing T10secasdefinedin2.3.1.2.d.

2~33Amendment.

No.March30,1976proposed Pressurizer Wheneverthereactorisathotshutdownorcriticalthepressurizer shallhaveatleast100kwofheatersoperableandawaterlevelmaintained between12%and87%oflevelspan.Ifthepressurizer isinoperable duetoheatersorwaterlevel,restorethepressurizer tooperablestatuswithin6hrs.orhavetheRHRsysteminoperation withinanadditional 6hrs.BasesTheplantisdesignedtooperatewithallreactorcoolantloopsinoperation andmaintaintheDNBRabovethelimitvalueduringallnormal3.1-4bChangeNo.Amendment No.Proposed

MinimumConditions forCriticalit Exceptduringlowpowerphysicstests,thereactorshallnotbemadecriticalatatemperature below500'F,andifthemoderatetemperature coefficient ismorepositivethana.5pcm/'F(below70percentofratedthermalpower)b.0pcm/'F(atorabove70percentofratedthermalpower)3.1.3.23.1.3.3BasisInnocaseshallthereactorbemadecriticalaboveandtotheleftofthecriticality limitlineshownonFigure3.1-1ofthesespecifications.

Whenthereactorcoolanttemperature isbelowtheminimumtemperature specified above,thereactorshallbesubcritical byanamountequaltoorgreaterthanthepotential reactivity insertion duetodepressurization.

PrevioussafetyanalyseshaveassumedthatforDesignBasisEvents(DBE)initiated fromthehotzeropowerorhigherpowercondition, themoderator temperature coefficient (MTC)was.eitherzeroornegative.

Beginning inCycle14,thesafetyanalyseshaveassumedthatamaximumMTCof+5pcm/Fcanexistupto70%power.Analyseshaveshownthatthedesigncriteriacanbesatisfied fortheDBE'swiththisassumption.

Atgreaterthan(3)70%powertheMTCmustbe'zeroornegative.

Proposed Thelimitations onMTCarewaivedforlowpowerphysicsteststopermitmeasurement oftheMTCandotherphysicsdesignparameters ofinterest.

Duringthesetestsspecialoperating precautions willbetaken.3.1-19proposed Therequirement thatthereactorisnottobemadecriticalaboveandtotheleftofthecriticality limitprovidesincreased assurance thattheproperrelationship betweenreactorcoolantpressureandtemperature willbemaintained duringsystemheatupandpressurization.

Heatuptothistemperature willbeaccom-plishedbyoperating thereactorcoolantpumps.Ifthespecified shutdownmarginismaintained, thereisnopossibility ofanaccidental criticality asaresultofanincreaseinmoderator temperature oradecreaseofcoolantpressure.

Reference (1)FSARTable3.2.1-1(2)FSARFigure3.2.1-8(3)SafetyEvaluation forR.E.GinnaTransition to14x14Optimized FuelAssemblies, Westinghouse ElectricCorporation, November1983.Amendment No.Proposed

topublichealthandsafety.Wheneverchangesarenotbeing(1)madeincoregeometryonefluxmonitorissufficient.

Thispermitsmaintenance oftheinstrumentation.

Continuous moni-toringofradiation levelsandneutronfluxprovidesimmediate indication ofanunsafecondition.

Theresidualheatpumpisusedtomaintainauniformboronconcentration.

Theshutdownmarginasindicated willkeepthecoresubcritical, evenifallcontrolrodswerewithdrawn fromthecore.Duringrefueling, thereactorrefueling cavityisfilledwithapproxi-'ately230,000gallonsofboratedwater.Theboronconcentration ofthiswaterat2000ppmboronissufficient tomaintainthereactorsubcritical byatleast5%Dk/kinthecoldcondition withallrodsinserted(bestestimateof10%subcritical),

andwillalsomaintainthecoresubcritical evenifnocontrolrodswereinsertedintothereactor.Periodicchecksofrefueling (2)waterboronconcentration insurethepropershutdownmargin.Communication requirements allowthecontrolroomoperatortoinformthemanipulator operatorofanyimpending unsafecondition detectedfromthemaincontrolboardindicators duringfuelmovement.

Inadditiontotheabovesafeguards, interlocks areutilizedduringrefueling toinsuresafehandling.

Anexcessweightinterlock is3.8-3Proposed

providedontheliftinghoisttopreventmovementofmorethanonefuelassemblyatatime.Thespentfueltransfermechanism canaccommodate onlyonefuelassemblyatatime.Inadditioninterlocks ontheauxiliary buildingcranewillpreventthetrolleyfrombeingmovedoverstoragerackscontaining spentfuel.Theoperability requirements forresidualheatremovalloopswillensureadequateheatremovalwhileintherefueling mode.Therequirement for23feetofwaterabovethereactorvesselflangewhilehandlingfuelandfuelcomponents incontainment iscon-sistentwiththeassumptions ofthefuelhandlingaccidentanalysis.

References:

(1)FSAR-Section9.5.2(2)ReloadTransition SafetyReport,Cycle14(3)FSAR-Section9.3.13.8-4Amendment No.

IIaveragepowertiltratioshallbedetermined onceadaybyatleastoneofthefollowing means:a.Movabledetectors b.Core-exit thermocouples 3.10.2.2Powerdistribution limitsareexpressed ashotchannelfactors.Atalltimes,exceptduringlowpowerphysicsteststhehotchannelfactorsmustmeetthefollowing limits:F(Z)=(2.32/P)*K(Z)

QF(Z)=4.64*K(Z)

QFDH=1.66[1+.3(1-P)]forPR.5forP<~5for0<P<1.00wherePisthefractionofratedpoweratwhichthecoreisoperating, K(Z)isthefunctiongivenbyFigure3.10N3,andZistheheightinthecore.ThemeasuredFshallbeincreased bygreepercenttoyieldF.IfthemeasuredForFhexceedsthelimitinsI value,withduea118wance Pcrmeasurement error,themaximumallowable reactorpowerlevelandtheNuclearOverpower TripsetpointshallbereducedonpercentforeachpercentwhichFDorFexceedsthelimitingvalue,whichever ismor8restiRctive.

Ifthehotchannelfactorscannotbereducedbelowthelimitingvalueswithinoneday,theOverpower hTtripsetpointandtheOvertemperature hTtripsetpointshallbesimilarly reduced.3.10.2.3Exceptforphysicstests,ifthequadranttoaveragepowertiltratio,exceeds1.02butislessthan1.12,thenwithintwohours:a.Correctthesituation, orb.Determine bymeasurement thehotchannelfactors,andapplySpecification 3.10.2.2, orC.Limitpowerto75%ofratedpower.3.10-3Amendment No.g,P4Proposed

Ifthequadranttoaveragepowertiltratioexceeds1.02butislessthan1.12forasustained periodofmorethan24hourswithoutknowncause,orifsuchatiltrecursintermittently withoutknowncause,thereactorpowerlevelshallberestricted soasnottoexceed50%ofratedpower.Ifthecauseofthetiltisdetermined, continued operation atapowerlevelconsistent with3.10.2.2above,shallbepermitted.

Exceptforphysicstest,ifthequadranttoaveragepowertiltratiois1.12orgreater,thereactorshallbeputinthehotshutdowncondition utilizing normaloperating procedures.

Subsequent operation forthepurposeofmeasuring andcorrecting thetiltisper-mittedprovidedthepowerleveldoesnotexceed50%ofratedpowerandtheNuclearOverpower Trip"setpointisreducedby50%".Following anyrefueling andatleasteveryeffective fullpowermonththereafter, fluxmaps,usingthemovabledetectorsystem,shallbemadetoconfirmthatthehotchannelfactorlimitsofSpecification 3.10.2.2aremet.Thereference equilibrium indicated axialfluxdifference asafunctionofpowerlevel(calledthetargetfluxdifference) shallbemeasuredatleastonceperequivalent fullpowerquarter.Thetargetfluxdifference mustbeupdatedatleasteachequivalent fullpowermonthusingameasuredvalueorbylinearinterpolation usingthemostrecentmeasuredvalueandthepredicted valueattheendofthecyclelife.Exceptduringphysicstests,controlrodexercises, excoredetectorcalibration, andexceptasmodifiedby3.10.2.9through3.10.2.12, theindicated axialfluxdifference shallbemaintained withini5%ofthetargetfluxdifference (definesthetargetbandonaxialfluxdifference).

Axialfluxdifference forpowerdistribution controlisdefinedastheaveragevalueforthefourexcoredetectors.

Ifoneexcoredetectorisoutofservice,theremaining threeshallbeusedtoderivetheaverage.3.10-4Amendment No.+Proposed 3.00eI.jilI~I(I;<<Ie-'(illeIII~lI'~e~~~jII.~IIIe~:I'.!e~ee~~IgjI~.I.:)Ie<<e-ITee~II~~~IIe<<[~~~~~ONELOOPOPERATION 2.00~eO'jOJOeIsrhJf-/-1.00OJOCC3ODIll0.0l~I:II~~~~I;Il00I~IIIl>eI;i<<~,"~IJIeli<iislIeLII'Is>i~~e~I-.l.erlibel1000ell:Iej;I<<:.I<<II;Jlje50!I.'I'I'.!jile~~.e.I)'Ij.ejj<<fTMOLOOPOPERATION COOLANTBORONCONCENTRATION (PPN)RE(VIREOSHUT00MNMARGINF'TrierStn-rPROPOSED 1.500uFIGURE3.10-31.2500NORMALIZEO AXIALOEPENOENCE FACTORFORFQVS.ELEVATION 1.00000.7500Ihi00000.2500TOTAI.FO2'.320COREHEIGHT0.0006.00010.80012.000K{7)f.000f.0000.9400.6470.0CDCDCDEUCDCDCD~t'DEOCOREHEIGHT(FT)C)CDCDEX)CDamendment t]o.gg,gpPROPOSEDCDAJ AppendixBBeginning withthereloadforCycle14,scheduled forinsertion inthespringof1984,Rochester Gas6ElectricwillusetheWestinghouse Optimized FuelAssembly(OFA)14x14designwithnaturaluraniumaxialblankets.

Inordertostoreandusefuelassemblies ofthisdesignseveralchangestoGinnaTechnical Specifications arerequired.

OnFebruary23,1982,RG&Erequested achangetotheTechnical Specification topermitstorageofthehigherenrichment OFAfuelinthespentfuelpool.Inresponsetoquestions fromtheNRCstaffconcerning thissubmittal RGSEprovidedacriticality analysisofthenewfuelstorageracksonSeptember 12,1983.Attachedarethreereportscomprising thesafetyanalysispreparedbyWestinghouse coveringthetransition fromanallExxonfueledcoretoafullcoreoftheOFAdesign.Thissafetyanalysisisnotcyclespecific, butusesparameters whichwillbound,thoseexperienced duringthetransition period.Thesafetyanalysisiscomposedofasummaryofthemechanical, thermal-hydraulic andaccidentanalysisanddetailedresultsofthenon-LOCAandLOCAanalysis.

Theseanalysesincorporate theproposedchangestotheTechnical Specifications andshowthattheapplicable designcriteriafortheExxonandOFAaresatisfied..

Inbrief,theproposedchangestotheTechnical Specifications arethefollowing:

1.AllowingapositiveMTC(+5pcm'F)upto70%power.2.Areduction inshutdownmarginatEOCfrom1900,pcmto1800pcm.N3.AchangeintheFhHlimitsat,lessthan100%power.4.Achangeinthecoreprotection limits(OTDTandOPDTsetpointequations).

5.AdeletionofthelimitsonTargetAxialOffset.Thefirstfourchangesareincorporated intotheaccidentanalyses.

Thedeletionofthelimitontargetaxialoffset(TAO)isnottreatedexplicitly intheWestinghouse safetyanalysis.

Worstcasepowerdistributions thatboundanythat,wouldoccurduringoperation areassumedbyWestinghouse.

ForeveryreloadWestinghouse mustassurethatthepotential worstcasepowerdistribution doesnotexceedthoseassumedinthesafetyanalysis.

Therefore, thelimitation onTAOisunnecessary.

Thedeletionofthelimitations isconsistent withtheprovisions oftheStandardTechnical Specification.

0~y Fourmigsoxideassemblies (MOX)willremaininthecoreforCycle14..Theseassemblies aremechanically identical tothewestinghouse HIFARdesignusedasreloadfueltoGinnaprjy~toCycle8.Exxonpreviously hasperformed asafetyanalysisandconcluded onabestestimatebasisthatinamixedcoreconfiguration theflowtoeachassemblywaswithinonepercentofthecoreaverage.ApplyingaDNBRpenaltyequivalent toadecreaseinonepercentofflowtotheminimumDNBRforExxonfuelcalculated byWestinghouse indicates thatsufficient margintothedesignDNBRlimitexists.pferanalysesremainvalid,aspreviously approvedbytheNRC.

Reference 1.Letter,D.C.Ziemann,USNRCtoL.D.White,RG&EApril15,1980.2.R.E.GinnaNuclearPlantCycle8SafetyAnalysisReport,ExxonNuclearCompany,December, 1977.

Attachment CInaccordance with10CFR50.91thesechangestotheTechnical Specifications havebeenevaluated againstthreecriteriatodetermine iftheoperation ofthefacilityinaccordance withtheproposedamendment would:1.involveasignificant increaseintheprobability orconsequences ofanaccidentpreviously evaluated; or2.createthepossibility ofanewordifferent kindofaccidentfromanyaccidentpreviously evaluated; or3.involveasignificant reduction inamarginofsafety.Asoutlinedbelow,Rochester Gas6Electricsubmitsthat,theissuesassociated withthisamendment requestareoutsidethecriteriaof10CFR50.91,andtherefore, anosignificant hazardsfindingiswarranted.

Theproposedchangesarerequiredtoallowtheinsertion of,andsubsequent transition to,afullcoreoffuelassemblies oftheWestinghouse Optimized FuelAssemblyDesign(W-OFA).Thesechangeshavebeenincorporated intotheassumptions andmethodology usedbyWestinghouse toverifythataDesignBasisEventdoesnotcausetheappropriate acceptance criteriatobeviolated.

Inallcasestheassumptions, methodsandresultsareconsistent withWestinghouse standardreloadsafetyevaluation techniques andotherplantsubmittals totheNRCforinsertion ofW-OFA.Therefore, anosignificant hazardsfindingiswarranted forthefollowing reasons:Theinsertion ofW-OFAfuelassemblies willnotcauseanincreaseintheprobability ofanyaccident, andbecausetheacceptance criteriaaresatisfied, theconsequences ofanaccidentarenotincreased.

2.Thepossibility ofanewordifferent kindofaccidentisnotcreated.3.Whileitisnotpossibletosimplycomparetheresultstopreviousanalysesbecauseofthedifferent analytical techniques usedbyvendors,andtheconstantevolution oftheirmethods,theWestinghouse analysishasdemon-stratedthatappropriate marginexistsbetweenresultsandtheacceptance criteria.

SafetyEvaluation forR.E.GinnaTransition toWestinghouse 14xl4Optimized FuelAssemblies EditedbyJ.C.MillerRochester GasandElec:ricCorporation OocketNo.50-244Approved:

M.G.Arlotti,ManagerFuelLicensngandProoramSuppor.Westinghouse Elec=ricCorporation NuclearEnergySys.emsNuclearFuelDivis'.on P.0.Box39'.2Pittsburgh, PA152300710L:6

TABLEOFCONTENTSSectionPage

2.0 SUMMARYANDCONCLUSIONS

2"13.0MECHANICAL EVALUATION 3-14.0NUCLEAREVALUATION 4-15.0THERMALANDHYDRAULIC EVALUATION

6.0 ACCIDENTEVALUATION

6-

17.0REFERENCES

ATTACHMENT ATECHNICAL SPECIFICATIONS 7-1ATTACHMENT BNON-LOCAACCIDENTANALYSISFSARCHAPTER14ATTACHMENT CLOCAACCIDENTANALYSISFSARSECTIONS14.3.1/14.3.20710L:6 LISTOFTABLESTableNo.~Pae1FSARChapter14AccidentAnalysisSensitivity toProposedChanges6-6LISTOFFIGURES~FiureNu.~Pae1R.E.Ginna14x14OFA3-50710L:61-2

1.0INTRODUCTION

R.E.GinnaisaWestinghouse designedPWRandiscurrently operating withanallExxonNuclearCompany(ENC)14x14fueledcoreexceptforfourWestinghouse MixedOxide(MOX)assemblies.

R.E.GinnawaslastsuppliedwithWestinghouse fuelduringthecycle7reload.Cycle14isthefirstcycleinatransition phasefromENCtoWestinghouse 14x149gridOptimized FuelAssembly(OFA)fuelwithcoreloadingsrangingfromapproximately a15:oOFAand85MENCfueledcoretoeventually anall-OFA-fueled core.TheOFAfuelisverysimilartotheWestinghouse 7grid14x14lowparasitic fuelwhichhashadsubstantial operating performance inanumberofnuclearplants.Thisreportsummarizes thesafetyevaluation/analysis fortheregion-by-region reloadtransition fromthepresentENC-fueled coretoanall-Westinghouse OFA-fueled core.Thisreportexaminesthedifferences betweentheOFAandENCfuelassemblydesignsandevaluates theeffectofthesedifferences onthecoresduringthetransition toanall-OFA-fueled core.Theevaluation considers thestandardreloaddesignmethodsdescribed inReference 1,andthetransition effectsdescribed inChapter18ofReference 2.Reference 3presentstheoperating experience throughDecember1981ofOFAdemonstration assemblies.

Therearefour14x147griddemonstration assemblies thathavecompleted twocyclesofoperation (established burnup-20,000MWD/MTU).

Post-test examination atthecompletion ofthefirstcycleofirradiation indicated noabnormalities.

However,onedemonstration assemblyattheendofthesecondcycleofirradiation wasdamagedandremoved.Itwasconcluded thatthecauseofthedamagewasanisolatedeventandnotagenericOFAdesignproblem(seeLetterReportIT-83-222, "FailureInvestigation ofPointBeachUnit2OFARods,"July1983).Thedemonstration assemblies willhaveexperienced approximately 35,000MWD/MTUofburnupin1984.Sections3.0through6.0summarize theMechanical, Nuclear,ThermalandHydraulic, andAccidentEvaluations, respectively.

0710L:61"3

2.0 SUMMARYANDCONCLUSIONS

Consistent withtheWestinghouse standardreloadmethodology (Reference 1),parameters arechosentomaximizetheapplicability ofthetransition evaluations presented hereinforfuturecycles.Theobjective ofsubseqvent cyclespecificReloadSafetyEvaluation Reports(RSE's)willbetoverifythatapplicable safetylimitsaresatisfied basedonthereference evaluation/analyses established bythisreport.Thetransition designandsafetyevaluations presented hereinconsiderthefollowing nominaloperating conditions:

1520MWtcorepower,2250psiasystempressure, 573.5'Fvesselaveragecoolanttemperature (HFP)at2250psia,and174,000gpmprimarysystemthermaldesignflow.Theresultsofevaluation/analyses andtestsdescribed hereinleadtothefollowing conclusions:

1.TheWestinghouse OFAsaremechanically andhydraulically compatible withtheENCfuelassemblies, controlrods,andreactorinternals interfaces.

AlldesigncriteriafortheWestinghouse OFA'saresatisfied.

2.Generally changesinthenuclearcharacteristics becauseofthetransition fromENCto,OFAfuelwillliewithinthecycle-to-cycle variations observedforpastfuelreloaddesigns.Themoderator temperature coefficient isthemostsignificant exception tothis.SincetheH/UratioislargerforOFA,themoderator temperature coefficient ismorepositivethanobservedinpastWestinghouse fueledR.E.Ginnacores.Thishasbeenaccounted forintheaccidentevaluations.

0710L:62-1 3.Demonstration experience withWestinghouse OFAscontaining Zircaloygridsprovidesreasontoexpectsatisfactory operation fromOFAZircaloygrids.4.Theproposedtechnical specifications changes(Attachment A)areapplicable tocorescontaining anycombination ofOFAandENCfuelandplantoperating limitations willbesatisfiedwiththeseproposedchanges.5.Areference isestablished uponwhichtobasefuturecyclesafetyevaluations forWestinghouse OFAreloadfuel.0710L:62-2

3.0 MECHANICAL

EVALUATION Themechanical designrequirements andcriteriaapprovedbytheNRCforthe17x17OFAdesignaredescribed inReference 2.The14x14OFAdesignmeetsthesesamebasicdesignrequirements andcriteria.

ENC,inestablishing theirassemblydesign,demonstrated theirfuel'scompatibility withtheWestinghouse designwhichwastheinitialR.E.Ginnafuel.Westinghouse hasdemonstrated thecompatibility ofitsOFAdesignwithitsinitial9griddesignandhasperformed thereviewsdescribed belowtherebydemonstrating compatibility oftheWestinghouse OFAandENCfuelassemblies.

Thesimilarities betweentheOFAdesignandpreviousWestinghouse fuelincludethenumberoffuelrods,grids,guidethimblesandinstrumenta-tiontube.Thematerials ofthetopandbottomnozzles,fuelrod,andtopandbottomgridsarethesameinboththeWestinghouse OFAandinitialdesigns.Thedesignchangesbetweenthetwodesignsincludeareduction infuelrod,guidethimble,andinstrumentation tubediameters, andchangeofmaterial(SStozirc)andsevenintermediate gridsmadeofZircaloywiththethickness andheightincreased toretaintherequiredgridstrength.

Inadditiontothereduction ofthefuelroddiameter, 6.2inchesofnaturaluraniumpelletsreplacethestandardslightlyenrichedpelletsatbothendsofthefuelstack(axialblanket).

Alsochangedisthebottomnozzlewhichincludesalockingcupfeaturewhichfacilitates reconstitutability ofthefuel.assembly.

CThisisidentical tothestandardbottomnozzleexceptforthereconstitution feature.Thisdesignchangeofthebottomnozzleandgridmodifications wereevaluated anddetermined tohavenoimpactonthesafeoperation oftheplantandtheperformance ofthefuel.Thesechangesweremadeasallowedpertherequirements of10CFR50.59.

ThefueldesignbasesandcriteriaforWestinghouse 14x14OFA'sarethesamea0thosediscussed inSections4.2and4.4.1.2ofReference 2fortheWestinghouse 17x17OFAdesign.Verification thatthesecriteriaare0710L:63-1

metforWestinghouse fuelintheR.E.Ginnaplantisperformed usingthedesignmethodology andmodelsdiscussed inReference 1.Animprovedthermalsafetymodel,Reference 4,isbeingusedtogeneratefueltemperatures forsafetyanalysis.

ThetopandbottomgridsoftheOFAarefabricated fromInconelandthesevenintermediate gridsarefabricated fromZircaloy.

Theelevation ofthecenterline ofeachoftheOFAgridsmatchthatoftheENCgridsinordertominimizecrossflow duringoperation.

Figure1showstheOFA.TheZircaloygridheightis2.25inchesascomparedtotheInconelgridwhichis1.5inches.Thesedimensional changesweremadetocompensate fordifferences inmaterialstrengthproperties.

Eachfuelrodisgivensupportatsixcontactpointswithineachgridcellbyacombination ofsupportdimplesandsprings.TheWestinghouse OFAthimbletubesarefabricated fromZircaloy.

Therearetwosectionswithalargediameterandtwowithasmallerdiameter.

Thelargerdiameteratthetoppermitsrapidcontrolrodinsertion.

Bothofthereduceddiametersectionsproduceadashpotactionneartheendofthecontrolrodtraveltodecelerate thecontrolrodandreduceimpactforces.Theinstrumentation tubeisalsofabricated fromZircaloy.

Thistubeisofconstantdiameterandisdesignedtoaccept,theR.E.Ginnaincoreinstrumentation.

TheOFAinstrumentation tubehasa0.004inchdiametral increasewhencomparedtotheENCassemblyinstrumentation tube.Thereissufficient diametral clearance fortheinstrumentation thimbletotraversetheOFAinstrumentation tube.TheOFAtopandbottomnozzlesarefabricated fromstainless steel.Bothnozzlesindexthefuelassemblyinthecoreanddirectflowintoandoutoftheassemblythroughperforated nozzleplates.Theaxialspacingbetweenthetopandbottomnozzleisestablished toaccommodate thegrowthofthefuelrodsduetoirradiation effectsontheZircaloyfueltube.TheOFAbottomnozzledesignhasareconstitution featurewhichfacilitates easyremovalofthenozzlefromthefuelassembly.

0710L:e3-2 HolddownoftheOFAisprovidedbyfoursetsoftwoleafsprings.TheInconel,718springdesignpermitsbothahighspringrateandlargetravel,whichisrequiredtoaccommodate thedifference inthermalexpansion betweenthe2ircaloythimblesandthestainless steelreactorinternals.

Thisspringdesignalsoaccommodates thegrowthofthe2ircaloythimblesduringserviceandpreventsfuelassemblyliftoffduringnormaloperation.

Thefuelrodfrettingevaluation performed ontheWestinghouse 14x14sevengridOFAdesignhasshownthatevenwithnogridspringforceactingonthefuelrodbythefiveZircaloygridsatendoflife,thecladwearcriterion ismet.SincetheR.E.GinnaOFAdesigncontainsninegridsincluding sevenZircaloygrids,considerable additional wearmarginexistsfortheR.E.Ginnafueldesignthanfortheseven-grid OFAdesign.TherodbowbehavioroftheR.E.GinnaOFAisexpectedtobebetterthanthatofthe7gridWestinghouse fuelassembly.

TheR.E.GinnaOFAwillhavereducedgridspringforcesduetotheZircaloygridsshorterspanlengthsandahigherfueltubethickness-to-diameter ratiothanthe7gridfuelassembly.

Thesedesignchangesshouldresultinreducedrodbow.The-Zircalloy gridspringforcesarelowerduringservicethanthosetypically usedonInconelgrids.Therefore, lowerfrictionforcesaregenerated bythedifferential thermalexpansion andirradiation growthofthefuelrods.Thisresultsinlowerloadsappliedtotheskeletoncomponents thanarepresentinthe7gridWestinghouse assemblies.

Theskeletoncomponents are'conservatively designedtoaccepttheseloadswithmargin.0710L:63-3 e

Newthimble-plugging devicesandsecondary sourceassemblies weredesignedtobecompatible withtheOFA'sonly.Thesenewcorecomponents weredesignedtoaccommodate thegrowthofthefuel'ssembly andthedifference inthermal-.expansion betweentheZircaloythimblesofthefuelassemblyandstainless steelreactorinternals.

ThecontrolrodsusedintheR.E.Ginnareactorcorearecompatible withtheOFA.Thecurrentthimblepluggingdevicesandsecondary sourceswillcontinuetobeusedwithpreviously suppliedfuel.0710L:63-4

~II~~

0

4.0 NUCLEAREVALUATION

Thekeysafetyparameters evaluated fortheconceptual transition andfullOFAdesignsshowthattheexpectedrangesofvariation formanyoftheparameters willliewithinthenormalcycle-to-cycle variations observedforpastENCfuelreloaddesigns.Theparameters whichfalloutsideoftheserangesarethosewhicharesensitive tofueltype,e.g.,themoderator temperature

-coefficient.

Theaccidentevaluation, documented inSection6.0,hasconsidered rangesofparameters whichareappropriate forthetransition cyclesandbeyond.Themethodsandcoremodelsusedinthereloadtransition analysisareidentical tothoseemployedanddescribed inReferences 1,2,and5.ThesearethesamemethodsandmodelswhichhavebeenusedinotherWestinghouse reloadcycledesigns.Nochangestothenucleardesignphilosophy, methods,ormodelsarenecessary duetothetransition toOFAfuel.AnumberofchangestotheR.E.GinnaTechnical Specifications (Attachment A)willbeproposedaspartofthetransition toOFAfuel.Someofthesechanges,whetherdirectlyrelatedtoOFAfuelornot,impactthecorenucleardesign.Thesechangesinclude:(1)thepositivemoderator temperature coefficient (MTC)specification; and(2)the0.3multiplier intheFlimitfunction; (3)areduction intherequiredshutdownmargin(SDM)to1.8"hp.Powerdistributions andpeakingfactorsareprimailyloading"pattern-dependent.

Theusualmethods,suchasenrichment variation canbeemployedinthetransition andfullOFAcorestoensurecompliance withthepeakingfactorTechnical Specifications.

0710L'64-1

5.0 THERMALANDHYDRAULIC

EVALUATION HYDRAULIC COMPATIBILITY Thehydraulic characteri'sties ofanENCfuelassemblywereevaluated byperforming testsonaclean,unirradiated.

ENCfuelassemblyattheR.E.GinnasiteusingtheWestinghouse FuelAssemblyCompatiblity Systems(FACTS)loop.Asimilartestwasconducted onacleanunirradiated sevengridWestinghouse OFAinthesameloop.SincetheWestinghouse OFAdesignforR.E.Ginnaisslightlydifferent fromtheregularsevengridOFAtested(twoextramixingvanegridsandaslightlyshorterfuelrodlength)theeffectofthesedesigndifferences onthehydraulic characteristics ofthetestassemblywasaddressed.

Theresultsshowedthatthenetmismatchinoverallcorelosscoefficient waslessthanonepercent.Itwastherefore concluded thatthetwoassemblies arehydraulically compatible.

CALCULATIONAL METHODSThecalculational methodsusedintheanalysisemploythreechangesfrommethodspresently employedfortheR.E.Ginnathermal-hydraulic analysis.

Thesemethodsare:(1)theTHINCIVcomputercode,(2).theWRB-1DNBCorrelation fortheOFA,and(3)theImprovedThermalDesignProcedure (ITDP).TheTHINCIVprogramisusedtoperformthermal-hydraulic calculations.

TheTHINCIVcodecalculates coolantdensity,massvelocity,

enthalpy, voidfractions, staticpressure, andDNBRdistributions alongflowchannelswithinareactorcoreunderallexpectedoperating conditions.

TheTHINCIVcodeisdescribed indetailinReferences 8and9.0710L:6 Inthis.application, theWRB-1DNBCorrelation (Reference 6)isemployedinthethermal-hydraulic designoftheWestinghouse OFA.TheWRB-1Correlation (References 6and13)providesasignificant improvement inCriticalHeatFlux(CHF)predictions overpreviousDNBcorrelati'ons.

The17x17OFADNBtestsshowedthattheWRB-1Correlation correctly accounted forthegeometrychangesingoingfromthe17x170.374"rodODdesigntothe17x170.360"rodODdesign,andthatthedesignlimitof1.17w'asstillapplicable, Reference 13.The14x14OFAdesigninvolvedverysimilargeometrychangesfromthe7grid14x14STDfueldesign,namely,thereduction oftherodODfrom0.422"to0.400"andtheincorporation ofagriddesignwithanincreased heightandstrapthickness duetothechangefromInconeltoZircaloy.

Confirmatory DNBtestsperformed onthe)4xl4OFAtypicalcellgeometryverifiedthattheWRB-1Correlation accurately predicted CHFvaluesforthisgeometrytypeandthatthedesignlimitof1.17wasstillappropriate.

TheW-3DNBRCorrelation (Reference 14and15)wasusedinthedesignoftheENCfuelassembly.

Acorrelation limitDNBRof1.30isapplicable.

ThedesignmethodemployedtomeettheDNBdesignbasisistheITDP,Reference 7.Uncertainties inplantoperating parameters, nuclearandthermalparameters, andfuelfabrication parameters areconsidered statistically suchthatthereisatleasta95percentprobability thattheminimumDNBRwillbegreaterthanorequaltoDNBRforthepeakpowerrod.Plantparameter uncertainties areusedtodetermine theplantDNBRuncertainty.

This--DNBR'ncertainty, combinedwiththeGNBR..l,imit, establishes adesign.DNBRvaluewhichmustbemetinplantsafetyanalyses.

Sincetheparameter uncertainties areconsidered indetermining thedesignDNBRvalue,theplantsafetyanalysesareperformed usingvaluesofinputparameters withoutuncertainties.

Inaddition, thelimitDNBR-valuesareincreased tovaluesdesignated asthesafetyanalysi's limitDNBR's.Theplantallowance available betweenthesafetyanalysislimitDNBRvaluesandthedesignlimitDNBRvalues0710L:65-2 isnotrequiredtomeetthedesignbasis.Theallowance willbeusedforflexibility inthedesignandoperation ofthispl'ant.TheDNBRmarginisdefinedasSafetyanalysisDNBRvalue=1-MarginThetablebelowindicates therelationship betweenthecorrelation limitDNBR,designlimitDNBR,andthesafetyanalysislimitDNBRvaluesusedrsforthisdesign.W14x14OFATypicalThimbleENC14x14Typica1ThimbleCorrlelation LimitDesignLimitSafetyAnalysisLimit1.171.34.1.521.171.331.511.301.58'.62~1.301.50)1.54)ThemarginbetweenthedesignlimitandthesafetyanalysislimitDRBRismorethanenoughtooffsettherodbowpenaltyandthetransition corepenalty.RODBOWTheOFAforR.E.Ginnahasninegridsandanactivefuellengthof141.4inches.BasedonthecurrentNRCapprovedlicensing basis,Reference 16,thefractional closureatanygivenburnupfortheOFAforR.E.Ginnacanbecomparedtothatofthe7-gridassembly.

Thereleventparameters formakingsuchacomparison areL/I(L=span2lengthbetweengrids,I=fuelrodmomentofinertia)andtheinitial0710L:65-3 rod-to-rod gap.The1/IratioishigherfortheOFA,buttheinitialrod-to-rod gapisalsolarger,therefore, theseeffortsoffseteachother.Thefractional closureatanyburnupforthe9-gridWestinghouse OFAcanbeobtainedbydirectLscalingfromthatofthe7-gridassembly.

Theresultsindicated thatamaximumrodbowpenaltyof4.Z4DNBRisapplicable fortheR.E.GinnaOFA.Sufficient marginbetweenthesafetyanalysislimitDNBRandthedesignlimitDNBRhasbeenmaintained toaccommodate thispenaltyaswellasthetransition coreDNBpenalty.TheENCfuelassemblywouldbeexpectedtohavelessgapclosurethantheWestinghouse OFA,duetotheENCfuel'sthickercladdingasshowninReference 17.Dataobtainedbyotherinvestigations, References 18and19,showthatgapclosuresupto55%%uhavenomeasurable effectonDNB.Therefore, noresultant rodbowDNBRpenaltyisrequiredforENCfuel.TRANSITION COREDNBMETHODOLOGY TheOFAhasalargerhydraulic diameterandflowareacomparedtotheENCfuelassembly.

Thus,ifitisassumedthatthesamemassflowexistsinanENCassemblyandanOFAandthatthereisnoallowance forflowredistribution tooccur,theENCfuelassemblywillhaveahighervelocityintherodbundle.Thehighervelocity, togetherwiththelowervalueofrodbundlehydraulic

diameter, willcausetherodbundlepressuredroptobehigherintheENCfuelassembly.

Thus,forthesamevalueofmassflowrateintoanadjacentsetofENCandOFA,theflowwouldhaveatendencytoredistribute fromtheENCtotheOFAintherodbundleregion.Inthegriddedregions,however,theOFAhasahighervalueofmixingVanegridlosscoefficient.

Thiswillinducelocalized flowredistribution fromtheOFAtotheENCattheaxialzonesnearthemixingvanegridpositions.

0710L:65"4

Thenetconsequence ofthisflowredistribution onDNBRisprimarily duetotheeffectthisredistribution hasonthehotchannelmassvelocityandthelocalquality.Depending ontheaxiallocationoftheminimumDNBR,aDNBpenaltycanbepostualted oneithertypeoffuelassemblywhencomparedtoafullcoreofsimilarfuel.A2Xtransition coreDNBpenalty,ontheWestinghouse OFAandalXDNBpenaltyontheENCfuelweredetermined tobeapplicable byanalyzing different assemblyloadingpatternsatvariouscoreconditions inamannerconsistent withpreviously approvedanalysis, Reference 20.Thusthetransition coreswillbeanalyzedinthefollowing manner:theENCfuelinatransition corewillbeanalyzedasafullcoreofENCfuelapplyinga1%DNBtransition corepenlty;andtheOFAfuelinatransition corewillbeanalyzedasfullcoreofOFAfuelapplyinga2XDNBtransition corepenalty.TheDNBmarginspreviously described fortheENCandOFAfuelaremorethanenoughtoaccomodate thetransition corepenaltyandtherodbowpenalty.0710L65-5 0

6.0 ACCIDENTEVALUATION

Thissectionaddresses theimpactonaccidentanalysesofthefollowing proposedchangesforR.E.Ginna.oOFAPositiveMTCF>HMultiplier ChangeArevisedFSARChapter14giveninAttachment Bcontainsthedescriptions, methodology, resultsandconclusion foreachaccidentreanalyzed.

OFATheprincipal mechanical designcharacteristic oftheOFAdesignwhichcouldhaveaneffectonaccidents isthesmallerfuelrod.Thisleadstoahigherfuelrodtemperature, surfaceheatflux,andaDNBpenalty.Thelargerhydraulic diameterandlowercoolantflowvelocitycauseareduction inheattransferafterDNB.Thesmallerfuelrodalsoleadstoafasterheatuprateforseverereactivity transients suchasRodClusterControlAssembly(RCCA)ejection.

Asaresultofthesmallerfuelrod,forthesamepowerlevel,theOFAdesignwillhavealowerDNBratiothantheinitialdesign.TheDNBpenaltywasoffsetfortheOFAcorethroughtheuseoftheWRB-1DNBCorrelation, Reference 6,andtheITDP,Reference 7.Thosetransients impactedbytheOFAdesignareshowninTable1.Adiscussion ofLoss-of-Coolant Accidents (LOCA)isaddressed laterinthissection.PositiveMTCThepresentR.E.GinnaTechnical Specifications requiretheMTCtobezeroornegativeatalltimeswhilethereactoriscritical.

This0710L:66-1 requirement isoverlyrestrictive, sinceasmallpositivecoefficient atreducedpowerlevelscouldresultinasignificant increaseinfuelcycleflexibility, butwouldhaveonlyaminoreffectonthesafetyanalysisoftheaccidenteventspresented intheFSAR.TheproposedTechnical Specification change,giveninAttachment A,allowsa+5pcm/FMTCbelow70percentofratedpower,changingtoa0pcm/'FMTCat70percentpowerandabove.Apower-level dependent MTCwaschosentominimizetheeffectofthespecification onpostulated accidents athighpowerlevels.Moreover, asthepowerlevelisraised,theaveragecorewatertemperature becomeshigherasallowedbytheprogrammed averagetemperature fortheplant,tendingtomakethemoderator coefficient morenegative.

Also,theboronconcentration canbereducedasxenonbuildsintothecore.Thus,thereislessneedtoallowapositivecoefficient asfullpower'isapproached.

Asfuelburnupisachieved, boronisfurtherreducedandtheMTCwillbecomenegativeovertheentireoperating powerrange.TheimpactofapositiveMTContheaccidentanalysespresented inChapter14oftheR.E.GinnaFSAR,Reference 10,hasbeenassessed.

Thoseincidents whichwerefoundtobesensitive,to minimumornear-zero moderator coefficients werereanalyzed.

Ingeneral,theseincidents arelimitedtotransients whichcausereactorcoolanttemperature toincrease.

Withtheexceptions below,theanalysespresented hereinwerebasedona+5pcm/~FMTC,whichwasassumedtoremainconstantforvariations intemperature.

Thebankwithdrawal fromsubcritical andcontrolrodejectionanalysesarebasedonacoefficient whichisatleast+5pcm/'Fatzeropowernominalaveragetemperature, andwhichbecomeslesspositiveforhighertemperatures.

Thisisnecessary sincetheTWINKLEcomputercode,onwhichtheanalysisisbased,isadiffusion-theory coderatherthanapoint-kinetics approximation andthemoderator temperature feedbackcannotbeartificially heldconstantwithtemperature.

Forall0710L:66-2 accidents whicharereanalyzed, theassumption ofapositiveMTCexistingatfullpowerisconservative, sinceasnotedinAttachment A,theproposedTechnical Specification requiresthatthecoefficient bezeroornegativeatorabove70percentpower.Accidents notreanalyzed includedthoseresulting inexcessive heat.removalfromthereactorcoolantsystemforwhichalargenegativeMTCisconservative, andthoseforwhichheatupeffectsfollowing reactortripareinvestigated, whicharenotsensitive tothemoderator coefficient.

F>HMultiplier ChangeAproposedchangefromK=0.2toK=0.3inthefollowing equationfortheNuclearHotChannelFactor(FH)wasevaluated withregardtoNitseffectonaccidentanalyses:

FhH-1.66[1.0+.3(1-P)3wherePisthefractionoffullpowerand.3isthepowercorrection constant.

Theeffectonaccidentanalysesisthroughthecoresafetylimitsatveryhighpressureandlowpowerlevels.Sincethesteamgenerator safety,valvespreventtheplantfromreachingtheselimitingconditions, theprotection setpoints areunaffected bythischange.Thechangesometimes impactstheaxialoffsetenvelopesuch,thatthef(EI)changes.However,nocreditforthef(hI)protection isassumedintheaccidentanalyses.

Therefore, thesafetyanalysesarenotimpactedbytheproposedFmultiplier change.0710L66"3

Non"LOCATheimpactof.theproposedchangesasidentified earlierinthissectionhasbeenassessedforthenon-LOCAasprovidedinChapter14oftheR.E.GinnaFSARgiveninAttachment B.Thefollowing accidents have.beenreanalyzed:

Uncontrolled RCCABankWithdrawal FromaSubcritical Condition Uncontrolled RCCABankWithdrawal atPowerRCCADropChemicalandVolumeControlSystemMalfunction StartupofanInactiveReactorCoolantLoopReduction inFeedwater EnthalpyIncidentExcessive LoadIncreaseIncidentLossofReactorCoolantFlow/Locked RotorLossofLossofExternalElectrical LoadNormalFeedwater/Station BlackoutRuptureofaSteamPipeRuptureofaControlRodMechanism Housing-RCCA EjectionForeachofthesafetycriteriaaccidents

analyzed, itwasfoundthattheappropriate aremet.'areBreakLOCAThelargebreakLOCAanalysisforR.E.Ginna,applicable totransition andfullOFAcorecycles,wasreanalyzed duetothedifferences betweenENCandWestinghouse OFAdesigns.Thisanalysisisconsistent withthemethodology employedinWCAP-9500, IReference CoreReort17x170timizedFuelAssembl.Thecurrently approved1981largebreakEmergency CoreCoolingSystem(ECCS)Evaluation Model,Reference 11,wasutilizedforaspectrumofcoldlegbreaks.TherevisedPADFuelThermalSafetyModel,Reference 4,generated theinitialfuelrodconditions.

TheR.E.Ginnaanalysiswasperformed foranassumedsteamgenerator tubeplugginglevelof.12%.0710L:66-4 ArevisedFSARChapter14.3.2giveninAttachment Ccontainsafulldescription ofthemethodsandassumptions utilizedfortheWestinghouse OFAECCSLOCAanalysis, andtheresultsoftheanalyses.

ThelargebreakOFALOCAanalysisforR.E.Ginnautilizing thecurrently approved1981evaluation modelresultedinaPCTof1833Fforthe0.4COLOCAcaseatatotalpeakingfactorof2.32.AdditionoftheUPIpenaltyof21FresultsinafinalPCTof1854'F.Thesmallimpactofcrossflow fortransition corecyclesisconservatively evaluated asatmosta4'FeffectontheWestinghouse fuel,whichiseasilyaccommodated inthemarginto10CFR50.46limits.SmallBreakLOCAThesmallbreakLOCAanalysisforR.E.Ginnaapplicable totransition andfullOFAcorecycles,wasreanalyzed duetothedifferences betweenEHCandWestinghouse OFAdesigns.Thisisconsistent withthemethodology employedinWCAP-9500.

Thecurrently approvedOctober1975smallbreakECCSevaluation modelwasutilizedforaspectrumofcold-legbreaks,Reference 12.TherevisedPADfuelthermalsafetymodelgenerated theintialfuelrodconditions.

TherevisedFSARChapter14.3.1,giveninAttachment C,containsafulldescripition oftheanalysisandassumption utilizedfortheWestinghouse OFAECCSLOCAanalysis.

ThesmallbreakOFALOCAanalysisforR.E.Ginnautilizing thecurrently approved1975SmallBreakEvaluation modelresultedinaPCTof1092Fforthe6inchdiametercoldlegbreak.Theanalysisassumedtheworstsmallbreakpowershapeconsistent withaLOCAFenvelopeof2.32atcoremidplaneqelevation and1.5atthetopofthecore.AnalysesshowthatthehighandlowheadportionsoftheECCS,togetherwiththeaccumulators, providesufficient corefloodingtokeepthecalculated PCTwellbelowtherequiredlimitsof10CFR50.46.Adequateprotection istherefore affordedbytheECCSintheeventofasmallbreakLOCA.0710L:66-5 TABLE1FSARCHAPTER14ACCIDENTANALYSISSENSITIVITY TOPROPOSEDCHANGESAccidents OFA+MTC1.Uncontrolled RodWithdrawal fromaSubcritical Condition FSARSection14.1.1..2.Uncontrolled RCCAWith-drawalatPower.FSARSection14.1.2.3.RodClusterControlAssembly(RCCA)DropFSARSection14.1.4.4.ChemicalandVolumeControlSystemMal-functionFSARSection14.1.5.5.StartupofanInactiveReactorCoolantLoopFSARSection14.1.7.6.Reduction inFeedwater EnthalpyIncidentFSARSection14.1.10.7~Excessive LoadIncreaseIncidentFSARSection14.1.11.0710L:66"6 TABLE1(Con't)FSARCHAPTER14ACCIDENTANALYSISSENSITIVITY TOPROPOSEDCHANGES(CONTINUED)

Accidents OFA+MTC8.LossofReactorCoolantFlowFSARSection14.1.6..9.LossofExternalElectrical LoadFSARSection14.1.8.10.LossofNormalFeed-waterFSARSection14.1.9.11.LossofAllACPowertotheStationAux-iliariesFSARSection14.4.12.12.RuptureofaSteamPipeFSARSection14.2.5.13.RuptureofaControlRodMechanism Housing-RCCAEjectionFSARSection14.2.6.14.LOCAFSARSection14.3.10710L:66"7 01'

7.0REFERENCES

1.Bordelon, F.M.,etal.,"Westinghouse ReloadSafetyEvaluation Met'hodology,"

WCAP-9272 (Proprietary) andWCAP-9273 (Non-Proprietary),

March1978.2.Davidson, S.L.;Iorii,J.A.,"Reference CoreReport-17xl7Optimized FuelAssembly,"

WCAP-9500-A, May1982.3.Skaritka, J.;Iorii,J.A.,"Operational Experience withWestinghouse Cores,"WCAP-8183, Revision12,August1983.4.LetterfromE.P.Rahe(Westinghouse) toC.O.Thomas(NRC),NS-EPR-2673, "Westinghouse RevisedPADCodeThermalSafetyModel,"WCAP"8720, Addendum2,October27,1982,(Proprietary).

5.Camden,T.M.,etal.,"PALADON-Westinghouse NodalComputerCode,"WCAP-9485A (Proprietary) andWCAP-9486A (Non-Proprietary),

December1979,andSupplement 1,September 1981.6.Motley,F.E.,etal.,"NewWestinghouse Correlation WRB-1ForPredicting CriticalHeatFluxInRodBundlesWithMixingVaneGrids,"WCAP-8762, July1976.7.Chelemer, H.,etal.,"Improved ThermalDesignProcedure,"

WCAP-8567, July1975.8.Chelemer, H.,etal.,"THINCIV-AnImprovedProgramforThermal-Hydraulic AnalysisofRodBundleCores,"WCAP-7956, June1973.9.Hochreiter, L.E.,etal.,"Application oftheTHINCIVProgramtoPWRDesign,"WCAP-8054, September 1973.10.FinalSafetyAnalysisReport,R.EDGinnaNuclearPowerPlant,DocketNo.50-244.0710L:67-1

7.0REFERENCES

(Continued) 11.Eicheldinger, C.,"Westinghouse ECCSEvaluation Model,1981Version,"

WCAP-9200-P-A (Proprietary),

WCAP-9221-A (Non-Proprietary)

Revision1,1981.12.Skwarek,R.J.;Johnson,W.J.;andMeyer,P.E.,"Westinghouse Emergency CoreCoolingSystemSmallBreak,"October1975Model,WCAP-8970-P-A (Proprietary) andWCAP-8971-A (Non-proprietary),

January1979.13.Davidson, S.L.;Iorii,J.A.(Eds.),"Verification TestingandAnalysesofthe17xl7Optimized FuelAssembly,"

WCAP-9401-P-A andWCAP-9402, March1979.14.Tong,L.S.,"Critical HeatFluxesinRodBundles,TwoPhaseFlowandHeatTransferinRodBundles,"

AnnualASMEWinterMeeting,November1969,p.3146.15.Tong,L.S.,"BoilingCrisisandCriteriaHeatFlux,...,"AECOfficeofInformation

Services, TID-25887, 1972.16.LetterfromR.A.Clark(NRC)toW.G.CounsilNortheast NuclearEngineering Company,

Subject:

FuelSafetyEvaluation ofMillstone, UnitNo.2BSR,NRCDocketNo.50-336.P17.LetterfromG.F.Owsley(ENC)toT.A.Ippolito(NRC),XN-75-32, Supplement 1,"Computational Procedure forEvaluating FuelRodBowing,"July17,1979.18.Markowski, etal.,"EffectofRodBowingonCHFinPWRFuelAssemblies,"

ASMEpaper77-HT-91.

0710L:67"2

7.0REFERENCES

(Continued) 19.LetterfromJ.H.TaylortoS.A.Varga,"StatusReportonREDProgramsdescribed inSemi-Annual TopicalReportBAW-10097A,"

Revision2,November13,1978.20.LetterfromC.0.Thomas(NRC)toE.P.Rahe(Westinghouse),

"Supplemental Acceptance Number2forReferencing ofLicensing TopicalReportsWCAP-9500 andWCAPs9401/9402,"

January24,1983.0710L:67-3 ATTACHMENT AAlistoftheTechnical Specification changesrequiredbvtheuseoftheOFAdesignandapositiveMTCisprovidedasAttachment AtotheApplication forAmendment toOperating License.

0 ATTACHMENT BNON-LOCAACCIDENTANALYSISRESULTSFSARCHAPTER140710L6

ATTACHMENT BNON-LOCASAFETYANALYSISPresented inAttachment Barethosenon-LOCAaccidentanalysesoftheR.E.GinnaFSARChapter14impactedbytheproposedchangesasdetermined inSection5.Providedbelowisadiscussion ofinitialconditions, assumptions, andcomputer'codesusedtoanalyzetheaccidents presented.

Furtherdiscussion isprovidedforeachindividual analysis.

Sectionnumbersinthisappendixcorrespond tothoseusedintheFSAR.B-10710L:6

~J TABLEOFCONTENTSSectionDescription Page14.0AccidentAnalysis14-114.1.1Uncontrolled RCCAWithdrawal FromASubcritical Condition 14.1.1-114.1.2Uncontrolled RCCAWithdrawal AtPower14.1.2-114.1.4RodClusterControlAssembly(RCCA)Drop14.1.4-114.1.5ChemicalandVolumeControlSystemMalfunction 14.1.5-114.1.6LossofReactorCoolantFlow14.1.6-114.1.8~~LossofExternalElectrical Load14.1.8"114.1.10Excessive HeatRemovalDueToFeedwater Temperature Decrease14.1.10-1 14.1.11Excessive LoadIncreaseIncident14.1.11-1 14.2.5RuptureofaSteamPipe14.2.5-114.2.6RuptureofaControlRodMechanism Housing-RCCA Ejection14.2.6-10710L:6

,4 LISTOFTABLESTableDescription Page14"1SummaryofInitialConditions andComputerCodesUsed14-2NominalVa1uesofPertinent PlantParameters forNon-LOCAAccidentAnalysis14"1214"3TripPointsandTimeDelaystoTripAssumedinAccidentAnalysis14-1314-4Determination ofMaximumOverpower TripPoint-PowerRangeNeutronFluxChannel-BasedonNominalS'etpoint Considering InherentInstrument Errors14-1514.1.1-1Time'Sequence ofEventsforUncontrolled RCCAWithdrawal FromaSubcritical Condition 14.1.2-1TimeSequenceofEventsforUncontrolled RCCAWithdrawal atPower14.1.2-814.1.6"1TimeSequenceofEventsforLossofReactorCoolantFlow14.1.6"10 14.1.6"2.

SummaryofLimitingResultsforLockedRotorAccident14.1.6-11 14F1.6-3TimeSequenceofEventsforLockedRotorIncident14.1.6-12 0710L:6 LISTOFTABLES(continued)

TableDescription Page14.1.8"1TimeSequenceofEventsforLossofExternalElectrical Load14.1'-614.1~11-1TimeSequenceofEventsforExcessive LoadIncreaseIncident14.1.11"4 14.2.5-1TimeSequenceofEventsforSteamline Rupture14.2.5-914.2.6"1Parameters UsedintheAnalysisoftheRodClusterControlAssemblyEjectionAccident14.2.6-14 14.2.6"2TimeSequenceofEventsRCCAEjection14.2.6-15 0710L:6 LISTOFFIGURESFigureDescription Page14-1CoreLimitsandOverpower-Overtemperature hTSetpoints 14-17.14.2Reactivity Coefficients UsedinNon-LOCASafetyAnalysis14-1814-3Reactivity Insertion SCRAMCurvesUncontrolled RCCABankWithdrawal FromSubcritical 14.F1-914.1.1-2Uncontrolled RCCABankWithdrawal FromSubcritical 14.1.1-10 14.1.2-1Uncontrolled RCCABankWithdrawal AtPower,MaximumFeedback, 100%Power,90pcm/sec14'.2-914.1.2-2Uncontrolled RCCABankWithdrawal atPower,MaximumFeedback, 100ÃPower,90pcm/sec14.1.2-10 14.1.2-3Uncontrolled RCCABankWithdrawal atPower,MaximumFeedback, 1005Power,90pcm/sec14.1.2-11 14.1.2-4Uncontrolled RCCABankWithdrawal atPower,MaximumFeedback, 1005Power,7pcm/sec14.1.2-12 iv0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.1.2-5Uncontrolled RCCABankWithdrawal atPower,MaximumFeedback, 1005Power,7pcm/sec14.1.2"13 14.1.2-6Uncontrolled RCCABankWithdrawal atPower,MaximumFeedback; 10P%%dPower,7pcm/sec14.1.2-14 14.1.2"7Uncontrolled BankWithdrawal From100KPower14.1.2"15 14.1.2-8Uncontrolled BankWithdrawal From60%Power14.1.2-16 14.1.2-9Uncontrolled BankWithdrawal From10/Power14.1.2"17 14.1.4-1DroppedRod-100pcm14.1.4-414.1.4-2DroppedRod-100pcm14.1.4-514.1.4-3.Dropped'od "100pcm14.1.4-614.1.4-4DroppedRod-800pcm14.1.4"714.1.4"5DroppedRod-800pcm14.1.4-814.1.4-6DroppedRod-800pcm14.1.4-90710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.1.6-1FullLossofFlow14.1.6-13 14.1.6"2FullLossofFlow14.1~6-1414.1.6-3FullLossofFlow14.1.6-15 14.1.6-4PartialLossofFlow14.1.6-16 14.1.6-5PartialLossofFlow14.1.6-17 14.1.6-6PartialLossofFlow14.1.6-18 14.1.6-7PartialLossofFlow14.1.6-19 14.1.6"8LockedRotor14.1.6"20 14.1.6-9,LockedRotor14.1.6-21 14.1.6-10 LockedRotor14.1.6-22 14.1.8"1LossofLoad-MinimumFeedbackWithAutomatic PressureControl14.1.8"914.1.8-2LossofLoad-MinimumFeedbackWithAutomatic PressureControl14.1.8-10 14.1.8-3LossofLoad-MinimumFeedbackWithAutomatic PressureControl14.1'-11vi0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.1.8"4LossofLoad-MaximumFeedbackWithAutomatic PressureControl14.1.8-12 14.).8-5LossofLoad-MaximumFeedbackWithAutomatic PressureControl14.1.8-13 14.1.8-6LossofLoad-MaximumFeedbackWithAutomatic PressureControl14.1.8-14 14.1.8-7LossofLoad-MinimumFeedbackWithoutPressureControl14.1.8-15 14.1.8-8LossofLoad-MinimumFeedbackWithoutPressureControl14.1.8-16 14.1.8-9LossofLoad-MinimumFeedbackWithoutPressureControl14.1.8-17 14.1.8"10 LossofLoad-MaximumFeedbackWithoutPressureControl14.1'-1814.1.8-11 LossofLoad-MaximumFeedback.

WithoutPressureControl14.1.8-19 14.1.8-12 LossofLoad-MaximumFeedbackWithoutPressureControl14.1.8"20 vii0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.1.10-1 Feedwater EnthalpyDecrease-Automatic RodControl14.1.10-5 14.1.10-2 Feedwater EnthalpyDecrease-Automatic RodControl14.1.10-6 14.1.10-3 Feedwater EnthalpyDecrease-Automatic RodControl14.1.10"7 14.1.10-4 Feedwater EnthalpyDecrease-ManualRodControl14.1.10"8 14.1.10"5 Feedwater EnthalpyDecrease-ManualRodControl14.1.10-9 14.1.10"6 Feedwater EnthalpyDecrease-ManualRodControl14.1.10"10 14.1.11"1 ExcessLoadIncrease"MinimumFeedbackWithoutRodControl14.1.11-5 14.1.11-2 ExcessLoadIncrease-MinimumFeedbackWithoutRodControl14.1.11-6 14.1.11-3 ExcessLoadIncrease-MinimumFeedbackWithoutRodControl14.1.11-7 viii0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.1.11-4 ExcessLoadIncrease-MaximumFeedbackWithoutRodControl14.1.11-8 14.1.11"5 ExcessLoadIncrease-MaximumFeedbackWithoutRodControl14.1.11-9 14.1.11-6 ExcessLoadIncrease"MaximumFeedbackWithoutRodControl14.1.11-10 14.1.11-7 ExcessLoadIncrease-MinimumFeedbackWithAutomatic RodControl14.1.11-11 14.1.11-8 ExcessLoadIncrease-MinimumFeedbackWithAutomatic RodControl14.1.11-12 14.1.11"9 ExcessLoadIncrease-MinimumFeedbackWithAutomatic RodControl14.1.11-13 14.1.11-10 ExcessLoadIncrease-MaximumFeedbackWithAutomatic RodControl14.).11-1414.1.11-11 ExcessLoadIncrease-MaximumFeedbackWithAutomatic RodControl14.1.11-15 14.1.11-12 ExcessLoadIncrease-MaximumFeedbackWithAutomatic RodControl14.1.11-16 14.2.5"1Steamline Rupture14.2.5-11 ix0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.2.5"2Steamline Rupture14.2.5"12 34.2.5"3Steamline Rupture-4.6ftBreakWithPower2LoopsinService14.2.5-1314.2.5-4Steamline Rupture-4.6ft2BreakWithPower2LoopsinService14.2.5-14 14.2.5"5Steamline Rupture-4.6ft2BreakWithPower14.2.5-15 14.2.5-6Steamline Rupture-4.6ft.2BreakWithPower14.2.5-16 14..2.5"7 Steamline Rupture-4.6ft2BreakWithPower14.2.5-17 14.2.5-8Steamline Rupture-BreakWithoutPower4.6ft,2LoopsinService14.2.5-18 14.2.5-9Steamline Rupture-4.6ft2BreakWithoutPower2LoopsinService14.2.5-19 14.2.5-10 Steamline Rupture-4.6ft2BreakWithoutPower2LoopsinService'4.2.5-20 14.2.5-11 Steamline Rupture-4.6ft2BreakWithoutPower2LoopsinService14.2.5-21 0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.2.5-12 Steamline Rupture-4.6ftBreakWithoutPower14.2.5"22 14.2.5-13 Steamline Rupture-FailedSafetyValve14.2.5-23 14.2.5-14 Steamline Rupture-FailedSafetyValve2LoopsinService14.2.5-24 14.2.5-15 Steamline Rupture-FailedSafetyValve2LoopsinService14.2.5-25 14.2.5-16 SteamlineRupture-FailedSafetyValve14.2.5-26 14.2.5-17 Steamline Rupture-FailedSafetyValve14.2.5-27 14.2'-18Steamline Rupture-4.6ftBreakWithPowerOneLoopinService14.2.5-28 14.2.5"19 Steamline Rupture-4.6ft2BreakWithPowerOneLoopinService14.2.5-29 14.2.5-20 Steamline Rupture-4.6ft2BreakWithPower'neLoopinService14.2.5-30 14.2.5-21 Steamline Rupture-4.6ft2BreakWithPowerOneLoopinService14.2.5"31 14.2.5-22 Steamline RuptureBreakWithPowerOneLoopinService14.2.5"32 xi0710L:6 LISTOFFIGURES(continued)

FigureDescription Page14.2.5-23 Steamline RuptureFailedSafetyValveOneLoopinService14.2.5-33 14.2.5-24 Steamline RuptureFailedSafetyValveOneLoopinService14.2.5-34 14.2.5-25 Steamline RuptureFailedSafetyValveOneLoopinService14.2.5-35 14.2.5"26 Steamline RuptureFailedSafetyValveOneLoopinSevice14.2.5-36 14.2.6-.1 RCCAEjection-Beginning ofLife,FullPower14.2.6-16 14.2.6-2RCCAEjection-Beginning ofLife,ZeroPower14.2.6-17 0710L:6X11

~Chater14InitialConditions Formostaccidents whichareDNB-limited, nominalvaluesofinitialconditions areassumed.Theallowances onpower,temperature, andpressurearedetermined onastatistical basisandareincludedinthelimitDNBR,asdescribed inWCAP-8567 (Reference 1).Thisprocedure isknownasthe"Improved ThermalDesignProcedure,"

andisdiscussed morefullyinSection4.Foraccidents whicharenotDNB-limited orinwhichtheImprovedThermalDesignProcedure isnotemployed, theinitialconditions areobtainedbyaddingthemaximumsteadystateerrorstoratedvalues.Thefollowing conservative steadystateerrorswereassumedintheanalysis:

1.CorePower+2percentallowance forcalorimetric error2.AverageReactorCoolant+4oFallowance forcontroller deadbandandmeasurement error.3.Pressurizer Pressure+30poundspersquareinch(psi)allowance forsteadystatefluctuations andmeasurement errorTables14-1and14"2summarize initialconditions'and computercodesusedintheaccidentanalysis, andshowwhichaccidents employedaDNBanalysisusingtheImprovedThermalDesignProcedure.

0710L:614-1

PowerDistribution Thetransient responseofthereactorsystemisdependent ontheinitialpowerdistribution.

Thenucleardesignofthereactorcoreminimizes adversepowerdistribution throughtheplacement ofcontrolrodsandroperating instructions.

Theconstantaxialoffsetcontrol(CAOC)strategyisusedforR.E.Ginna.Powerdistribution maybecharacterized bytheradialfactor(F)andthetotalpeakingaHfactor(F~).ThepeakingfactorlimitsaregivenintheTechnical Specifications andinSection5.0ofthisreport.Fortransients whichmaybeDNB-limited, theradialpeakingfactorisofimportance.

Theradialpeakingfactorincreases withdecreasing powerlevelduetorodinsertion.

ThisincreaseinFHisincludedintheaHcorelimitsillustrated inFigure14-1.Alltransients thatmaybeDNBlimitedareassumedtobeginwithaF>Hconsistent withtheinitial,powerleveldefinedintheTechnical Specifications.

TheaxialpowershapeusedintheDNBcalculations arediscussed inSection4.Theradialandaxialpowerdistributions described aboveareinputtotheTHINGCodeasdescribed inSection4.Fortransients whichmaybeoverpower limited,thetotalpeakingfactor(F~)isofimportance.

Alltransients thatmaybeoverpower limitedareassumedtobeginwithplantconditions including powerdistributions whichareconsistent withreactoroperation asdefinedintheTechnical Specifications.

Foroverpower, transients whichareslowwithrespecttothefuelrodthermaltimeconstant, forexample,theChemicalandVolumeControlSystemmalfunction thatresultsinadecreaseintheboronconcentration inthereactorcoolantincidentwhichlastsmanyminutes,andtheexcessive increaseinsecondary steamflowincidentwhichmayreachequilbrium withoutcausingareactortrip,.thefuelrodtherm'alevalua-tionsareperformed asdiscussed inSection4.Foroverpower transients 0710L:614"2 whicharefastwithrespecttothefuelrodthermaltimeconstant, forexample,theuncontrolled rodclustercontrolassemblybankwithdrawal fromsubcritical androdclustercontrolassemblyejectionincidents whichresultinalargepowerriseoverafewseconds,adetailedfuelheattransfercalculation mustbeperformed.

Althoughthefuelrodthermaltimeconstantisafunctionofsystemconditions, fuelburnupandrodpower,atypicalvalueatbeginning-of-life forhighpowerrodsisapproximately fiveseconds.h.Reactivit Coefficients AssumedintheAccidentAnalsesPThetransient responseofthereactorsystemisdependent onreactivity feedbackeffects,inparticular themoderator temperature coefficient andtheDopplerpowercoefficient.

Thesereactivity coefficients andtheirvaluesarediscussed indetailinSection3.0ofthemaintext.Intheanalysisofcertainevents,conservatism requirestheuseoflargereactivity coefficient values,whereasintheanalysisofotherevents,conservatism requirestheuseofsmallreactivity coefficient values.SomeanalysessuchaslossofcoolantfromcracksorrupturesintheReactorCoolantSystemdonotdependonreactivity feedbackeffects.Thejustification foruseofconservatively largeversussmallreactivity coefficient valuesistreatedonanevent-by"event basis.Insomecasesconservative combinations ofparameters areusedtoboundtheeffectsofcorelife,althoughthesecombinations maynotrepresent possiblerealistic situations.

Thelimitingvaluesofthemoderator densityandDopplerpowercoefficients usedinthesafetyanalysesareshowninFigure14-2.RodClustersControlAssemblInsertion Characteristics Thenegativereactivity insertion following areactortripisafunctionofthepositionversustimeoftherodclustercontrolassemblies andthevariation inrodworthas.afunctionofrodposition.

Withrespecttoaccidentanalyses, thecriticalparameter isthetimeofinsertion up'tothedashpotentry.0710L:614-3 TherodclustercontrolassemblypositionversustimeassumedinaccidentanalysesisshowninFigure14-3.Therodclustercontrolassemblyinsertion timetodashpotentryisnormalized to1.8seconds.Figure14-3alsoshowsthefractionoftotalnegativereactivity insertion versusnormalized rodposition.

Thiscurveisusedtocomputethenegativereactivity insertion versustimefollowing areactortrip.Atotalnegativereactivity insertion following atripof4percenthkisassumedinthetransient analysesexceptwherespecifically notedotherwise.

Thisassumption isconservative withrespecttothecalculated tripreactivity worthavailable.

TriPointsandTimeDelastoTriAssumedtoAccidentAnalsesAreactortripsignalactstoopentwotripbreakersconnected inseriesfeedingpowertothecontrolroddrivemechanisms.

Thelossofpowertothemechanism coilscausesthemechanisms toreleasetherodclustercontrolassemblies whichthenfallbygravityintothecore.Therearevariousinstrumentation delaysassociated witheachtripfunction, including delaysinsignalactuation, inopeningthetripbreakers, andinthereleaseoftherodsbythemechanisms.

Thetotaldelaytotripisdefinedasthetimedelayfromthetimethattripconditions arereachedtothetimetherodsarefreeandbegintofall.Limitingtripsetpoints assumedinaccidentanalysesandthetimedelayassumedforeachtrip'unction aregiveninTable14-3.Reference ismadeinthattabletoOvertemperature andOverpower hTtripshowninFigure14-1.Thisfigurepresentstheallowable ReactorCoolantLoopAverageTemperature andhTforthedesignflowandpowerdistribution, asdescribed inSection4,asafunctionofprimarycoolantpressure.

Theboundaries ofoperation definedbytheoverpower hTtripandtheovertemperature hTtriparerepresented as"Protection Lines"onthisdiagram.Theprotection linesaredrawnto0710L:614-4

includealladverseinstrumentation andsetpointerrorssothatundernominalconditions tripwouldoccurwellwithintheareaboundedbytheselines.TheutilityofthisdiagramisthatthelimitimposedbyanygivenDNBRcanberepresented asaline.TheDNBlinesrepresent thelocusofconditions forwhichtheDNBRequalsthelimjtvalue.ThelimitvaluesforWestinghouse fuel.are1.52(typicalcell)and1.51(thimblecell).ForEXXONfuel,thevaluesare1.62(typicalcell)and..1.54(thimblecell).AllpointsbelowandtotheleftofaDNBlineforagivenpressurehaveaONBRgreaterthanthelimitvalue.ThediagramshowsthatDNBisprevented forallcases,iftheareaenclosedwiththemaximumprotection linesisnottraversed bytheapplicable ONBRlineatanypoint.Theareaofpermissible operation (power,pressure, andtemperature) isboundedbythecombination ofreactortrips:highneutronflux(fixedsetpoint);

highpressure(fixedsetpoint);

lowpressure(fixedset-point);overpower andovertemperature hT(variable setpoints).

Thelimitvalue,whichwasusedastheONBRlimitforallaccidents analyzedwiththeImprovedThermalDesignProcedure (seeTable14-1),isconservative comparedtotheactualdesignDNBRvaluerequiredtomeettheONBdesignbasisasdiscussed inSection4.Thedifference betweenthelimitingtrippointassumedfortheanalysisandthenormaltrippointrepresents anallowance forinstrumentation channelerrorandsetpointerror../Nominal tripsetpoints arespecified intheplantTechnical Specifications.

Instrumentation DriftandCalorimetric Errors-PowerRaneNeutronFluxTheinstrumentation driftandcalorimetric errorsusedinestablishing thepowerrangehigh'eutron fluxsetpointarepresented inTable14-4.Thecalorimetric erroristheerrorassumedinthedetermination ofcore'tthermalpowerasobtainedfromsecondary plantmeasurements.

Thetotalionchambercurrent(sumofthetopandbottomsections) iscalibrated (setequal)tothismeasuredpoweronaperiodicbasis.0710L:614"5 Thesecondary powerisobtainedfrommeasurement offeedwater flow,feedwater inlettemperature tothesteamgenerators andsteampressure.

High-accuracy instrumentation isprovidedforthesemeasurements withaccuracytolerances muchtighterthanthosewhichwouldberequiredtocontrolfeedwater flow.ComuterCodesUtilizedSummaries ofsomeoftheprincipal computercodesusedintransient analysesaregivenbelow.Thecodesusedintheanalysesofeachtransient havebeenlistedinTable14-1.FACTRANFACTRANcalculates thetransient temperature distribution inacrosssectionofmetalcladU02fuelrodandthetransient heatfluxatthesurfaceofthecladusingasinputthenuclearpowerandtime-dependent coolantparameters (pressure, flow,temperature, anddensity).

Thecodeusesafuelmodelwhichexhibitsthefollowing featuressimultaneously:

l.Asufficiently largenumberofradialspaceincrements tohandlefasttransients suchasrodejectionaccidents.

2.Materialproperties whicharefunctions oftemperature andasophisticated fuel-to-clad gapheattransfercalculation.

3.Thenecessary calculations tohandlepost-DNBtransients:

filmboilingheattransfercorrelations, Zircaloy-water

reaction, andpartialmeltingofthematerials.

0710L:614"6 FACTRANisfurtherdiscussed inReference 2.LOFTRANTheLOFTRANprogramisusedforstudiesoftransient responseofaPWRsystemtospecified perturbations inprocessparameters.

LOFTRANsimulates amultiloop systembyamodelcontaining reactorvessel,hot-andcold-legpiping,steamgenerator (tubeandshellsides)andthepressurizer.

Thepressurizer heaters,spray,relief,andsafetyvalvesarealsoconsidered intheprogram.Pointmodelneutronkinetics, andreactivity effectsofthemoderator, fuel,boron,androdsareincluded.

Thesecondary sideofthesteamgenerator utilizesahomogeneous, saturated mixtureforthethermaltransients andawater-levelcorrelation forindication andcontrol.TheReactorProtection Systemissimulated toincludereactortripsonhighneutronflux,Overtemperature bT,Overpower hT,highandlowpressure, lowflow,andhighpressurizer level.Controlsystemsarealsosimulated including rodcontrol,steamdump,feedwater control,andpressurizer pressurecontrol.TheEmergency CoreCoolingSystem,including theaccumulators andupper-head injection, isalsomodeled.LOFTRANisaversatile programwhichissuitedtobothaccidentevalua-tionandcontrolstudiesaswellasparameter sizing.LOFTRANalsohasthecapability ofcalculating thetransient valueofDNBRbasedontheinputfromthecorelimitsillustrated inFigure14-1~Thecorelimitsrepresent theminimumvalueofDNBRascalculated fortypicalorthimblecell.LOFTRANisfurtherdiscussed inReference 3.0710L:614-7

TWINKLETheTWINKLEprogramisamulti-dimensional spatialneutronkineticscode.Thecodeusesanimplicitfinite-difference methodtosolvethetwo"group transient neutrondiffusion equations inone,two,andthreedimensions.

Thecodeusessixdelayedneutrongroupsandcontainsadetailedmulti-region fuel-clad-coolant heattransfermodelforcalculating pointwise Dopplerandmoderator feedbackeffects.Thecodehandlesupto2000spatialpoints,andperformsitsownsteadystateinitialization.

Asidefrombasiccross-section dataandthermal-hydraulic parameters, thecodeacceptsasinputbasicdrivingfunctions suchasinlettemperature,

pressure, flow,boronconcentration, controlrodmotion,andothers.Variouseditsareprovided, e.g.,channelwise power',axialoffset,enthalpy, volumetric surge,pointwise power,andfueltemperatures.

TheTWINKLECodeisusedtopredictthekineticbehaviorofareactorfortransients whichcauseamajorperturbation inthespatialneutronfluxdistribution.

TWINKLEisfurtherdiscussed inReference 4.THINCTheTHINGCodeisdescribed inReferences 18and19,(maintext).References 1.Chelemer, H.,etal.,"Improved ThermalOesignProcedure,"

WCAP-8567-P (Proprietary),

July,1975,andWCAP-8568 (Non-Proprietary),

July1975.2.Hargrove, H.G.,"FACTRAN-AFortran-IV CodeforThermalTransients inaUOFuelRod,"WCAP-7908, June1972.II0710L:6 3.Burnett,T.W.T.,etal.,"LOFTRANCodeDescription,"

WCAP-7907, June1972.4.Risher,D.H.,Jr.;Barry,R.F.,"TWINKLE"AMulti-Dimensional NeutronKineticsComputerCode,"WCAP-7979-P-A (Proprietary),

andWCAP-8028-A (Non-Proprietary),

January1975.0710L614-9

aTABI.E14"1SUMMAIIYOFINITIALCONOITIOITSINO COIIPUTER CODESUSEDAccidents improvedThermaIComputerDNB"DesignCodesUtilizedCorrelation Procedure InitiaINSSSTIEermaIPowerOutput(NWT)ReactorVesselCoolantFlow(GPH)VesselAveragoTemp(oF)Pressurizer Pressure(psia)Uncorltrolled RCCATWINKLEWithdrawal fromaFACTRANSubcritica ITHINGCondition WRB-IH-3Yes824325472250UncontroIledRCCALOFTRANWithdrawal atPowerHRB-IW-3Yes1520912152179200573.5562.9549.72250RodClusterLOFTRANControlAssemblyTHING(RCCA)DropWRB-1H-3Yes1520179200573'2250.ChemlcaIandNAVolumeControISysternHaIfunctIon0and1520~NANANAReduction inFeedwater EnthaIpyLOFTRANWRB"IW-3Yes0and1520179200547573.52250Excessive LoadLOFTRANIncr'ease WRB-IW-3Yes1520179200573'2250LossofLoadTurbineTripLOFTRANWRB-1W-3Yes1520179200573.52250SteamllneBreakLOFTRANTHINGW-3No174000800405472250"Wheretwocorrelations aroIisted,HRB-IappliostoWestinghouse fuelH-3appliestoEXXONfuel<<<<OnepumpInoporation.

AccountsforreverseI'lowthroughotherloop.0710L:6

TABLE,14-1(Continued)

SUHHARYOFIITIALCODITIONSA0COHPUTERCODESUSEDAccidents ImprovedThermaIComputerDNBDesignCodesUtilizedCorrelation Procedure InitlaIReactorNSSSThermaIVesselVesselPowerCoolantAveragePressurizer Output(HWT)Flow(GPH)Temp.(4F)Pressure(psia)LossofFlowLOFTRANFACTRANTHINGWRB-1W-3Yes1520179200573.52250LockedRotorLOFTRANFACTRANN/ANo1550174000577.52280JRodEJectlon.

TWINKLEFACTRANN/ANo1550and017400080040>>"547andN/A573.5+onepumpInoporation.

Accountsforreversoflowthroughotherloop.0710L:6

0TABLE14-2NOMINALVALUESOFPERTINENT PLANTPARAMETERS FORNON-LOCAACCIDENTS ANALYSIS" Parameter WithITDPWithoutITDPThermalOutputofNSSS(MWt)g1520/1520CoreInletTemperature

(~F)543.7543.7VesselAverageTemperature

('F)573.5573.5ReactorCoolantSystem*Pressure(psia)22502250ReactorCoolantFlowPerLoop(gpm)TotalReactorCoolantFlow(10LBM/hr)8960067.98700065.9Al(g~<SteamFlowfromNSSS(10LBM/hr)6.586.58SteamPressureatSteamGenerator 746.5Outlet(psia)IAssumed,Feedwater Temperature at432.3SteamGenerator Inlet(F)746.5432.3AverageCoreHeatFlux(BTU/hr-ft

)21894401894400Thenon-LOCAanalysesassumeasteamgenerator tubeplugginglevelof10;~.0710L:614-12 TABLE14"3TRIPPOINTSANDTIMEDELAYSTOTRIPASSUMEDINACCIDENTANALYSESTripFunctionLimitingTripPointAssumedInAnalsisTimeDelaysSecondsPowerrangehignneutron/flux,highsetting118K0.5Powerrangehighneutronflux,lowsetting35K0.5Overtemperature hTVariableseeFigure14-16.0Overpower hTVariableseeFigure14-12.0Highpressurizer pressure2425psia2.0Lowpressurizer pressure1775psia2.0aTotaltimedelay(including RTDtimeresponse, andtripcircuit,channelelectronics delay)fromthetimethetemperature difference inthecoolantloopsexceedsthetripsetpointuntiltherodsarefreetofa11.0710L:614-13 TABLE14-3(Continued)

TRIPPOINTSANDTIMEDELAYSTOTRIPASSUMEDINACCIDENTANALYSESTripFunctionLimitingTrip.PointAssumed~*'TimeDelaysSecondsLowreactorcoolantflow875loopflow(Fromloopflowdetectors)

1.0 Undervoltage

tripNotapplicable 1.5TurbinetripNotapplicable 2.0Low-lowsteamgenerator.

level0~ofnarrowrangelevelspan2.00710L:614-14 TABLE14-4DETERMINATION OFMAXIMUMOVERPOWER TRIPPOINT"POWERRANGENEUTRONFLUXCHANNEL-BASEDONNOMINALSETPOINTCONSIDERING INHERENTINSTRUMENT ERRORSVariableAccuracyofMeasurement ofVariableIerrorEffectOnThermalPowerDetermination

~error(Estimated)

(Assumed)

'Calorimetric ErrorsintheMeasurement ofSecondary SystemThermalPower:Feedwater temperature

+0.5Feedwater pressure(smallcorrection onenthalpy)

+0.50.3Steampressure(smallcorrection onenthalpy)

Feedwater flow+1.251.25AssumedCalorimetric Error(Xofratedpower)+2(a)Axialpowerdistribution effectsontotalionchambercurrentEstimated Error(Iofratedpower)AssumedError(XofratedPower)+5(b)0710L:614-15 TABLE14-4(Continued)

DETERMINATION OFMAXIMUMOVERPOWER TRIPPOINT-POWERRANGENEUTRONFLUXCHANNEL-BASEDONNOMINALSETPOINTCONSIDERING INHERENTINSTRUMENT ERRORSVariableAccuracyofMeasurement ofVariableXerrorEffectOnThermalPowerDetermination 5error(Estimated)

(Assumed)

Instrumentation channeldriftandsetpointreproducibi lityEstimated Error(5ofratedpower)AssumedError(Xofratedpower)+2(c)Totalassumederrorinsetpoint(a)+(b)+(c)PercentofRatedPowerNominalSetpoint109Maximumoverpower trippointassumingallindividual errorsaresimultaneously inthemostadversedirection 1180710L:614-16 Figure14-1GinnaCoreLimitsandOverpower

-Overtemperature aTSetpoints 6664625654O52o58~~48464442L((oO\'\'\'\Qbf.C'>HIT5A%50rs(h'\'\'\wVOOs:(h0(4((5256IC(.~"PCfpylef5unc,rgSf'af~gvalve>apI5755885655985956886856l86I56286256B8TevgloF')14-17 Figure14-2Reactivity Coefficients UsedinNon-LocaSafetyAnalysisQQ2L~-10.ge-18020406080100Power,XMostPositivesgttgeyeast500540560T,'F14-18

.Figure14-3Reactivity Insertion ScramCurves1.0.8.6.4.20.2.4.6.81.0Normalized Position1.21.0Q.8OO7g.6p.4OashpotII.20.2.4.6.81.01.2Norma1ized Time14-19.

14.1.1Uncontrolled RCCAWithdrawal fromaSubcritical Condition ARCCAwithdrawal incidentisdefinedasanuncontrolled additionofreactivity tothereactorcorebywithdrawal ofrodclustercontrolassemblies resulting inapowerexcursion.

Whilethepr'obabilityofatransient ofthistypeisextremely low,suchatransient couldbecausedbyamalfunction ofeitherthereactorcontrolorcontrolroddrivesystems.Thiscouldoccurwiththereactoreithersubcritical oI'tpower.The"atpower"caseisdiscussed inSection14.1.2.Reactivity isaddedataprescribed andcontrolled rateinbringing, thereactorfromashutdowncondition toalowerpowerlevelduringstartupbyRCCAwithdrawal.

Althoughthe'nitial startupprocedure usesthe"Imethodofborondilution, thenormalstartupiswithRCCAwithdrawal.

RCCAmotioncancausemuchfasterchangesinreactivity thancanbemadebychangingboronconcentration.

Therodclusterdrivemechanisms arewiredintopreselected groups,andthesegroupconfigurations arenotalteredduringcorelife.Therodst..aretherefore physically prevented fromwithdrawing inotherthantheirrespective groups.Powersuppliedtotherodgroupsiscontrolled suchthatnomorethantwogroupscanbewithdrawn atanytime.Theroddrivemechanism isofthemagneticlatchtypeandthecoilactuation issequenced toprovidevariablespeedrodtravel.Themaximumreactivity insertion rateisanalyzedinthedetailedplantanalysisassumingthesimultaneous withdrawal ofthecombination ofthetworodgroupswiththemaximumcombinedworthatmaximumspeed.0710L614'.1-1 Theneutronfluxresponsetoacontinuous reactivity insertion ischar-acterized byaveryfastrise,terminated bythereactivity feedbackeffectofthenegativeDopplercoefficient.

Thisself-limitation ofthepowerexcursion isofprimaryimportance, sinceitlimitsthepowertoatolerable levelduringthedelaytimeforprotective action.Ifacon-tinuousrodclustercontrolassemblywithdrawal accidentoccurs,thetransient isterminated bythefollowing automatic featuresofthereac-torprotection system:1.Sourcerangeleveltrip-actuatedwheneitheroftwoindepen-dentsourcerangechannelsindicates afluxlevelaboveapre-selected, manuallyadjustable value.Thistripfunctionmaybemanuallybypassedwheneitheroftheintermediate rangefluxchannelsindicateafluxlevelabovethesourcerangecutoffpowerlevel.Itisautomatically reinstated whenbothinter-mediaterangechannelsindicateafluxlevelbelowthesourcerangecutoffpowerlevel.2.Intermediate rangerodstop-actuatedwheneitheroftwoinde-pendentintermediate rangechannelsindicates afluxlevelaboveapreselected, manuallyadjustable value.Thisrodstopmaybemanuallybypassedwhentwooutofthefourpowerrangechannelsindicateapowerlevelaboveapproximately lOXpower.Itisautomatically reinstated whenthreeofthefourpowerrangechannelsarebelowthisvalue.3.Intermediate rangefluxleveltrip-actuatedwheneitheroftwoindependent intermediate rangechannelsindicates afluxlevelaboveapreselected, manuallyadjustable value.Thistripfunc-tionmaybemanuallybypassedwhentwoofthefourpowerrangechannelsarereadingaboveapproximately 10Kpowerandisauto-matically reinstated whenthreeofthefourchannelsindicateapowerlevelbelowthisvalue.0710L:614.1.1"2 4.Powerrangefluxleveltrip(lowsetting)-actuatedwhentwooutofthefourpowerrangechannelsindicateapowerlevelaboveapproximately 25~.Thistripfunctionmaybemanuallybypassedwhentwoofthefourpowerrangechannelsindicateapowerlevelaboveapproximately 105powerandisautomatically reinstated whenthreeofthefourchannelsindicateapowerlevelbelowthisvalue.5.Powerrangefluxleveltrip(highsetting)-actuatedwhentwooutofthefourpowerrangechannelsindicateapowerlevelaboveapresetsetpoint.Thistripfunctionisalwaysactive.MethodofAnalsisArodclustercontrolassemblywithdrawal accidentisanalyzedbythreedigitalcomputercodes.Theanalysisi'sperformed inthreestages:first,anaveragecorenuclearpowertransient calculation; thenanaveragecoreheattransfercalculation; andfinallytheDNBRcalcula-tion.Theaveragenuclearcalculation isperformed usingaspatialneutronkineticscode,TWINKLE,averagepowergeneration withtimein-cludingthevarioustotalcorefeedbackeffects,i.e.,Dopplerandmoderator reactivity.

TheFACTRANcodeisthenusedtocalculate thethermalheatfluxtransient, basedonthenuclearpowertransient calculated byTWINKLE.FACTRANalsocalculates thefuelandcladtemperatures.

TheaverageheatfluxisnextusedinTHING,References 18and19,fortransient DNBRcalculation.

ThisaccidentisanalyzedusingtheImprovedThermalDesignProcedure asdescribed inReference 6.Plantcharacteristics andinitialconditions arediscussed inSection4.0710L:614.1.1-3 Inordertogiveconservative resultsforastartupaccident, thefol-lowingadditional assumptions aremadeconcerning theinitialreactorconditions:

1.Sincethemagnitude ofthenuclearpower'peakreachedduringthe"initialpartofthetransient, foranygivenrateofreactivity insertion, isstrongly.

dependent ontheDopplercoefficient, conservatively lowvalues(lowabsolutemagnitude) asafunctionoftemperature areused.2.,Thecontribution ofthemoderator reactivity coefficient isnegligible duringtheinitialpartofthetransient becausetheheattransfertimebetweenthefuelandthemoderator ismuchlongerthanthenuclearfluxresponsetimeHowever,aftertheinitialnuclear'fluxpeak,the'ucceeding rateof'owerincrease's affectedbythemoderator reactivity coefficient.

Accord-ingly,aconservative valueof+5.0pcm/'Fatzeropower.is,used,becausethisyields.themaximumpeakheatflux.I3.Thereactorisassumedtobejustcriticalathotzeropower(no-loadaveragetemperature)'.

Thisassumption ismorecon-servativ'e thanthatofalower'nitial systemtemperature.

The.Ihigherinitial'ystem temperature yieldsalargerfuelwaterheattransfercoefficient, largerspecificheats,andalessnegative(smallerabsolutemagnitude)

Oopplercoefficient

-allofwhichtendtoreducetheOopplerfeedbackeffect,therebyincreasing theneutronfluxpeak.Theinitialeffective multi-plication factorisassumedtobe1.0,sincethisresultsinmaximumneutronfluxpeakingand,thus,themostseverenuclearpowertransient.

0710L:614.1.1"4 4.Reactortripis,assumedtobeinitiated bythepowerrangeflux.leveltrip(lowsetting).

Themostadversecombination ofin-strumentandsetpointerrors,aswellasdelaysfortripsignalactuation androdclustercontrolassemblyrelease,istakenintoaccount.A10percentincreaseisassumedforthepowerrangefluxtripsetpoint, raisingitfromthenominalvalueof25percentto35percent.Previousresults,however,showthatrise,intheneutronfluxissorapidthattheeffectoferrorsin.thetripsetpoint,on theactualtimeatwhichtherodsarereleasedisnegligible.

Inaddition, thereactortripinsertion characteristic isbasedontheassumption thatthehighestworthrodclustercontrolassemblyisstuckinitsfullywithdrawn position.

5.Themaximumpositivereactivity insertion rateassumed(97.5pcm/sec)isgreaterthanthatforthesimultaneous withdrawal ofthecombination ofthetwocontrolbankshavingthegreatestcombinedworthatmaximumspeed(45inches/minute).

6.Themostlimitingaxialandradialpowershapes,associated withhavingthetwohighestcombinedworthsequential banksintheirhighestworthposition, areassumedforDNBanalysis.

7.Theinitialpowerlevelwasassumedtobebelowthepowerlevel"9expectedforanyshutdowncondition (10ofnominalpower).Thecombination ofhighestreactivity insertion rateandlowestinitialpowerproducesthehighestpeakheatflux.8;Onereactorcoolantpumpisassumedtobeinoperation.

Thislowestinitialflowminimizes theresulting DNBR.0710L:614.1.1-5 ResultsThecalculated sequenceofeventsisshowninTable14.1.1-1.Figures14.1.1-1and14.1.1-2showthetransient behaviorfortheindicated reactivity insertion ratewiththeaccidentterminated byreactortripat35percentnominalpower.Thisinsertion rateisgreaterthanthatforthetwohighestworthcontrolbanks,bothassumedtobeintheirhighestincremental'orth region.Figure14.1.1-1showstheneutronfluxtransient.

Theenergyreleaseandthefueltemperature increases arerelatively small..Thethermalfluxresponse, ofinterestfordeparture fromnucleateboilingconsiderations, isshowninFigure14.1.1-1.Thebene-ficialeffectontheinherentthermallaginthefuelisevidenced bya(peakheatfluxlessthanthefull-power nominalvalue.Thereisalarge,'~margin-to-departure fromnucleateboilingduringthetransient, since<therodsurfaceheatfluxremainsbelowthedesignvalue,and.thereisahighdegreeofsubcooling atalltimesinthecore.Figure14.1.1"2showstheresponseofthehotspotaveragefuelandcladding'empera-ture.Theaveragefueltemperature increases toavaluelowerthanthenominalful1-powervalue.TheminimumDNBRatalltimesremainsabovethelimitvalue.Thecal-culatedsequenceofeventsforthisaccidentisshowninTable14.1.1-1.Withthereactortripped,theplantreturnstoa'stablecondition.

Theplantmaysubsequently becooleddownfurtherbyfollowing normalplantshutdownprocedures.

0710L:614.1.1"6 Conclusion Ifarodclustercontrolassemblywithdrawal accidentfromthesubcrit-icalcondition occurs,thecoreandthereactorcoolantsystemarenotadversely

affected, sincethedeparture fromnucleateboilingratioremainsabovethelimitvalue.I0710L:614.1.1-7 TABLE14.1.1-1TIMESEQUENCEOFEVENTSFORUNCONTROLLED RCCAWITHDRAWAL FROMASUBCRITICAL CONDITION EventTimeofEachEventSecondsInitiation ofuncontrolled rodwithdrawal, 97.5pcm/second reactivity insertion rate,from-910ofnominalpowerPowerrangehighneutronfluxlowsetpointreached8.09Peaknuclearpoweroccurs8.21Rodsbegintofallintocore8.59PeakheatfluxoccursMinimumDNBRoccurs10.25Peakcladtemperature occurs10.53Peakaveragefueltemperature occurs10.630710L:614.1.1-8 Figure14.1.1-'1 GinnaUncontrolled RCCABankWithdrawal fromSubcritical, 1.0O.1.01t.0000.80000.60000I+.iooOO.@XYLO0.0C5CDCDCIIDCDCICICICDCDCICICCJCDnICJCDCImTlHE(SEC)

Figure14.1.1-2GinnaUncontrolled RCCABankWithdrawal fromSubcritical i100.0INN.0a1750.01500.0M50.01000.0027.00UJaO.00oC00.00SCICtTilelSKC)CICIAJClCICIm14.I.1-10 14.1.2Uncontrolled RCCAWithdrawal atPowerAnuncontrolled RCCAwithdrawal atpowerresultsinanincreaseincoreheatflux.Sincetheheatextraction fromthesteamgenerator remainsconstant, thereisanetincreaseinreactorcoolanttemperature.

Un"lessterminated bymanualorautomatic action,thispowermismatchandresultant coolanttemperature risewouldeventually resultinDNB.Therefore, topreventthepossibility ofdamagetothecladding, theReactorProtection Systemisdesignedtoterminate anysuchtransient withanadequatemargintoDNB.Theautomatic featuresoftheReactorProtection Systemwhichpreventcoredamageinarodwithdrawal accidentatpowerincludethefollowing:

1.Nuclearpowerrangeinstrumentation actuatesareactortripiftwooutofthefourchannelsexceedanoverpower setpoint.

Lv2.ReactortripisactuatedifanytwooutoffourhTchannelsexceedanovertemperature hTsetpoint.

Thissetpointisauto-matically variedwithaxialpowerimbalance, coolanttemperature andpressuretoprotectagainstDNB.3.ReactortripisactuatedifanytwooutoffourhTchannelsexceedanoverpower hTsetpoint.

Thissetpointisautomati-callyvariedwithaxialpowerimbalance andcoolanttemperature toensurethattheallowable heatgeneration rate(kw/ft)isnotexceeded.

4.Ahighpressurereactortrip,actuatedfromanytwooutofthreepressurechannels, issetatafixedpoint.Thissetpressurewillbelessthanthesetpressureforthepressurizer safetyvalves.0710L614.1.2"1 5.Ahighpressurizer waterlevelreactortrip,actuatedfromanytwooutofthreelevelchannels, isactuatedatafixedset-point.Thisaffordsadditional protection forRCCAwithdrawal accidents.

Themannerinwhichthecombination ofoverpower andovertemperature hTtripsprovidesprotection overthefullrangeofreactorcoolantsystemconditions isillustrated inFigure14-1.Figure14-1presentsallowable reactorloopaveragetemperature andhTforthedesignpowerdistribution andflowasafunctionofprimarycoolantpressure.

Theboundaries ofoperation definedbytheoverpower hTtripandtheover-temperature hTtriparerepresented as"protection lines"onthisdiagram.Theseprotection linesaredrawntoincludealladverseins-trumentation andsetpointerrors,sothatundernominalconditions tripwouldoccurwellwithintheareaboundedbytheselines.Amaximumsteady-state operating condition forthereactorisalsoshownontheFigure.Theutilityofthediagramjustdescribed isinthefact'hattheopera-tinglimitimposedbyanygivenDNBratiocanberepresented asalineonthiscoordinate system.TheDNBlinesrepresent thelocusofcondi-tionsforwhichtheDNBRequalsthelimitvalue.AllpointsbelowandtotheleftofthislinehaveaDNBratiogreaterthanthisvalue.ThediagramshowsthatDNBisprevented forallcasesiftheareaenclosedwithinthemaximumprotection linesisnottraversed bytheapplicable DNBratiolineatanypoint.Theregionofpermissible operation (power,pressureandtemperature) iscompletely boundedbythecombination ofreactortrips:nuclearover-power(fixedsetpoint);

.highpressure(fixedsetpoint);

lowpressure0710L:6'4.1.2-2

(fixedsetpoint);

overpower andovertemperature hT(variable set-points).Thesetripsaredesignedtopreventoverpower andaDNBratiooflessthanthelimitvalue.MethodofAnalsisqCjVUncontrolled rodclustercontrolassemblybankwithdrawal isanalyzedbytheLOFTRANcode.Thiscodesimulates theneutronkinetics, reactorcoolantsystem,pressurizer, pressurizer reliefandsafetyvalves,pres-surizerspray,steamgenerator, andsteamgenerator safetyvalves.Thecodecomputespertinent plant'variables, including temperatures, pres-sures,andpowerlevel.Thecorelimits,asillustrated inFigure14-1,areusedasinputtoLOFTRANtodetermine theminimumdeparture fromnucleateboilingratioduringthetransient.

ThisaccidentisanalyzedwiththeImprovedThermalDesignProcedure asdescribed inReference 6.Plantcharacteristics andinitialconditions arediscussed inSection14.Inordertoobtainconservative valuesofdeparture fromnucleateboil-ingratio,thefollowing assumptions aremade:1.InitialConditions

-Initialreactorpower,reactorcoolantaveragetemperatures, andreactorcoolantpressureareassumedtobeattheirnominalvalues.Uncertainties ininitialconditions areincludedinthelimitDNBRasdescribed inReference 6.2.Reactivity Coefficients

-Twocasesareanalyzed.

a.MinimumReactivity Feedback-Apositive(Spcm/'F)modera"torcoefficient ofreactivity isassumed,corresponding tothebeginning-of-core-life.

AvariableDopplerpowercoef-ficientwithcorepowerisusedintheanalysis.

Aconser-vativelysmall(inabsolutemagnitude) valueisassumed.0710L:614.1.2-3

(/b.HaximumReactivity Feedback-Aconservatively largeposi-tivemoderator densitycoefficient andalarge(inabsolutemagnitude) negativeDopplerpowercoefficient areassumed.3.Therodclustercontrolassemblytripinsertion characteristic isbasedontheassumption thatthehighestworthassemblyis,stuckinitsfullywithdrawn position.

4.Thereactortriponhighneutronfluxisassumedtobeactuatedataconservative valueof1185ofnominalfullpower.Theovertemperature 4Ttripincludesalladverseinstrumentation andsetpointerrors;thedelaysfortripactuation areassumedtobethemaximumvalues.Nocreditwastakenfortheotherexpectedtripfunctions.

5.Themaximumpositivereactivity insertion rateisgreaterthanthatforthesimultaneous withdrawal ofthecombination ofthetwocontrolbankshavingthemaximumcombinedworthatmaximumspeed.Theeffectofrodclustercontrolassemblymovementontheaxialcorepowerdistribution isaccounted forbycausingadecreaseinover-temperature andoverpower bTtripsetpoints proportional toadecreaseinmargintoDNB.ResultsFigures14.1.2-1through14.1.2-3showtheplantresponse(including neutronflux,pressure, averagecoolanttemperature, anddeparture fromnucleateboilingratio)toarapidrodclustercontrolassemblywithdrawal incidentstartingfromfullpower.Reactortriponhighneutronfluxoccursshortlyafterthestartoftheaccident.

Sincethisisrapidwithrespecttothethermaltimeconstants oftheplant,smallchangesinTandpressureresult,andalargemargintoDNBisavgmaintained.

0710L:614.1.2-4 Theplantresponseforaslowcontrolrodassemblywithdrawal fromfullpowerisshowninFigures14.1.2-4through14.1.2-6.Reactortriponovertemperature hToccursafteralongerperiod,andtheriseintemperature andpressureisconsequently largerthanforrapidrodclustercontrolassemblywithdrawal.

Again,theminimumDNBRisgreaterthanthelimitvalue.Figure14.1.2-7showstheminimumdeparture fromnucleateboilingratioasafunctionofreactivity insertion ratefrominitialfull-power oper-ationfortheminimumandmaximumreactivity feedbackcases.Itcanbeseenthattworeactortripchannelsprovideprotection overthewholerangeofreactivity insertion rates.Thesearethehighneutronfluxa'ndovertemperature hTtripchannels.

TheminimumDNBRisneverlessthanthelimitvalue.Figures'4.

1.2-8and14.1.2-9showtheminimumdeparture fromnucleateboilingratioasafunctionofreactivity insertion rateforrodclustercontrolassembly'withdrawal incidents startingat60/and10'opowerrespectively.

Theresultsaresimilartothe100Kpowercase,exceptthatastheinitialpowerisdecreased, therangeoverwhichtheover-temperature hTtripiseffective isincreased.

Inneithercasedoesthedeparture fromnucleateboilingratiofallbelowtheDNBRlimitvalue.Inthereferenced figures,theshapeofthecurvesofminimumdeparture fromnucleateboilingratioversusreactivity insertion rateisduebothtoreactorcoreandcoolantsystemtransient responseandtoprotection systemactionininitiating areactortrip.0710L:614.1.2-5 Referring toFigure14.1.2-9, forexample,itisnotedthat:1~For.highreactivity insertion rates(i.e.,between-3"5-lxlOhk/second and-3.0x10hk/second),

reactortripisinitiated bythehighneutronfluxtripfortheminimumreactivity feedbackcases.Then'eutronfluxlevelinthecorerisesrapidlyfortheseinsertion rates,whilecoreheatfluxandcoolantsystemtemperature lagbehindduetothethermalcapacityofthefuelandcoolantsystemfluid.-Thus,thereac-,'oristrippedpriortosignificant increaseinheatfluxorwatertemperature withresultant highminimumdeparture fromnucleateboilingratiosduringthetransient.

Withinthisrange,asthereactivity insertion ratedecreases, coreheatfluxandcoolanttemperatures canremainmorenearlyinequili-briumwiththeneutronflux;minimumDNBRduringthetransient thusdecreases withdecreasing insertion rate.2.Withfurtherdecreaseinreactivity insertion rate,theover-temperature hTandhighneutronfluxtripsbecomeequallyeffective interminating thetransient (e.g.,at"5-3.0x10hk/second reactivity insertion rate).Theovertemperature hTreactortripcircuitinitiates areac-tortripwhenmeasuredcoolantloophTexceedsasetpointbasedonmeasuredreactorcoolantsystemaveragetemperature andpressure.

Itisimportant inthis'contexttonote,however,thattheaveragetemperature contribution tothecircuitislead-lagcompensated inordertodecreasetheeffectofthethermalcapacityofthereactorcoolantsysteminresponseto'owerincreases.

-5Forreactivity insertion ratesbetween-3.0x10hk/second

-6and-6.0xlOhk/second, theeffecti'veness oftheover-temperature hTtripincreases (intermsofincreased minimumdeparture fromnucleateboilingratio)due.tothefactthat,0710L:614.1.2-6 withlowerinsertion rates,thepowerincreaserateisslower,therateofriseofaveragecoolanttemperature isslower,andthelead-lag.compensation providedcanincreasingly accountforthecoolantsystemthermalcapacitylag.3.Formaximumreactivity feedbackcasesreactivity insertion rates-4lessthan-5.0x10hk/second, theriseinreactorco'olanttemperature issufficiently highsothatthesteamgenerator safetyvalvesetpointisreachedpriortotrip.Openingthesevalves,whichactasanadditional heatloadonthereactorcoolantsystem,sharplydecreases therateofriseofreactorcoolantsystemaveragetemperature.

Thisdecreaseinrateofriseoftheaveragecoolantsystemtemperature duringthetran-sientisaccentuated bythelead-lagcompensation, causingtheovertemperature hTtripsetpointtobereachedlaterwithresulting lowerminimumdeparture fromnucleateboilingratios.Figures14.1.2-7, 14.1.2-8,and14.1.2-9i.llustrate minimumdeparture fromnucleateboilingratiocalculated forminimumandmaximumreac-tivityfeedback.

Thecalculated sequenceofeventsforthisaccidentisshowninTable14.1.2-1.

Conclusions Intheunlikelyeventofacontrolrodwithdrawal

incident, fromfull-poweroperation orlowerpowerlevels,thecoreandreactorcoolantsystemarenotadversely affectedsincetheminimumvalueofDNBratioreachedisinexcessoftheDNBlimitvalueforallrodreactivity rates.Protection isprovidedbynuclearfluxoverpower andovertemperature hT.Additional protection wouldbeprovidedbythehighpressurizer level,overpower hT,andthehighpressurereactor.trip.Thepreceding sectionshavedescribed theeffectiveness oftheseprotection channels.

0710L:614.1.2-7 TABLE14.1.2-1TIMESEQUENCEOFEVENTSFORUNCONTROLLED RCCAWITHDRAWAL ATPOWER,EventTimeofEachEventSecondsCaseA:Initiation ofuncontrolled rodclustercontrolassemblywithdrawal atfullpowerandmaximumreactivity insertion rate(90pcm/sec)Powerrangehighneutronfluxhightrippointreached3.21Rodsbegintofallintocore3.71Minimumdeparture fromnucleateboilingratiooccurs4.00CaseB:Initiation ofuncontrolled rodclustercontrolassemblywithdrawal atfullpowerandatasmallreactivity insertion rate(7pcm/sec)Overtemperature hTreactortripsignalinitiated 264.7Rodsbegintofallintocore266.7Minimumdeparture fromnucl'cate boilingratiooccurs267.00882L:614.1~2-8 Figure14.1.2-1GinnaUncontrolled RCCABankMithdrawal atPowerfiAXIfiUM FEEDBACK100%Power,90pcm/secl.20001.0000x0X.soooo0Pg.SDOOO.ioooo4.200000.0l.2000l.0000.soooo~~".60000K.loooo.20000d00b0000000O00OO0O000000000~Alsrn0g000OTlHE(SEC)14.1.2-9 Figure14.1.2-2GinnaUncontrolled RCCAHankWithdrawal atPower100'4Power,90pcm/seciiAXIMUMFEEDBACK2S00.02100.02300.0I82200.0~~~2loo.02000..0$1800.0ILlSOO.0l700.0looo.00900.00LJ800.004J700.00600.00S00.00ioo.000088000080000000000000000'00~Alfoal+V1lD>00TIRElSKC)14.1.2-10 Figure14.1.2-3GinnaUncontro11ed RCCABankWithdrawa1 atPowerHAXItlUt1 FEEDBACK100Power,90pcm/sec620.00610.00600.00590.00560.00570.00560.00550.00510.005.00001.5000'.0000/3.5000I3.00002.50002.00001.5000l.20000O80000380000OO0O0O0tve'nco000O01TlHEtSEC)14.1.2-11 Figure14.1.2-4GinnaUncontrolled RCCABankWithdrawal atPowert.2000100%Power,7pcm/secllAXItlUf3 FEEDBACK.

xl.0000XCD.80000IPK.60000CD0CC.i0000~200000.0t.2000l.0000XF80000.60000.l0000I.200000.0~DCDCDCDCCJCDCDCCJCDCDCDTAHEtSEC)14.1.2-12

Figure14.1.2-5GinnaUncontr01led RCCAHankWithdrawa1 atPower100%Power,7pcm/sec.f1AXItlUfl FEEOBACK2500.0Zioo.02300.0Xc2200.021oo.o2000.0]800.041800.01?00.01000.00500.DD800.00X100.00ClSDD.00500.00ioo.00ClCICl3VIClClClClClT1HEt5KQClClAJClClV7Clm14.1.2-13 Figure14.1.2-6GinnaUncontrolled RCCABankWithdrawal atPower620.00100~Power,7pcm(secMAXIMUMFEEOBACK610.00600.00SSO.OO5sso.oo570.00560.00550.00Sio.001.00003.50003.0000O2.50OO2.0000l.50001.2OO0OCIlOCIClClOClClTlHE<SEC)OOOOAJOOOAlClClm14.1.2-]4 Figure14.1.2-7GinnaUncontrolled BankMithdrawal from100"PowerMaximumFeedback---MinimumFeedback2.12.01.9rrHighFlux1.8OTzT1.7~I10100Reactivity Insertion Rate,pcm/sec

Figure14.1.2-8GinnaUncontrolled BankWithdrawal from60~PowerMaximumFeedback---MinimumFeedbackIIIII2.52.42a32.22.12.0OTLT///////////////HighFluxOTaT/t/)q310100Reactivity Insertion Rate,pcm/sec14.1.2-16 Figure14.1.2-9GinnaUncontrolled Bank'Withdrawal from10Power,HaximumFeedback---MinimumFeedbackIII>>>>II~2>>52.42.32.2////If/0,///'////OTBT/HighFlUX///'L///2.12.010100Reactivity Insertion Rate,pcm/sec14.1.2-17 0

14.1.4RodClusterControlAssembly(RCCA)DropDroppingofafull-length RCCAoccurswhenthedrivemechanism is<deen-ergized.Thiswouldcauseapowerreduction andanincreaseinthehotchannelfactor.Ifnoprotective actionoccurred, theReactorControlSystemwouldrestorethepowertothelevelwhichexistedbeforetheincident.

ThiswouldleadtoareducedsafetymarginorpossiblyDNB,depending uponthemagnitude oftheresultant hotchannelfactor.IfanRCCAdropsintothecoreduringpoweroperation, itwouldbedetectedeitherbyarodbottomsignal,byanout-of-core chamber,orbyboth.Therodbottomsignaldeviceprovidesanindication signalforeachRCCA.Theotherindependent indication ofadroppedRCCAisob-tainedbyusingtheout-of-core powerrangechannelsignals.Thisroddropdetection circuitisactuateduponsensingarapiddecreaseinlocalfluxandisdesigned'suchthatnormalloadvariations donotcauseittobeactuated.

Aroddropsignalfromanyrodpositionindication channel,orfromoneormoreofthefourpowerrangechannels, initiates thefollowing pro-tectiveaction:reduction oftheturbineloadbyapresetadjustable amountandblockingoffurtherautomatic rodwithdrawal.

Theturbinerunbackisachievedbyactingupontheturbineloadlimitandontheturbineloadreference.

Therodwithdrawal blockisredundantly achieved.

Thetransient following adroppedRCCAaccidentisdetermined byade-taileddigitalsimulation oftheplant.Thedroppedrodcausesastepdecreaseinreactivity andthecorepowergeneration isdetermined usingtheLOFTRANcode.Theoverallresponseiscalculated bysimulating theturbineloadrunbackandpreventing rodwithdrawal.

Theanalysisispresented forthecaseinwhichtheloadcutbackveryclosely'matches thepowerdecreasefromthenegativereactivity foradroppedrod0882L:614.1.4-1 (800pcm)andalsoforthecaseinwhichtheloadcutbackisgreaterthanthatrequiredtomatchtheworthofthedroppedrod(100pcm).Inbothcasestheloadisassumedtobecutbackfrom100to84Koffullloadataconservatively slowrateofapproximately 1%persecond.Themostnegativevaluesofmoderator andDopplertemperature coeffi-cientsofreactivity areusedinthis'analysis resulting inthehighestheatfluxduringthetransient.

Theseareamoderator densitycoefficient ofreactivity of.43hp/gm/ccandaDopplertemperature

-5coefficient ofreactivity of-2.9x10hk/F.ThisaccidentisanalyzedwiththeImprovedThermalDesignProcedure asdescribed inWCAP-8567 (Reference 6).Plantcharacteristics andinitialconditions arediscussed inSection14.ResultsFigures14.1.4-1through14.1.4-3illustrate thetransient responsefol-lowingadroppedrodofworth100pcm.Thecoolanttemperature decreases initially duetothefactthatmoreenergyistakenoutfromthesecondary thanproducedintheprimary,thenincreases undertheinfluence ofthenegativereactivity effectofthemoderator andDopplertemperature coefficients.

Thepeakheatfluxfollowing theinitialresponsetothedroppedrodis97Kofnominal.Figures14.1.4-4and14.1.4-6illustrate thetransient responsefollow-ingadroppedrodofworth800pcm.Againthereactorcoolantaveragetemperature decreases initially, andthenincreases becauseofthenegativereactivity feedback.

Forthiscase,thepeakheatfluxfollowing theinitialresponsetothedroppedrodis84%ofnominal.Atthesametimethecoreaveragetemperature drops11.8'Fandthepres-suredrops130psia.0882L:614.1.4-2 Ananalysishasbeenmadeforthedroppedrodsattheconditions ofpeakheatfluxfollowing theinitialresponsetothedroppedrod.Thisanalysisincorporates theincreaseinradialhotchannelfactorcausedbythedroppedrods.ItwasfoundthattheDNBRdoesnotfallbelowthelimitvalue.Ananalysishasbeenalsomadeoftheamountofastatically misaligned RCCAforthemaximumfullpoweroperating conditions (100%power;corewaterinlettemperature of543.7'F;primarypressureof2250psia).Theeffectofthestaticmisaligned rodincidentwasrepresented byanincreaseintheradialheatfluxhotchannelfactor.Itwasfoundthat~theincreased FHcouldbeaccommodated withoutthe'NBRfallingbelowthelimitvalue.Conclusions Protection foradroppedRCCAisprovidedbyautomatic turbinerunbackandblockingofautomatic rodwithdrawal.

Astheanalysesshow,theprotection system,inconjunction withtheturbinerunback,protectsthecorefromDNB.Additionally, forastaticmisalignment atmaximumfull-powerconditions, DNBwillnotoccur.0882L:614.1.4-3

'igure14.1.4-1, GinnaDroppedRod-100pcm~~1.20001.0000C).80000ICC.6oooo.40000u.20000?0.01.20001.0000XC),80000?.').60000.loooo.200000.0ClEDC7C7C)C)C)O0OIO0OC)C>C>TlkE(SEC114.1.4-4 Figure14.1.4-2GjnnaDroppedRod100pcm2500.02400.02300.0cX2200.0CCleCCCCl4/1CC02300.02000.0>8OO.Ol800.01700.0620.006'!0.00sao.oo590.,00580.00510.00Sppp550.005i0.00oCDPPPPCVPPPPP~PCDCDCDPPCDmlACDCDCDCDCDCDCDTlÃE<SEC)14.1.4-5 Figure14.1.4-3GinnaDroppedRod-100pcml2000XCI1.0000F80000~60000.10000IC~200000.0C7CICICICICICICICICICImCICICICICICICICIIVICICICIIDCICICII1M'HAEC)14.1.4-6 Figure14.1.4-4GinnaDroppedRod-800pen1.2000z1.0000XCIZCDClILJCL.80000.60000cD.F0000.200000.01.20001.0000.80000.60000.10000lx~200000.0CDCDCICDCIClCICIJClCICIClmClClCIClv\ClClCDClClClClTIME(SEC)14.1.4-7 Figure14.1.4-5GinnaDroppedRod-800pcm2500.02400.02300.0Xc2200.02l00.02000.0gl900.0l800.0l700.0620.006lo.00600.00540.00580.00570.00560.00550.00540.00CDClClC)C)C'0y>C14.1.4-8 Figure14.1.4-'6 GinnaDroppedRod-.800pcm1.2000l.0000.SOOOOCIICC.60000.F0000E.200000.0CICICICIC)CICIC7CImCICICICICICICITIMEtCIiC>14.1.4-9 14.1.5ChemicalandVolumeControlSystemMalfunction Reactivity canbeaddedtothecorewiththeChemicalandVolumeControlSysembyfeedingreactormakeupwaterin.otheReactorCoolantSystemviathereac.ormakeupcontrolsystem.Thenormaldilutionprocedures callforalimitontherateandmagnitude foranyindividual dilution",

understrictadministrative controls.

Borondilutionisamanualopera-tion.Aboricacidblendsystemisprovidedtopermittheoperatortomatchtheconcentration ofreactorcoolantmakeupwatertohatexistinginthecoolantatthetime.TheChemicalandVolumeControlSys.emisdesignedtolimit,evenundervariouspostulated ailuremodes,the.potential rateofdilutiontoavaluewhich,afterindication throughalarmsandins.rumentation, providestheoperatorsuficienttimetocorrectthesituatjon inasafeandorderlymanner.ThereisonlyasinglecommonsourceofreactormakeupwatertotheReactorCoolantSystemfromthereactormakeupwaterstoragetank,andinadvertent dilutioncanbereadilyterminated byisolating thissinglesource.Theoperation ofthereactormakeupwaterpumpswhichtakesuctionfromhistankprovidestheonlysupplyofmakeupwatertotheReac.orCoolantSystem.InorderformakeupwatertobeaddedtotheReactorCoolantSystemthechargingpumpsmustberunninginadditiontothereactormakeupwaterpumps.Therateofadditionofunborated watermakeuptotheReac.orCoolantSys.emislimitedtothecapacityofthemakeupwaterpumps.Thislimitingadditionrateis120gpmfortworeactormakeupwaterpumps.iorotallyunborated watertobedelivered atthisratetotheReactorCoolantSystematpressure, twochargingpumpsmustbeoperatedatfullspeed.Normally, twochargingpumpsareoperating athalfspeed,whilethethirdpumpisidle.Theboricacidfromtheboricacidtankisblendedwiththereactormakeupwaterintheblenderandthecomposition isdetermined bythepresentflowratesofboricacidandreac.ormakeupwaterontheReactor0882L:614.1.5-1

MakeupControl.Twoseparateoperations arerequired.

First,theoper-atormusswitchfromtheautomatic makeupmodetothedilutemode.Second,thesartbuttonmustbedepressed.

Omittingei.herstepwouldpreventdilution.

Thismakesthepossibility ofinadvertent dilutionverysmall.Information onthestatusofthereacorcoolantmakeJpiscontinuously available totheoperator.

Lightsareprovidedonthecontrolboardtoindicatetheoperating condition ofpumpsintheChemicalandVolumeControlSystem.Alarmsareactuatedtowarntheopera.orifboricacidordemineralized waterflowratesdeviatefrompresetvaluesasaresultofsystemmalfunc.ion.

Tocoverallphasesofplantoperation, borondilutionduringrefueling, startup,andpoweroperation areconsidered inthisanalysis.

IMethodofAnalsisandResultsDilutionDurinRefuelinDuringrefueling'the followiag conditions exist:a)Oneresidualheatremovalpumpisrunningtoensurecontinuous mix-inginthereactorvessel,b)Thevalvein.hesealwaterheadertothereactorcoolantpumpsisclosed,c)Thevalvesonthesuc.ionsideofthechargingpumpsareadjustedforadditionofconcentrated boricacidsolution.

d)Theboronconcentration oftherefueling waterisaminimumof2000ppm,corresponding toashutdownof""percenthkwithallcontrolrodsin;periodicsamplingensuresthatthisconcentration ismain-tained,and0882L:614.1.5-2

~'1 e)NeutronsourcesareinsalledinthecoreandBFdetecorscon-3nec.edtoinstrumentation givingaudiblecountratesoprovidedirectmonitoring ofthecore.AminimumwatervolumeintheReactorCoolantSystemof2724ftis3considered.

Thiscorresponds tothevolumenecessary tofillthereac-torvesseltothemidplaneofthenozzlestoensuremixingviathepesidualheatremovalloop.Themaximumdilutionflowof120gpmanduniformmixingarealsoconsidered.

Administrative procedures limithechargingf1owtoonepumpavailable (twopumpslockedout).Themaximumdilutionflowassumesthesinglefailure,suchthattwopumpsaredelivering max.'mumflow.Theoperatorhaspromptanddefiniteindication ofanyborondiluionfromtheaudiblecountrateinsrumentation.

.Highcountrateisalarmedinthereactorcontainment andthemaincontrolroom.Thecountrateincreaseisproportional totheinversemultiplication fac.or.At1420ppm,forexample,atypicalcoreis4percentshutdownandthecountrateisincreased bya'factorof3.3overthecountrateat2000ppm.Theboronconcentration mustbereducedfrom.2000opmtoapproximately j500ppmbeforethereacorwillgocritical.

Thiswouldtakeatleast'8.8 minutes.This'isampletimefortheoperatortorecognize theaudiblehighcountratesignalandisolatethereac.ormakeupwatersourcebyclosingvalvesandstopping.hereactormakeupwaterpumps.DilutionDurinStar.uoPriortorefueling, theReactorCoolantSystemisfilledwithboratedwaterfromtherefueling waterstoragetank.Coremonitoring isbyexternalBF3detectors.

Mixingofreactorcoolantisaccomplished byoperation ofthereactorcoolantpumps.Again,themaximumdilutionflow(120gpm)isconsidered.

Thevolumeofreac.orcoolantis0882L:614.1.5"3 01'1 approximately 4255ftwhichishevolumeofheReac.orCoolant--.3Systemexcluding thepressurizer.

Thisvolumeaccountsfor10.percentsteamgenerator tubeplugging.

Highsourcelevelandallreac.ortripalarmsareeffec.ive.

Theminimumtimerequiredtoreducethereactorcoolantboronconcentration to1500ppm,wheretherectorcouldgocriticalwithal"1rodsattheinsertion.

limits,isabout.64.1minutes.Onceagain,thisrshouldbemorethanadequatetimeforoperatoractiontothe)'ighcountratesignal,andtermination ofdilutionflow.Inanycase,ifcontinued dilutionoccurs,thereactivi.yinsertion ra:eandconsequences thereofareconsiderably lessseverethanthoseassociated withtheuncontrolled rodwithdrawal analyzedinSection14.1.1,Uncontrolled RCCAWithdrawal fromaSubcritical Condition.

DilutionatPowerFordilutionatpower,itisnecessary thatthetime.oloseshutdownmarginbesufficient toallowidentification oftheproblemandtermination ofthedilution.

Asinthedilutionduringstar.upcase,theRCSvolumereduc.ion duetosteamgenerator tubepluggingisconsidered.

Theeffec.ive reaciviyadditionrateisafunctionof'thereac.orcoolanttemperature andboronconcentration.

Thereactivity insertion ratecalculated isbasedonaconse.vativelyhighvaluefortheexpectedboronconcentration atpower(1500ppm)aswellasaconservatively highchargingflowratecapacity(127gpm).Thereactori.sassumedtohaveallrods:outineitherautomatic ormanualcontrol.Withthereac.orin,manualcontrolandnoopera.oractionto.erminate Ithetransient, thepowerandtemperature risewillcausethereactortoreachthereactorprotection (i.e.,OThT,highnuclearflux)trip~isetpoint, resulting inareactortrip.Afterreac.ortripthereisatleast53.5minutesforoperatoractionpriortoreturnto0882L:614.1.5-4 t~.~0 criticality.

Theborondilutiontransient inthiscaseisesseniallytheequivalent toanuncontrolled rodwithdrawal

'atpower.Themhximumreac.ivity insertion rateforaborondilutiontransient isconservatively estimated tobe,1.6pcm/secandiswi.hintherangeofinsertion ratesanalyzedforuncontrolled rodwithdrawal atpower.Prior'toreachingthereactorprotection trip,theopera'orwillhavereceivedanalarmonOvertemperature hTandturbinerunback.>>

Withthereactorinautomatic control,aborondilutionwillresultina'owerandtemperature increasesuchthattherodcontroller willat.empttocomoensate byslowinsertion ofthecontrolrods.Tnisac.ionby.heI,controller willresultinrodinsertion limitandaxialfluxalarms.-Tneminimumtimetolosetheshutdownmarginatbeginning oflifewouldbegreaterthan54.4minutes.'he timewouldbesignificantly longeratendoflifeduetothelowinitialboronconcentration.1 Conclusions Becauseoftheprocedures involvedinthedilutionprocess,anerroneous dilutionisconsidered incredible.

Nevertheless, ifanunintentional dilutionofboroninthereactorcoolantdoesoccur,numerousalarmsandindications areavailable toaler.theoperatortothecondition."

Themaximumreactivity additionduetothedilutionisslowenoughtoaIllowtheoperatortodetermine thecauseoftheaddiionandtakecorrec.ive actionbeforeexcessive shutdownmarginislost.0882L:614.1.S"5 14.1.6LossofReactorCoolantFlowFlowCoas.down Acciden.s Aloss-of-coolant flowincidentcanresultfromamechanical or'lectrical failureinoneormorereac.orcoolantpumps,orfromafaulinthepowersupplytothesepumps.If,thereactor'is atpoweratthetimeoftheincident, theimmediate effectofloss"of-coolant flowisarapidincreaseincoolanttemperature.

Thisincreasecouldresulindepar.ure fromnucleateboiling(DNB)withsubsequent fueldamageifthereac:orisnot.rippedpromptly.

Thefollowing tripcircuitsprovidethenecessary protec:ion agains.alossofcoolantflowincidentandareactuatedby:1.Lowvoltageonpumppowersupplybus2.Pumpcircuitbreakeropening(lowfrequency on-pumppowersupplyn,busopenspumpcircuitbreaker)3.LowreactorcoolantflowThesetripcircuitsandtheirredundancy arefurtherdescribed inSec.ion7.2oftheFSAR,ReactorControlandProtec.ion System.Simultaneous lossofelectrical powertoallreactorcoolantpumpsatfullpoweristhemostseverecredibleloss-of"coolant flowcondit;on.

Forthiscondition reac.ortriptogetherwithflowsustained bytheincr.iaofthecoolantandrotatingpumppar.swillbesuficienttopreventfuelfailureandreactorcoolantsystemoverpressure andtopreventheONBratiofromgoingbelowthelimitvalue.MethodofAnalsis-Thefollowing loss-of-flow casesareanalyzed:

1.Lossoftwopumpsfromareactor'oolant system,heatoutputof1520MWtwithtwoloopsoperating.

0882L:614.1.6-1 2.Lossofonepumpfromareactorcoolantsystem,heatoutputof1520MWtwithtwoloopsooerating.

Theirstcaserepresents theworst,crediblecoolantflowloss.Thesecondcaseislesssevere.Lossofonepumpabovegpresetpowerlevelcausesareactortripbyalowflowsignal.Thepowerlevelabovewhichthistripoccursisassumedtobesetat49.~offullload.Thenormalpowersupplies.

forthepumpsarethetwobusesconnected tothegenerator, eachofwhichsuppliespowertooneofthetwopumps.Whenagenerator ripoccurs,thepumpsareautomatically transferred toabussuppliedfromexternalpowerlines.Therefore, thesimultaneous lossofpowertoallreac.orcoolantpumpsisahighlyunlikelyevent.Following anyturbinetrip,wheretherearenoelectrical faultswhichrequiretrippingthegenerator fromthenetwork,thegenerator remainsconnec.ed tothenetworkforaleastoneminute.Sincebothpumpsarenotonthesamebus,asinglebusfaultwouldnotresultinthelossofallpumps.Thistransient isanalyzedbythreedigitalcomputercodes.Firs.,theLOFTRANcodeisusedtocalculate theloopandcoreflowduring.hetransient, thetimeof,thereactortripbasedonthecalcula.ed flow,thenuclearpowertransient, andtheprimarysys.empressureandtemperature transients.

TheFACTRANcodeisthenusedtocalculate theheatfluxtransient basedonthenuclearpowerandflowfromLOFTRAN.0882L:614.1.6"2 Finally,theTHINCcodeisusedtocalculate theDNBRduringthetransient basedontheheatfluxfromFACTRANand'lowfromLOFTRAN.TheDNBRtransients presented represent theminimumofthetypicalorthimblecell.This'accident isanalyzedwiththeImprovedThermalDesignProcedure asdescribed inWCAP-8567 (Reference 6).Plantcharacteristics andinitialconditions arediscussed inSection14.InitialOperatinConditions Initialreactorpower,pressure, andRCStemperature areassumedtobeattheirnominalvalues~Uncertainties ininitialconditions areincludedinthelimitDNBRasdescribed inWCAP-8567.

Reactivit Coefficients Aconservatively largeabsolutevalueoftheDoppler-only powercoefficient isused.Thisservestomaximizepowerlevelwhileitisdecreasing afterreactor'trip.

Thetotalintegrated Dopplerreactivity (powerdefect)between(4and100io'ower isassumedtobe0.0166k,consistent withFigure14-2.Themostpositivevalueofthemoderator temperature coefficient

(+5pcm/~F)isassumed,sincethisresultsinthemaximumcorepowerduringtheinitialpartofthetransient, whentheminimumdeparture fromnucleateboilingratioisreached.FlowCoastdown Theflowcoastdown analysisisbasedonamomentumbalancearoundeachreactorcoolantloopandacrossthereactorcore.Thismomentumbalance0882L:614.1.6-3

iscombinedwiththecontinuity

equation, apumpmomentumbalanceandthepumpcharacteristics andisbasedonhighestimates ofsystempressurelosses.Nosingleactivatefailureintheplantsystemsandequipment whicharenecessary tomitigatetheeffectsoftheaccidentwilladversely affecttheconsequences oftheaccidentduringthetransient mostlyasaresultofthechangeoffuelgapconductance.

Aconservatively evaluated overallheattransferwasusedintheanalysis.

ResultsReactorcoolantflowcoastdown curvesareshowninFigure14.1.6-1.Figures14.1.6-1and14.1.6-3showthenuclearflux,theaveragechannelheatflux,andthehotchannelheatfluxresponseforthetwo-pumploss.Figure14.1.6-2showstheDNBratioasafunctionoftimeforthiscase.TheminimumWRB-1DNBratioisreached3.0secondsafterinitiation oftheincident.

Figures14.1.6-4through14.1.6-6showthetransient forlossofonepumpwithbothloopsoperating andFigure14.1.6-7showstheDNBratioasafunctionoftimeforthiscase.TheminimumDNBratiooccurs3.5secondsafterinitiation ofthetransient.

Conclusions SinceDNBdoesnotoccurinanyloss-of-coolant flowincident, thereisnocladdingdamageandnoreleaseoffissionproductsintothereactorcoolant.Therefore, oncethefaultiscorrected, theplantcanbereturnedtoserviceinthenormalmanner.Theabsenceoffuelfailureswould,ofcourse,beverifiedbyanalysisofreactorcoolantsamples.0882L:614.1.6-4

LockedRotorAccidentAhypothetical transient analysisisperformed forthepostulated instantaneous seizureofareactorcoolantpumprotor.Flowthroughthereactorcool'antsystemisrapidlyreduced,leadingtoareactortriponalow-flowsignal.Following thetrip,heatstoredinthefuelrodscontinues topassintothecorecoolant,causingthecoolanttoheatupandexpand.Atthesametime,heattransfertotheshellsideofthesteamgenerator isreduced,firstbecausethereducedflowresultsinadecreased tubesidefilmcoefficient andthenbecausethereactorcoolantinthetubescoolsdownwhiletheshellsidetemperature increses(turbinesteamflowisreducedtozerouponplanttrip).Therapidexpansion ofthecoolantinthereactorcore,combinedwiththereducedheattransferinthesteamgenerator, causesaninsurgeintothepressurizer andapressureincreasethroughout thereactorcoolantsystem.Theinsurgeintothepressurizer compresses thesteamvolume,actuatestheautomatic spraysystem,opensthepower-operated reliefvalves,andopensthepressurizer safetyvalves,inthatsequence.

Thetwopower-operated reliefvalvesaredesignedforreliableoperation andwouldbeexpectedtofunctionproperlyduringtheaccident.

However,forconservatism, theirpressure-reducing effect,aswellasthepressure-reducing effectofthespray,isnotincludedintheanalysis.

MethodofAnalsisTwodigitalcomputercodesareusedtoanalyzethistransient.

TheLOFTRANcodeisusedtocalculate theresulting loopcoreandflowtransients following thepumpseizure,thetimeofreactortripbasedonloopflowtransients, andthenuclearpowerfollowing reactortrip,andtodetermine peakpressure.

Thethermalbehaviorofthefuellocatedatthecorehotspotisinvestigated usingtheFACTRANcode,whichusesthecoreflowandnuclearpowercalculated byLOFTRAN.TheFACTRANcodeincludesafilmboilingheattransfercoefficient.

0882L:614.1.6-5 I

Onecaseisanalyzedwithbothloopsoperating andonelockedrotor.Atthebeginning ofthepostulated lockedrotoraccident(i.e.,atthetimetheshaftinoneofthereactorcoolantpumpsisassumedtoseize),theplantisassumedtobeinoperation underthemostadversesteady"state operating conditions withrespecttothepressure, i.e.,maximumsteady-state powerlevel,maximumsteady-state pressure(2280psia),andmaximumsteady-state coolantaveragetemperature.

Thelockedrotoreventisnotanalyzedwithaconsequential lossofoffsitepower.AttheR.E.Ginnaplant,thegenerator breakerswill.notopenuntiloneminuteafterthelossofoffsitepower.Thus,powerwillbemaintained totheintactreactorcoolantpumpthroughout thelimitingportionofthetransient.

Thisiswithinthefirst10secondswhenthepeakcladtemperature occurs.Forthepeakpressureevaluation, theinitialpressureisconservatively estimated as30psiabovenominalpressure(2250psia)toallowforerrorsinthepressurizer pressuremeasurement andcontrolchannels.

Thisisdonetoobtainthehighestpossibleriseinthecoolantpressureduringthetransient.

Toobtainthemaximumpressureintheprimaryside,conservatively highlooppressuredropsareaddedtothecalculated pressurizer pressure.

Thepressure" responseshowninFigure'N14.1.6-9istheresponseatthepointinthereactorcoolantsystemhavingthemaximumpressure.

Evaluation ofthePressureTransient

-Afterpumpseizure,theneutronfluxisrapidlyreducedbycontrolrodinsertion effect..Rodmotionisassumedtobeginonesecondaftertheflowintheaffectedloopreaches87%ofnominalflow.Nocreditistakenforthepressure-reducing effectofthepressurizer reliefvalves,pressurizer spray,steamdump,orcontrolled feedwater flowafterplanttrip.Althoughtheseoperations areexpectedtooccurandwouldresultinalowerpeakpressure, anadditional degreeofconservatism isprovidedbyignoringtheireffect.0882L:614.1.6-6 N

Thepressurizer safetyvalvesarefullopenat2575psia,andtheirtotalcapacityforsteamreliefis20ft/s.3Evaluation of

DearturefromNucleateBoilinintheCoreDurintheAccident-Forthisaccident,

departure fromthenucleateboilingisassumedtooccurinthecore,andtherefore, anevaluation oftheconsequence withrespecttofuelrodthermaltransients isperformed.

Resultsobtainedfromanalysisofthishotspotcondition represent theupper1-imitwithrespecttocladtemperature andzirconium-water reaction.

Intheevaluation, therodpoweratthehotspotisassumedtobethreetimestheaveragerodpower(Fg3)attheinitialcorepowerlevel.FilmBoil,inCoefficient

-Thefilmboilingcoefficient iscalculated intheFACTRANcodeusingtheBishop-Sandberg-Tong filmboilingcorrelation.

Thefluidproperties areevaluated atfilmtemperture, whichistheaveragebetweenthewallandbulktemperatures.

Theprogramcalculates thefilmcoefficient ateverytimestep,basedonthe.actualheattransferconditions atthetime.Theneutronflux,systempressure, bulkdensity,andmassflowrateasafunctionoftimeareusedasprograminput.Forthisanalysis, theinitialvaluesofthepressureandthebulkdensityareusedthroughout thetransient, sincetheyarethemostconservative withrespecttocladtemperature response.

Forconservation, departure fromnucleateboilingwasassumedtostartatthebeginning oftheaccident.

FuelCladGaCoefficient

-Themagnitude andthetimedependence oftheheattransfercoefficient betweenfuelandclad(gapcoefficient) haveapronounced influence onthethermalresults.Thelargerthevalueofthegapcoefficient, themoreheatistransferred betweenthepelletandtheclad.Basedoninvestigations oftheeffectofthegapcoefficient onthemaximumcladtemperature duringthetransient, thegapcoefficient isassumedtoincreasefromastedy-state valueconsistent withaninitialfueltemperature to10,000Btuperhour-square 0882L:614.1.6-7 feet-Fattheinitiation ofthetransient.

Thus,thelargeamountofenergystoredinthefuelbecauseofthesmallinitialvalueisreleasedtothecladattheinitiation ofthetransient.

Zirconium-Steam Reaction-Thezirconium-steam reactioncanbecomesignificant aboveacladtemperature of1,800'F.TheBaker-Just parabolic rateequationshownbelowisusedtodefinetherateofthezirconium-steam reaction:

26d~w333w30~45000dt1.986Twhere:w=amountreacted(mg/cm)2t=time(seconds)

T=temperature

('F).Thereactionheatis1,510cal/gm.ResultsFigure14.1.6"18andFigure14.1.6-9showthenuclearpower,coreflow,andloopflowtransients andFigure14.1.6-18showsthepressuri.zer pressuretransients.

Theheatfluxandcladtemperature transients aregiveninFigure14.1.6-10.Theresultsofthesecalculations aresummarized inTable14.1.6-2.ThesequenceofeventsisshowninTables14.1.6-1and14.1.6-3.

0882L:614.1.6-8 Conclusions Sincethepeakreactorcoolantsystempressure(2836psia)reachedduringanyofthetransients islessthan120Kofdesignpressuretheintegrity oftheprimarycoolantsystemisnotendangered.

Thisvaluecanbeconsidered anupperlimit,sincetheassumptions usedinthe.analysisareconservative.

Sincethepeakcladsurfacetemperature (2176'F)calculated forthehotspotduringthemoreseveretransient remainsconsiderably lessthan2,700'Fandtheamountofzirconium-water reactionissmall,thecoreremainsinplaceandintactwithnoconsequential lossofcorecoolingcapability.

0882L:614.1.6-9 TABLE14.1.6"1TIMESEQUENCEOFEVENTSFORLOSSOFREACTORCOOLANTFLOWCaseEventTimeofEachEventSecondsa.Partiallossofreactorcoolantflow(twoloopsoperating, onepumpcoastingdown)Coastdown beginsLowflowreactortrip1.27Rodsbegintodrop2.27MinimumDNBRoccurs3.5b.CompletelossofforcedreactorcoolantflowBothoperating pumpslosepowerandbegincoastingdownReactorcoolantpumpundervoltage trippointreachedRodsbegintodrop1.5MinimumDNBRoccurs3.00882L:614.1.6-10

TABLE14.1.6-2SUMMARYOFLIMITINGRESULTSFORLOCKEDROTORACCIDENTMaximumReactorCoolantSystemPressure(psia)2836MaximumCladdingTemperature

('F)CoreHotSpot2176Zr-H20ReactionatCoreHotSpot(5byweight).99350882L:614.1.6-11

TABLE14.1.6-3TIMESEQUENCEOFEVENTSFORLOCKEDROTORINCIDENTEventTimeofEachEventSecondsRotorononepumplocksLowflowtrippointreached.09Rodsbegintodrop1.09MaximumRCSpressureoccurs3.20Maximumcladtemperature occurs3.410882L:614.1~6-12 Figure14.1.6-1~irmaFull'psofFlowt.2CCO.<0000t:2OCOt.COCO'.20000X.CCCCCIA'VIva,2CCCC-CPCPCL(SEC)CICIChCt14.1.6-13 Figure14.1.6-2GinnaFullLossofFlowZICOOr23'.CIA~~~QCOO.ZCCO.C.tOO:0'.>cQcQ,COO':oP(CAR~Agel,gCCC(SEU14.1.6-14

Figure14.1.6-3Gr)pQu1IQ55!.2NQ1.0000.60000p,~ro.,e C.'nanncl,IdQCC1;CCCC..CCOQC4%~QQQQ%CCgus>A~-(14.1.6-15

Figure14.1.6-4GinnaPartialLossofF1ow~libIII~IIIhOIIA0II50000IO00200000.0IIlM~'IT000wl50000I'IF0000+nnn2'NOOfr00CI7lltE<SEO14.1.6-16 Figure14.1.6-5GinnaPartialLossof.Flow~EllIl8F009)111I1P82".pp8!pppZQOOpt".00p~oqpn~1PP~)IlI1I1I~~!".Qpp0900060000~nppp29000f0.0CICI.(Hf.5Eil14.1.6-17 0

Figure14.1.6-6pinnaPar-ialLssofF'ow,I\A(~:".000~?CCQQ+A>8"a.c.'g~he~aaelSCQQQF000020000~)IlhQ~~:0000OhQQPSpoonipppp20900IIIIggrChcahne(00ICPCT(meCie~>14.1.6-18

Figure14.1.6-7CI~Q>(t<El5cCl14.1.6-19

Figure14.1.6-8GinnaLockedRotor2.0000~7CQn!SQQQ'QQ3I.00007SQQQ~cSQQQQ250000.0~2tin0~nQQQ.80000X'60000ClF0000Xn200000.0CI~5OQtO%sf5='4.1.6-20 Figure14.1.6-9GonnaLockedRotor20<<'..,0000>"OCSCOOQ25CQOn-25000IfbllQIl<p,g($(=bLCCP2800.0O2500.02<00.Q2200.02000.0!800,0CICICIC7CICICICICIaDTtv((SEC)14.1.6-21

Figure14.1.6-10 GinnaLockedRotor!,@VVVl.00001ISO0004X'ICLS00004F0000tI20000jI0.022hh20000tII'.1S0,0~I:S00.l'tIl2S0.9ViI1000004838OTlg'SEC)14.1.6-22 14.1.8LossofExternalElectrical LoadTheplantisdesignedtoaccepta50Klossofelectrical loadwhileoperating atfullpoweroracompletelossofloadwhileoperating below50%powerwithoutactuating areactortrip.Theautomatic steambypasssystemwith405steamdumpcapacitytothecondenser isabletoaccommodate thisloadrejection byreducingthetransient imposeduponthereactorcoolantsystem.Thereactorpowerisreducedtothenewequilibrium powerlevelatarateconsistentwiththecapability oftherodcontrolsystem.Shouldthereactorsufferacompletelossofloadfromfullpower,thereactorprotection systemwouldautomatically actuateareactortrip.Themostlikelysourceofacompletelossofloadonthenuclearsteamsupplysystemisatripoftheturbine-generator.

Inthiscase,thereisadirectreactortripsignalderivedfromeithertheturbineautostopoilpressureoraclosureoftheturbinestopvalves,providedthereactorisoperating above505power.Reactortemperature andpressuredonotincreasesignificantly ifthesteambypasssystemandpressurizer pressurecontrolsystemarefunctioning properly.

However,theplantbehaviorisevaluated foracompletelossofloadfromfullpowerwithoutadirectreactortrip,primarily toshowtheadequacyofthepressurerelieving devicesandalsotoshowthatnocoredamageoccurs.Thereactorcoolantsystemandsteamsystempressurerelieving capacities aredesignedtoensurethesafetyoftheplantwithoutrequiring theautomatic rodcontrol,pressurizer pressurecontrol,and/orsteambypasscontrolsystems.MethodofAnalsisThetotallossofloadtransients areanalyzedbyemploying thedetaileddigitalcomputerprogramLOFTRAN.Theprogramsimulates theneutron0882L:614.1.8-1

kinetics, reactorcoolantsystem,pressurizer, pressurizer reliefandsafetyvalves,pressurizer spray,steamgenerator, andsteamgenerator safetyvalves.Theprogramcomputespertinent plantvariables, including temperatures, pressures, andpowerlevel.Inthisanalysis, thebehavioroftheunitisevaluated foracompletelossofsteamloadfrom100Koffullpowerwithoutdirectreactortrip,primarily toshowtheadequacyofthepressure-relieving devicesandalsotodemon'strate coreprotection margins.ThisaccidentisanalyzedwiththeImprovedThermalDesignProcedures inWCAP-8567, Reference 6(maintext).Plantcharacteristics andinitialconditions arediscussed inSection14.1.Initial0eratinConditions

-Theinitialreactorpowerandreactorcoolantsystemtemperatures areassumedattheirnominalvalues.Uncertainties ininitialconditions areincludedinthelimitDNBRasdescribed inWCAP-8567.

Moderator andDolerCoefficients ofReactivit

-Theloss-of-load accidentisanalyzedwithbothmaximumandminimumreactivity feedback.

Themaximumfeedbackcasesassumealargenegativemoderator temperature coefficient andthemostnegativeDopplerpowercoefficient.

Theminimumfeedbackcasesassumepositivemoderator temperature coefficient

(+5pcm/F)andtheleastnegativeDopplercoefficient.

ReactorControl-Fromthestandpoint ofthemaximumpressures

attained, itisconservative toassumethatthereactorisinmanualcontrol.SteamRelease'-Nocreditistakenfortheoperation ofthesteamdumpsystemorsteamgenerator power-operated reliefvalves.Thesteam0882L:614.1.8-2

generator pressurerisestothesafetyvalvesetpoint, wheresteamreleasethroughsafetyvalveslimitssecondary steampressureatthesetpointvalue.Pressurizer SraandPower-0cratedReliefValves-Twocases,for,bothmaximumandminimumfeedback, areanalyzed.

a.Fullcreditistakenfortheeffectofpressurizer sprayandpower-operated reliefvalves,inreducingorlimitingthecoolantpressure.

b.Nocreditistakenfortheeffectofpressurizer sprayandpower-operated reliefvalvesinreducingorlimitingthecoolantpressure.

Safetyvalvesareoperable.

Feedwater Flow-Mainfeedwater flowtothesteamgenerators isassumedtobelostatthetimeoflossofexternalelectrical load.Reactortripisactuatedbythefirstreactorprotection systemtripsetpointreached,withnocredittakenforthedirectreactortriponturbinetrip.ResultsThetransient responses foratotallossofloadfromfull"power operation areshownforfourcases-twocasesforminimumreactivity feedbackandtwocasesformaximumreactivity feedbackillustrated inFigures14.1~8-1through14.1.8-12.

Figures14.1.8-1through14.1.8-3showthetransient responses forthetotalloss-of-steam loadwithminimumreactivity

feedback, assumingfullcreditforthepressurizer sprayandpressurizer power-operated reliefvalves.Nocreditistakenforthesteamdump.0882L:614.1.8-3

Thereactoristripped.bythehighpressurizer pressuresignal.Theminimumdeparture fromnucleateboilingratioiswellabovethelimitvalue.Thepressurizer safetyvalvesarenotactuated.

Figures14.1.8-4through14.1.8-6showtheresponseforthetotalloss-of-steam loadwithalargenegativemoderator temperature coefficient.

Astemperature increases nuclearpowerdecreases duetonegativereactivity feedback.

Powerthenstabilizes atalowerpowerleveluntilthelowsteamgenerator leveltripsetpointisreached.TheDNBRincreases throughout thetransient andneverdropsbelowitsinitialvalue.Pressurizer reliefvalvesandsteamgenerator safetyvalvespreventoverpressurization inprimaryandsecondary systems,respectively.

Thepressurizer safetyvalvesarenotactuatedforthiscase.Following thelowsteamgenerator waterlevelreactortrip,auxiliary feedwater wouldbeusedtoremovedecayheatwiththeresultslessseverethanthosepresented inSection14.1.9oftheFSAR,LossofNormalFeedwater Flow.IThetotallossofloadaccidentwasalsostudiedassumingtheplanttobeinitially operating at100%offullpower,withnocredittakenforthepressurizer spray,pressurizer power"operated reliefvalves,orsteamdump.Thereactoristrippedonthehighpressurizer pressuresigna'1.Figures14.1.8-7through14.1.8-9showtheminimumfeedbacktransients.

Theneutronfluxincreases slightlyuntilthereactoristripped.Thedeparture fromnucleateboilingratioincreases throughout thetransient.

Inthiscase,thepressurizer safetyvalveisactuated.

0882L:614.1.8-4 Figures14.1.8-10 through14.1.8-12 showthetransients withmaximumfeedbackandallotherassumptions beingthesameasthoseinFigures14.1.8-7through14.1.8-9.

Again,thedeparture fromnucleateboilingratioincreases throughout thetransient, andthepressurizer safetyvalvesareactuated.

Thecalculated sequenceofeventsforthesefourcasesisshowninTable14.1.8-1.

Conclusions Resultsoftheanalysesshowthattheplantdesignissuchthatatotallossofexternalelectrical loadwithoutadirectorimmediate reactortrippresentsnohazardtotheintegrity ofthereactorcoolantsystemorthemainsteamsystem.Pressure-relieving devicesincorporated inthetwosystemsareadequatetolimitthemaximumpressures withinthedesignlimits.Theintegrity ofthecoreismaintained byoperation ofthereactorprotection system;i.e.,thedeparture fromnucleateboilingratioismaintained abovethelimitvalue.0882L:614.1.8-5 TABLE14.1.8"1TIMESEQUENCEOFEVENTSFORLOSSOFEXTERNALELECTRICAL LOADCaseEventTimeofEachEventSecondsa.Withpressurizer control(minimumfeedback)

Lossofelectrical loadHighpressurizer pressurereactortrippointreached12.6Rodbeginstodrop14.6Peakpressurizer pressureoccurs16.0Minimumdeparture fromnucleateboilingratiooccursb.Withpressurizer control(maximumfeedback)

Lossofelectrical loadPeakpressurizer pressureoccurs13.00882L:6P14.1.8"6 CaseTABLE14.1.8-1(continued)

EventTimeofEachEventSecondsLowsteamgenerator levelreactortrippoint81.3Rodsbegintodrop83.3Minimumdeparture fromnucleateboilingratiooccursc.Withoutpressurizer control(minimumfeedback)

Lossofelectrical loadHighpressurizer pressurereactor5.4trippointreachedRodsbegintodrop7.4Peakpressureoccurs9.0Initiation ofreleasefromS/Gsafetyvalves13.0Minimumdeparture fromnucleateboilingratiooccurs*DNBRdoesnotdecreasebelowitsinitialvalue.0882L:614.1.8-7 CaseTABLE14.1.8-1(continued)

EventTimeofEachEventSecondsd.Withoutpressurizer control(maximumfeedback)

Lossofelectrical

'loadHighpressurizer pressurereactortrippointreached05.3Rodsbegintodrop7.3Peakpressureoccurs8.0Initiation ofreleasefromS/GsafetyvalvesMinimumdeparture fromnucleateboilingratiooccursDNBRdoesnotdecreasebelowitsinitialvalue.0882L:614.1.8-8

Figure14.1.8-1GinnaLossofLoadMinimum;edback.vithAuomatcPressure"ontro1I.2000XICCCCZCIeooooCI50000o,>COCO'.2CCCO1IIIITI0.0I.00003~000~I3.0000I2.5000+~'.0000I.500020QQCICICICI5cclCID14.1.8-9 Figure14.1.8-2GinnaLossofLoadMinimumFeedbackwithAutomatic PressureControl660.005l0.QO600.00550.00d5.6O.QO570.00~'cl0QO550F005<O.QO600.QO560.QO~560.005iO.00lIII520.00g500.QOClCIClClAJClCl,lÃK(5EC)aClClClOClCl14.1.8-10 Figure14.1.8-3GinnaLossofLoadMinimumFeedbackwithAutomatic PressureControI25DQ.Q2400.0)400hj2300.3~22CQ.Q4!e2!00.0tIZCOO.Qts400I',800.0t!700.0!I!OCO.'0400.COtaCO.QO-4J700.00I500.005n4.!CO.CO400.CDCICICIClCICICICICICIrlHf(SKC)

Figure14.1.8-4GinnaLossofLoadt<aximumFeedbackwithAutomatic PressureControl~AI>>~~CCQXOltlWlv>>v>>JhIvvgl>>vh2&i>>%.S000i000>>52.0050,2!040Z.0000l.SD4Q.).2000.oCIflN(lSCQ14.1.8-1g

Figure14.1.8-5GinnaLossofLoadMaximumFeedbackwithAutomatic PressureControl0\40AA[WAQCt'AA~V~;~n.n+'ih1IVtvfbi'(sgA~VevTLHs14.1.8-13 0

Figure14.1.8-6GinnaLossofLoadMaximumFeedback,vith Automatic.

PressureControl2>C".VAAllC'22"Q.V5glAAQ)AAAA~VVV~4AQA~AAA~~I+I~V1AV~~AAQQ4AA'lA~~V3illAAAV~lllAAQAAAA4V4AAAQ~V4AAAQTIRE(44)14.1.8-14

Figure14.1.8-7GinnaIossofLoad'linimumFeedback'AithouPressureContro12000'000.SOOOO.SOQQOyI,iCCCOCJ.20000RII+5.0000i,!CQO~.00003.!000g3.aaaaIIlIIj2.!000nnQQIIl.!0002CCQCICI'SEC)14.1.8-15

Figure14.1.8-8GinnaLossofLoadMin'.numFeedbac!<

M:.hou.?ressureCcntro1520."0otp~004200.00<<III55ppp4l580.00tI570.004I5eO.CP~I550.00jIIII5~0.00SOO.OO5SO.PO<<560.00tCl540.00520.00500.00ClCIClClClClClC7ClClCCCCYlHE<5EC)14.1.8-16 Figure14.1.8-9GinnaLossofLoadMinimumFeedbackMithoutPressureControl2500.02400.0t2<00.04,2300.02200.9j42IhQhv2000.04!4CO.Q~F800.0I)00Q!000,00I300.00ycI400nP4.VE700."0CC500.CDinCO."0III400.COCIC5AyC)CIaC7CICI<SEC)14.1.8-17 II Figure14.1.8-10 GinnaLossoiLoadMaximumFeedbackWithoutPressureControl!.2000!.0000.80000+I)OCCO"jt).40000I20000TI!.CCCOt!000~t.0000~13.!COOiI3.COCOj.III2.!OOC+II)%IQCIIIt0tlA~vV!.2000CIO1!20)14;1.3-18

Figure141'8-11GinnaLossofLoadMaximumFeedbackMithoutPressure'.Control 820.00I8'.0.20~800.00I580.00~j1580.00~I5(llhQI560.00I550.005gOQIIt+I800"0580.00y580.0045io.00520.00~tQQhQCICI~qCCIC7t58C)14.1.8-19 Figure14.1.8-12 GinnaLossofLoadMaximumFeedbackMithoutPressureControl2600.5ciaoaj2300,522CO.Of2'!00.052CCO.5!400.54'L800.0!7005'"CO0I400.50fSCO.CO4baal'I~I5,np<<pTICCI500.50+I400.5400.00oCtC4I620)14.1.8-20 e

14.1.10Excessive HeatRemovalDuetoFeedwater Temperature DecreaseThereduction infeedwater enthalpyisanothermeansofincreasing corepowerabovefullpower.Suchincreases areattenuated bythethermalcapacityinthesecondary

'plantandintheReactorCoolantSystem.Theoverpower-overtemperature protection (nuclearoverpower andhTtrips)preventsanypowerincreasewhichcouldleadtoaDNBRlessthanlimit'(value.Anextremeexampleofexcessheatremovalbythefeedwater systemisthe'ransient associated withtheaccidental openingofthefee'dwater bypassvalvewhichdivertsflowaroundthelowpressurefeedwater heaters.Thefunctionofthisvalveistomaintainnetpositivesuctionheadonthemainfeedwater pumpintheeventthattheheaterdrainpumpflowislost,e.g.,duringalargeloaddecrease.

Intheeventofanaccidental openingthereisasuddenreduction ininletfeedwater temperature, tothesteamgenerators.

Theincreased subcooling

'willcreateagreaterloaddemandontheprimarysystemwhichcanleadtoareactortrip.Thethree-element feedwater controlsystemoperatestoregulatethefeedwater flowandmaintainawaterlevelapproximately constantinthesteamgenerator.

Actionofthethree-way boilercontrolunderemergency condition hasnobearingonsafetysinceemergency feedwater isinjecteddownstream ofthecontrolvalves.However,thefeedwater controlvalvesareusedforfeedwater lineisolation.

Anysafetyinjection signalwillredunantly isolatethe.feedwater linesbya)ventingthesupplyairtoallfeedwater control.valves,causingvalvestoclose;andbyb)trippingoffthemainfeedwater pumps,including closureofthefeedwater discharge valves.TheweteffectontheRCSduetoareduction infeedwater enthalpyissimilartotheeffectofincreasing secondary steamflow,i.e,thereactorwillreachanewequilibriumcondition atapowerlevel,corresponding tothenewsteamgenerator hT.0882L:614.1.10-1 MethodofAnalsisThisaccidentisanalyzedusingtheLOFTRANcode.Thecodesimulates theneutronkinetics, reactorcoolantsystem,pressurizer, pressurizer reliefandsafetyvalves,pressurizer spray,steamgenerator, steamgenerator safetyvalves,andfeedwater system.Thecodecomputes.

pertinent plantvariables, including temperatures, pressures, andpowerlevel.Thistransient isanalyzedbyreducingthefeedwater enthalpybytheamountcorresponding tothelossIofonefeedwater heater.Twocaseshavebeenanalyzedtodemonstrate theplantbehaviorintheeventofasuddenfeedwater temperature reduction resulting fromaccidental openingofthebypassvalve.'II1.Reactorcontrolinmanualwithmaximummoderator reactivity feedback.

2.Reactorcontrolinautomatic withmaximummoderator reactivity feedback.

Thereactivity insertion rateatnoloadfollowing anexcessive feedwater accidenthasalsobeencalculated withthefollowing assumptions:

1.Astepincreaseinfeedwater flowtoonesteamgenerator from0tothenominalfullloadvalueforonesteamgenerator.

2.Themostnegativereactivity moderator coefficient atendoflife.0882L:614.1.10-2 I.f 3.Aconstantfeedwater temperature of70F.4.Neglectoftheheatcapacityofthereactorcoolantsystemandsteamgenerator shellthickmetal.5.Neglectoftheenergystoredinthefluidoftheunaffected secondsteamgenerator..

Continuous additionofcoldfeedwater afterareactortripisprevented sincethereduction ofreactorcoolantsystemtemperature,

pressure, andpressurizer levelwillleadtotheactuation ofsafetyinjection onlowJpressurizer pressure.

Thesafetyinjection signalwilltripthelmainfeedwater pumpsandclosethefeedwater pumpdischarge valvesaswellasclosethemainfeedwater controlvalves.ThisaccidentisanalyzedwiththeImprovedThermalDesignProcedure asdescribed in'eference 12.Plantcharacteristics andinitialconditions arediscussed inSection14.1.Initialreactorpower,pressure, andRCStemperatures areassumedtobeattheirnominalvalues.Uncertainties

~ininitialconditions areincludedinthelimitDNBRasdescribed inReference 6.ResultsFigures14.1.10-1 through14.4.10-3 illustrate thetransient withthereactorintheautomatic controlmode.Duetotheactionofthecontrolrodsandmoderator

feedback, thenuclearpowerincreases whiletemperature andpressuredecreaseuntilasteady-,state condition isreached.Areduction indeparture fromnucleateboilingratioisexperienced, butthedeparture fromnucleateboilingratioremainsabovethelimitvalue.0882L:614.1.10-3

,J Figures14.1.10-4 through14.1.10-6 illustrate thetransient whenthereactorisassumedtobeinthemanualcontrolmode.Again,thecorepowerincreases duetothedecreaseincoolantaveragetemperature.

Thedeparture fromnucleateboilingratiodecreases butremainsabovethelimitvalue.Thefeedwater enthalpydecreaseincidentissimilartoanexcessive loadincreaseandisanoverpower transient forwhichthefueltemperatures rise.Whenareactortripdoesnotoccur,theplantreachesanewequilibrium condition atahigherpowerlevelcorresponding totheincreaseinsteamflow.Atzeropower,fortheexcessive feedwater flowtoonesteamgenerator, g~7&themaximumreactivity insertion ratewascalculated tobe4.5xl0hk/second.

Thisislessthanthemaximumreactivity insertion ratelanalyzedinSection14.1.1,Uncontrolled RCCAWithdrawal fromaaSubcritical Condition.

Itshouldbenotedthatiftheaccidentoccurswiththeplantjustcriticalatnoload,thereactorwillbetrippedby.thepowerrangefluxleveltrip(lowsetting)setatapproximately 25K.AsshowninSection'4.1.1, theDNBremainsabovethelimitvalue.-iss"LS+Conclusions Ithasbeendemonstrated that,forafeedwater enthalpydecreaseatfullpower,minimumDNBRdoesnotfallbelowthelimitvalue..Atzeropower,theresultsarelesslimitingthanthosepresented inSection14.1.1.0882L:614.1.10-4

Figure14.1.10-1 1.2000GinnaFeedwater EnthalpyDecreaseAutomatic RodControlt.oooo~.80000cK.60000o.40000CI.200000.01.2000l.0000XIDCICIILJCC.80000.60OOO40000I.200000.0CIC7C)C)AJC)C)VlCICICICIC7C)C)CII/ICDAJ1TIVEtSEC)14.1.10-5 Figure14.1.10-2 GinnaFeedwater EnthalpyDecreasekAutomatic RodControl2500.02400.02300.0cc2200'2IQQ.02000.0g1400.0.1800.017.00.0I000.00900.00800.004JK700.00CD)eso.soa.500.00tQQ.QQCDCDCDCDCDCDCDCDCDCDCDCDCDCDCDAJTIME<SEC)14.1.10-6 Figure14.1.10-3 GinnaFeedwater fnthalpyDecreaseAutomatic RodControl610.Oo600.00550.00c)580.00570.00560.Oot550.00540.003.00002.75002.50002.2500c..00001.7500I.5000CDCDCDCDAJCDCDI/>CDCDCD(s<<)CDCDCDCDCDCDCDAJ14.1.10-7

Figure14.1.10-4

!.2CCOGinnaFeedwater Entha1pyDecreaseManua1RodControIi.OCCOt.SCCOOZ'D.60000IcD.40000t..20000-,II0.0:.20001.0000ZCD.80000CD.60000T.40000.200000.0CDCDCDCDPgCDCDCD~CDCDV1T'iCDCDCDCDHE(SEC)CDCDCDCD14.1.10-8

Figure14.1.10-5 GinnaFeedwater EnthalpyDecrease'I2500.0ManualRodControl2400.02300.0+III2200.02100.02000.0i800.OCL)800.0-$700.0!"OC.Qo8CQ.Co800.00LJ7glhhoJSCO.Oo7I.500.Qo4QO.OOCDCDCDCDCICDfVIflCD,CD'D(,SEC)CDCD14.1.1G-9

Figurel4.1.10-6 ffGinnaFeedwater EnthaloyOecreasemanualRodControl520.006I0.00~600.00TI5000~a580.00-570.00t.I560.001550.00Sao.CC-.3.00002.75002.5000T2.2500T2.00001.7500TI.S000~IC)CICIC)CCCCIClC)C)CDCDTIME(SEC)14.1.10-10 14.1.11Excessive LoadIncrease'ncident Anexcessive loadincreaseincidentisdefinedasarapidincreaseinsteamgenerator steamflowthatcausesapowermismatchbetweenthereactorcorepowerandthesteamgenerato'r loaddemand.Thereactorcontrolsystemisdesignedtoaccommodate a10%steploadincreaseand/ora5~perminuteramploadincrease(withoutareactortrip)intherangeof155to10(Cfullpower.Anyloadingrateinexcessofthesevaluesmaycauseareactortripactuatedbythereactorprotection system.Iftheloadincreaseexceedsthecapability ofthereactorcontrolsystem,thetransient isterminated intimetopreventDNBRlessthanthelimitingvaluebyacombination ofthenuclearoverpower tripandtheoverpower-overtemperature hTtrips.Anexcessive loadincreaseincidentcouldresultfromeithe'ranadministrative violation, suchassteambypasscontrolorturbinespeedcontrol.Forexcessive loadingbytheoperatororbysystemdemand,theturbineloadlimiterkeepsmaximumturbineloadbelow100Kratedload.Duringpoweroperation, steambypasstothecondenser iscontrolled byreactorcoolantcondition signals,i.e.,abnormally highreactorcoolanttemperature indicates aneedforsteambypass.Asinglecontroller malfunction doesnotcausesteambypass;aninterlock isprovidedwhichblocksthecontrolsignaltothevalvesunlessalargeturbineloaddecreaseoraturbinetriphasoccurred.

MethodofAnalsisThisaccidentisanalyzedusingtheLOFTRANcode.Thecodesimulates theneutronkinetics, reactorcoolantsystem,pressurizer, pressurizer reliefandsafetyvalves,pressurizer spray,steamgenerator, steam0882L:614.1.11-1

generator safetyvalves,andfeedwater system.Thecodecomputespertinent plantvariables, including temperatures, pressures, andpowerlevel.Fourcasesareanalyzedtodemonstrate theplantbehaviorfollowing a105step-load increasefromratedload.Thesecasesareasfollows:1.Reactorcontrolinmanualwithminimummoderator reactivity feedback.

2.Reactorcontrolinmanualwithmaximummoderator reactivity feedback.

3.Reactorcontrolinautomatic withminimummoderator reactivity feedback.

4.Reactorcontrolinautomatic withmaximummoderator reactivity feedback.

Fortheminimummoderator feedbackcases,thecorehasa5.0pcm/'Fmoderator temperature coefficient ofreactivity and,therefore, theleastinherenttransient capability.

Forthemaximummoderator feedbackcases,themoderator temperature coefficient ofreactivity hasitsmostnegativevalue.Thisresultsinthelargestamountofreactivity feedbackduetochangesincoolanttemperature.

Aconservative limitontheturbinevalveopeningisassumed,andallcasesarestudiedwithoutcreditbeingtakenforpressurizer heaters.ThisaccidentisanalyzedwiththeImprovedThermalDesignProcedure asdescribed inReference 12.Plantcharacteristics andinitialconditions areasdiscussed inSection14.1.Initialreactorpower,pressure,.

and0882L:614.1.11-2 RCStemperatures areassumedtobeattheirnominalvalues.Uncertainties ininitialconditions areincludedinthelimitDNBRasdescribed inReference 6.ResultsFigures14.1.11"1through14.1.11-6illustrate thetransient withthereactorinthemanualcontrolmode.Fortheminimumfeedbackcase,thepositiveMTCcausesthenucl.earpowertodecreasewithtemperature andpressureunti1areactortriponlowpressurizer pressureoccurs.Thisresultsinadeparture fromnucleateboilingratiothatincreases aboveitsinitialvalue.For'hemaximumfeedback, manuallycontrolled case,thereisanincreaseinreactorpowerduetothemoderator feedback.

Areduction indeparture fromnucleateboilingratioisexperienced, butthedeparture fromnucleateboilingratioremainsabovethelimitvalue.Figures14.1.11-7through14.1.11-12illustrate thetransient whenthereactorisassumedtobeintheautomatic controlmode.Boththeminimumandmaximumfeedbackcasesshowthatcorepowerincreases, therebyreducingtherateofdecreaseincoolantaveragetemperature andpressurizer pressure.

Forboththeminimumandmaximumfeedbackcases,theminimumdeparture fromnucleateboilingratioremainsabovethelimitvalue.Thecalculated sequenceofeventsisshowninTable14.1.11-1.

Theexcessive loadincreaseincidentisanoverpower transient forwhichthefueltemperatures rise.'Whenareactortripdoesnotoccur,theplantreachesanewequilibrium condition atahigherpowerlevel'Icorresponding totheincreaseinsteamflow.Conclusion Ithasbeendemonstrated that,foranexcessive loadincrease, theminimumdeparture fromnucleateboilingratioduringthetransient willnotbebelowthelimitvalue.0882L:614.1.11-3 TABLE14.1.11"1 TIMESEQUENCEOFEVENTSFOREXCESSIVE LOADINCREASEINCIDENTCaseEventTimeofEachEventSecondsa.Manualreactorcontrol(minimumfeedback) 10%steploadincreaseLowerpressurizer pressuretripreachedRodsbegintofallintocore213.9215.9'.Manualreactorcontrol(maximumfeedback) 10%steploadincreaseEquilibriumconditions reached(approximate timesonly)50.0c.Automatic reactor10/ostepload'ncrease control(minimumfeedback)

Equilibrium conditions reached(approximate timesonly)35.0d.Automatic reactorcontrol(maximumfeedback) 105steploadincreaseEquilibrium conditions reached(approximate timesonly)60.00882L:614.1.11-4 Figure14.1.11-1 GinnaExcessLoadIncreaseMinimumFeedbackwithoutRodControll.2000z1.0000CIICI.80000II.60000c..l0000C4141.20000I.ZOOOl.0000XCIXII41.80000~60000.10000.200000.0CICIVlCICICI1IlCICICICImTIHK(SEC)14.1.11-5 Figure14.1.11-2 GinnaExcessLoadIncreaseMinimumFeedbackwithoutRodContro1Z800.0Zioo,o2200.02200.0VlCC02100.02000.01800.0!800.01700.0$000.00800.00~m800.00700.00'4J600.00.oo.ao400.00300.00200.00.ao.aaClCICSCIC5C)<SC)sabirizsr)CIClCIAJC7CI14.1.11-6 o

Figure14.1.11-3GinnaExcessLoadIncreaseMinimumFeedbackwithoutRodControl620.00600.00580.00560.00SIO.00S20.00500.005.0001Ki.5000l.00003.50003.00002.0000!.SOOO1.2000CDCICICDCICICDCD,CICITAHE(SEC)CDCDAJCDCDCDhJCICICIm

Figure14.1.11-4 GinnaExcessLoadIncreaseMaximumFeedbackwithoutRodControl1.2000,Il.0000gI.60000<(I.60000~a~<0000+CCI0.0I.0000XC.60000cIXCI~60000.10000II'i.200000.0CSCIys33s383000O000CICDClInC)C)AlAJit%TIME<SEC)14.1.11-8 Figure14.1.11-5 GinnaExcessLoadIncreaseMaximumFeedbackMithoutRodControlCiUV.Vjlao.0C2300.0I+2200.02!00.02COO.01eco.0:eco.o4'.Iao.otI+I!000.00Sao.aoEJcoo.ao>CO.CO5sao.co5sao.oo+00.00OOCICICICIOCICICICInCIONIOCICICIOCIINIIIlCIOCICINInHE(SEC)14.1.11-9 Figure14.1.11-6 GinnaExcessLoadIncreaseMaximumFeedbackwithoutRodControl620.00610.00,II600.00~I4I550.00~I560.00~c570.00y560.00550.005lO.003.00002.7500Z.2500Z.00001~7500l.5000I:9888OCI3CIvtCIOCIOVlCIOAJCIAJOgCIOCICImCIrtlCICIO3PPE(SEC)14.1.11-10 Figure14.1.11-7 GinnaExcessLoadIncreaseMinimumFeedbackwithAutomatic RodControlIJh0h<<xI000960hnh4l~a.600h09CC.C00.hgnnss00.0~gh0h~hhh0Pn0h00.60000CCC0hh'Iao3CIVlgInhlIIIplTIRE(SKC)CICICI eFigure14.1.11-8 GinnaExcessLoadIncreaseMinimumFeedbackwithAutomatic RodControlI2500.02100.0vs2300.0g22nn0<<00.0i2000"CCgl00.0l800.0l700.9l000.00600.00LI800.00100.00600.00II.500.00400.'0CICICICICICIAJ88C5CImCICICICITlHElSEC)14.1.11-12 Figure14.1.11-9 GinnaExcessLoadIncreaseMinimumFeedbackwithAutomatic RodControl6ZQ.00610.00600.00550.00580.""c57"..56".".55L.:5l0.:::3.0000!.oooojZ.5000t.350041.50001I:QSBCIgOOOOI.OOO00OOV1OoooOOosooooooTlHK(5EC)14.1.11-13

Figure14.1.11-10 GinnaExcessLoadIncreaseMaximumFeedbackwithAutomatic RodControlz1.0000.80000cc.60000o.<00004g.200000.0f.2000t.0000.80000a.60000.I0000I0.0OOInCIOOIno383OOOT?HE(SKC)14.1.11-14 Figure14.1.11-11 GinnaExcessLoadIncreaseMaximumFeedbackwithAutomatic RodControl2100.02200.0C4m2200.0III2100.02000.01900.0l800.01700.0!000.00800.00le800.00'4J~oo.oocecoo.ooo.800.00Ioo,00ClCICICICIVlCICIOOOCIOhJCIClCICICImCIOOOOOmTfHKtSEC)

Figure14.1.11-12 GinnaExcessLoadIncreaseMaximumFeedbackwithAutomatic RodContro1620.00610.00600.00580.004J560.00c570.00S60.00550.00540.003,%002.150020002.25002.0000K.750000000000>0CI8CICIovs8CIVtIJTIRE(SEC)00"0O0gO0mm14.1.11-16 14.2.5RuptureofaSteamPipeAruptureofasteampipeisassumedtoincludeanyaccidentwhichresultsinanuncontrolled steamreleasefromasteamgenerator.

Thereleasecanoccurduetoabreakinapipelineorduetoavalvemalfunction.

Thesteamreleaseresultsinaninitialincreaseinsteamflowwhichdecreases duringtheaccidentasthesteampressurefalls.TheenergyremovalfromtheReactorCoolantSystemcausesareduction ofcoolanttemperature andpressure.

Withanegativemoderator temperature coefficient, thecooldownresultsinareduction ofcoreshutdownmargin.Ifthemostreactivecontrolrodisassumedtobestuckinitsfullywithdrawn

position, thereisapossibility thatthecorewillbecomecriticalandreturntopowerevenwiththeremaining controlrodsinserted.

Areturntopowerfollowing asteampiperuptureisapotential problemonlybecauseofthehighhotchannelfactorswhichmayexistwhenthemostreactiverodisassumedstuckinitsfullywithdrawn position.

Assumingthemostpessimistic combination ofcircumstances

.whichcouldleadtopowergeneration following asteamlinebreak,thecoreisutimately shutdownbytheboricacidintheSafetyInjec~ion System.Theanalysisofasteampiperuptureisperformed todemonstrate thatwithastuckrodandminimally engineered safetyfeatures, thecoreremainsinplaceandessentially intactsoasnottoimpaireffective coolingofthecore.AlthoughDNBandpossiblecladperforation (nocladmeltingorzirconium-water reaction) following asteampiperupturearenotnecessarily unacceptable, thefollowing

analysis, infact,showsthatnoDNBoccursforanyrupture,assumingthatthemostreactiverodisstuckinitsfullywithdrawn position.

H0882L:614.2.5-1 Thefollowing systems'providethenecessary protection againstasteampiperupture:1.SafetyInjection Systemactuation on:1a.Twooutofthreepressurizer lowpressuresignals.b.Twooutofthreelowpressuresignalsinanysteamline.c.Twooutofthreehighcontainment pressuresignals.2.Theoverpower trips(neutronfluxandhT)andthereactortripoccurring uponactuation oftheSafetyInjection System.3.Redundant isolation ofthemainfeedwater lines.Sustained highfeedwater flowwouldcauseadditional cooldown; thus,inadditiontothenormalcontrolactionwhichwi11closethemainfeedwater valves,anysafetyinjection signalwillrapidlycloseallfeedwater controlvalves,tripthemainfeedwater pumps,andclosethefeedwater pumpdischarge valves.4.Tripofthefastactingsteamlineisolation valves(designed tocloseinlessthanfivesecondswithnoflow)on:a.Oneoutofthetwosteamflowsignalsinthatsteamlineincoincidence withanysafetyinjection signal.(Dualsetpoints areprovided, withthelowersetpointusedincoincidence withtwoout,offourindications oflowreactorcoolantaveragetemperature.)

0882L:614.2.5-2 b.Twooutofthreehighcontainment pressuresignals.Eachsteamlinehasafastclosingisolation valveandacheckvalve.Thesefourvalvespreventblowdownofmorethanonesteamgenerator foranybreaklocationevenifonevalvefailstoclose.Forexample,forabreakupstreamoftheisolation valveinoneline,closureofeitherthecheckvalveinthatlineortheisolation valveintheotherlinewillpreventblowdownoftheothersteamgenerator.

Steamflowismeasuredbymonitoring dynamicheadinnozzlesinsidethesteampipes.Thenozzles(16-in.IDversusapipediameterof28-in.ID)arelocatedinsidethecontainment nearthesteamgenerator andalsoservetolimitthemaximumsteamflowforanybreakfurtherdownstream.

Inparticular, thenozzleslimittheflowforallbreaksoutsidethecontainment.

MethodofAnalsisTheanalysisofthesteampiperupturehasbeenperformed todetermine:

1.Thecoreheatfluxandreactorcoolantsystemtemperature andpressureresulting fromthecooldownfollowing thesteamlinebreak.TheLOFTRANcodehasbeenused.2.Thethermalandhydraulic behaviorofthecorefollowing asteamlinebreak.Adetailedthermalandhydraulic digital-computer code,THINChasbeenusedtodetermine ifDNBoccursforthecoreconditions computedin(1)above.0882L:614.2.5"3 Thefollowing assumptions weremade:l.A0.018shutdownreactivity fromtherodsatnoloadconditions with2loopsinoperation.

Thisistheend-of-life designvalueincl'uding designmarginswiththemostreactiverodstuckinitsfullywithdrawn position.

Operation ofrodclustercontrolassemblybanksduringcoreburnupisrestricted insuchawaythatadditionofpositivereactivity inasecondary systemsteamreleaseaccidentwillnotleadtoamoreadversecondition thanthecaseanalyzed.

A0.0245shutdownreactivity isassumedforcaseswhereoneloopisinservice.2.Thenegativemoderator temperature coefficient corresponding totheendoflifecorewithallbutthemostreactiverodinserted.

Thevariation ofthecoefficient withtemperature andpressurehasbeenincluded.

Thekversustemperature at1000psiacorresponding tothenegativemoderator temperature coefficient usedisshowninFigure14.2.5-1.

Incomputing thepowergeneration following asteamlinebreak,thelocalreactivity feedbackfromthehighneutronfluxintheregionofthecorenearthestuckcontrolrodhasbeenincludedintheoverallreactivebalance.Thelocalreactivity feedbackiscomposedofDopplerreactivity fromthehighfueltemperatures nearthestuckcontrolrodandmoderator feedbackfromthehighwaterenthalpynearthestuckrod.Forthecasesanalyzedwheresteamgeneration occursinthehighflux.regionsofthecore,theeffectofvoidformation onthereactivity hasbeenincluded.

Theeffectofpowergeneration inthecoreonoverallreactivity ispresented inFigure14.2.5-2.

Thecurveassumesendoflifecoreconditions withallrodsinexceptthemostreactiverodwhichisassumedstuckinitsfullywithdrawn position.

3.Minimumsafetyinjection capability corresponding totwooutofthreesafetyinjection pumpsinoperation.

Twothousand(2000)ppmboronisassumedinthesafetyinjection system.Thetimedelaysrequiredtosweepthelowconcentration boricacidfromthesafety0882L:614.2.5-4 injection piping.priortothedeliveryoftheboronhavebeenincludedintheanalysis.

Twentythousand(20,000)ppmboronisassumedinthecaseswithoneloopinservice.4.Powerpeakingfactorscorresponding toonestuckRCCAandnonuniform coreinletcoolanttemperatures aredethrmined atendofcorelife.Thecoldestcoreinlettemperatures areassumedtooccurinthesectorwiththestuckrod.Thepowerpeakingfactorsaccountfortheeffectofthelocalvoidintheregionofthestuckcontrolrodassemblyduringreturntopowerphasefollowing thesteamline break.Thisvoidinconjunction withthelargenegativemoderator coefficient partially offsetstheeffectofthestuckassembly.

Thepowerpeakingfactorsdependuponthecorepower,temperature,

pressure, and,flow, and,thus,aredifferent foreachcasestudied.5.Threecombinations ofbreaksizesandinitialplantconditions havebeenconsidered indetermining thecorepowerandreactorcoolantsystemtransient.

a.Completeseverance ofapipeinsidethecontainment attheoutletofthesteamgenerator atinitialno-loadconditions withoutsidepoweravailable andtwoloopsinservice.The'quivalent breakareais4.6ft.b.Case(a)abovewithlossofoutsidepowersimultaneous withthesteambreak.c.Abreakequivalent tosteamreleasethroughonesteamgenerator safetyvalvewithoutsidepoweravailable andtwoloopsinservice.d.Case(a)abovewithonlyoneloopinservice.e.Case(c)abovewithonlyoneloopinservice.0882L:614.2.5-5 Theseverance ofapipedownstream ofthesteamflowmeasuring nozzleisnotanalyzed.

The.equivalent breakarea(1.4ft)2islessthanthatofcase(a)andwouldresultinalessseverecooldown.

Thus,thisbreakisboundedbycases(a)and(b).Thecasesaboveassumeinitialhotshutdownconditions withtherodsinserted(exceptforonestuckrod)attimezero.Shouldthereactorbe'ustcriticaloroperating atpoweratthetimeofasteamlinebreakthereactorwillbetrippedbythenormaloverpower protection systemwhenthepowerlevelreachesatrippoint.Following atripatpowerthereactorcoolantsystemcontainsmorestoredenergythanatno"load,theaveragecoolanttemperature ishigherthanatno-loadandthereisappreciable energystoredinthefuel.Thus,theadditional storedenergyisremovedviathecooldowncausedbythesteamlinebreakbeforethenoloadconditions ofreactorcoolantsystemtemperature andshutdownmarginassumedintheanalysesarereached.Aftertheadditional storedenergyhasbeenremoved,thecooldownandreactivity insertions proceedinthesamemannerasintheanalyseswhichassumeno-loadconditions attimezero.ResultsTheresultspresented areaconservative indication oftheeventswhichwouldoccurassumingasteamlinerupture.Theworstcaseassumesthat.allofthefollowing occursimultaneously.

1.Minimumshutdownreactivity marginequalto1.80%(2loopsinservice).

Minimumshutdownreactivity marginequalto2.45.o(1loopinservice).

2.Themostnegativemoderator temperature coefficent fortheroddedcoreatendoflife.3.Therodhavingthemostreactivity stuckinitsfullywithdrawn position.

0882L:614.2.5-6

4.Onesafetyinjection pumpfailstofunctionasdesigned.

CorePowerandReactorCoolantSstemTransient Figures14.2.5-3through14.2.5-7showthereactorcoolantsystemtransient andcoreheatfluxfollowing asteampiperupture(complete severence ofapipe)attheexitofasteamgenerator atinitialno-loadconditions withtwoloopsinoperation.

Thebreakassumedisthelargestbreakwhichcanoccuranywhereeitherupstreamordownstream oftheisolation valves.Offsitepowerisassumedavailable suchthatfullreactorcoolantflowexists.Thetransient shownassumestherodsinsertedattime0(withonerodstuckinitsfullywithdrawn position) andsteamreleasefrombothsteamgenerators.

Shouldthecorebecriticalatnearzeropowerwhentheruptureoccurs,theinitiation ofsafetyinjection bylowsteamlinepressurewilltripthereactor.Steamreleasefromatleastonesteamgenerator willbeprevented byeitherthecheckvalveorbyautomatic tripofthefastactingisolation valveinthesteamlinebythehighsteamflowsignalincoincidence withthesafetyinjection signal.Evenwiththefailureofonevalve,releaseislimitedtonomorethansevensecondsforonesteamgenerator whi,lethesecondgenerator blows'down.(Thesteamlineisolation valvesaredesignedtobefullyclosedinlessthanfivesecondswithnoflowthroughthem.Withthehighflowexistingduringasteamlinerupture,thevalveswillcloseconsiderably faster.)Thecorebecomescriticalwiththerodsinserted(withthe.designshutdownassumingonestuckrod)at14.5seconds.Boronsolutionat2,000ppmentersthereactorcoolantsystemfromthesafetyinjection system(initiated automatically bythelowsteamlinepressure) at41.0secondswhichincludesthedelayrequiredtoclearthesafetyinjection systemlinesoflowconcentration boricacid.Nocredithasbeentakenforthe2,000ppmboronwhichentersthereactorcoolantsystempriorthe2,000ppmboricacid.Thepeakcoreheatfluxis31%of1520Mwt.0882L:614.2.5-7

Figures14.2.5-,8 through14.2.5-12 showtheresponses forcaseaassumingalossofoutsidepowerattime0whichthenresultsinareactorcoolantsystemflowcoastdown.

Thesafetyinjection systemdelaytimeincludesthe-.timerequiredtostartasafetyinjection pump-'nthediesel.Onlyonedieselisassumedtostart.Creditistakenforonlythesafetyinjection flowenteringthecold-leglines,sincetheflowtothehotlegflowpathsarevalvedshut.Thepeakpoweris20Kofnominal.Figures14.2.5-13 through14.2.5-17 showtheresponses forafailedsteamgenerator safetyvalvewithtwoloopsinoperation.

Criticality occursat220seconds.Boronentersthecoreduetoalowpressurizer pressuresafetyinjection signalat200seconds.Figures14'.5-18through14.2.5-22 showthetransient foradoubleendedruptureassumingoneloopinservice.Theloophavingtheaffectedsteamgenerator isassumedtobeinoperation.

Thesequenceofeventsissimilartothecasewithbothloopsinoperation.

Thecorebecomescriticalat22.0seconds.Boronsolutionat20,000ppmentersthecoreat37.0seconds.Thepeakcoreheatfluxis27Kof1520NWth.Thetransient forafailedsafetyvalvewithoneloopinserviceispresented inFigures14.2.5-23 through14.2.5-26.

Boronsolutionat20,000ppmentersthecoreat160seconds.Criticality doesnotoccur.Thesequenceofeventsforeachcaseispresented inTable14.2.5-1.

Conclusion ADNBanalysiswasperformed foreachcase.ItwasfoundthatallcaseshaveaminimumDNBRgreaterthanthelimitvalue.TheanalysishasshownthatthecriteriastatedinSection14.2.5aresatisfied'lthough DNBandpossiblecladdingperforation following asteampiperupturearenotnecessarily unacceptable andnotprecluded bythecriteria, theaboveanalysis, infact,showsthattheDNBdesignbasisismet'asstatedinSection4.0882L:614.2.5-8 Casea.TABLE14.2.5-1TIMESEQUENCEOFEVENTSFORSTEAMLINE RUPTURETimeofEventEventSteamline rupturesPressurizer emptiesCriticality attainedBoronenterscore8.514.541.0Steamline rupture;offsitepowerlostPressurizer emptiesCriticality attainedBoronenterscore9.519.053.0C.SafetyvalvefailsopenPressurizer emptiesLowpressurizer pressureSIsetpointreachedBoronenterscore97100200Criticality occurs2200882L:614.2.5"9 TABLE14.2.5-1(Continued)

TIMESEQUENCEOFEVENTSFORSTEAMLINE RUPTURETimeofEventCaseEventSteamline rupturesPressurizer emptiesCriticality attainedBoronenterscoreSafetyvalvefailsopenPressurizer emptiesLowpressurizer pressureSIsetpointreachedBoronenterscore9.022.037.00.093.099.01600882L:614.2.5-10 FIGURE14.2.5-1GINNASTEAMLINE RUPTURE1.041.031.021.011.00.99.98200300400COREAVERAGETEMPERATURE,

'F14.2.5-11 Figure14.2.5-2GINNASTEAMLINERUPTURE2.01.81.61.41.21.0Cl40)C)SClCLGLO.8lg$Ol.4.20.1.2.3.4.5-fraction ofpower14.2.5-12 FIGURE14.2.5-3.50000GINNASTEAMLINE RUPTURE4.6ftBreakwithPower2LoopsinService2.ipoooCIILJC?:.30000.20000w.~pppp0.0.50000c.LpopoX.30000.20000C.toooo0.0COCIC)C)CtCDCCI1COCl.~W71HE(SEC)14.2,5-13 FIGURE14.2.5-4GINNASTEAMLINE RUPTURE4.6ftBreakwithPower22LoopsinService2000,01750.01500.01250.01000.00CI750.00500..001000.00500.00-800.00700.00600.00500.00CI400.00300.00I/I200.00100.000.0CICICICICIC)CI'IAdCICICICICICICICICICITIME15E6)14.2.5-14 FIGURE14.2.5-5GINNASTEAMLINE RUPTURE4.6ftBreakwithPower550.00500.00450.00400.00C)C)o35000IO300.00in$pgfFacetted250.00200.00600.00550.00500.00450.00lalChhhhCMM~vv350.00300.00250.00200.00C)CIC)ClCOCOCICOCVCOCOCOCDmC)CDCtC)COTlHE<SIC)14.2.S-1S FIGURE14.2.5-6GINNASTEAMLINE RUPTURE4.6ftBreakwithPower2cCCCI2.5000.25002.0000l.7500l,5000l.2500l.0000.75000.50000.25000oClC7Vl2.5C002.25002.0000l.7500c~cays.<<vipl.2500l.0000z.75000.50000ID.25000IntactFaulted0.0000CICOOOC)CVCICIC)COIDC>IDC)COC)TIME(52C)14.2.5-16 FIGURE14.2.5-7GINNASTEANLINE RUPTURE4.6ftBreakwithPower22500.02000.01000.00ZLJ0.0100Oc.0<000.'0<500.0500.00F00.00XcL300.00CD200.00CD\JIOO.000.0CDCDCDCDCDAJCDCDCDCDTIME<SEC>CDCDCDCD14.2.5-17 FIGURE14.2.5-86INNASTEAMLINERUPTURE4.6ftBreakw/oPower-2LoopsinService2.l00005.30000LJ~~.20000CI4.100000.0T'AI%15KO.NN00~.moo3.amo~100000.0CITINE(%C)14.2.5-18 FIGURE14.2.5-9GINNASTEAMLINE RUPTURE4.6ftw/oPower-2LoopsinService22000.01750.0~gnOO.OM50.0~l000.00750.00flylSEC)1000.00500.00NO.008700.00NO.00gg400.00WA4g300,00IA200.00Oef00.000.0CS8AlTlME(SEC)C5CI14.2.5-19 FIGURE14.2.5-10 GINNASTEAMLIHE RUPTURE4.6ftw/oPower-2LoopsinService2550.00Intact't<50.00F00.00Im.~l300.00Faulted200.00C7CI(AC)CIClCl550.00<50.005F00.00c350.00300.00~5TtHE(SEC)ClCIClClClCl14.2.5-20 FIGURE14.5.2-11GINNASTEAMLINE RUPTURE4.6ftw/0Power-2LoopsinService22.50002.25002.ON).7500).5e)51.2500),oooo.75000CI.50000.25N~0.0C)CIC)TTIC(5KC)2.50002.25002.0000).7500).5000I).2500).0000z.75000Intact0.0aulted88-.8gI3y3S%THE(SKC)14.2.5-21 FIGURE14.2.5-12 GINNASTEAMLINE RUPTURE4.6ftBreakw/outPower2~2500.02000.01000.00K0.0-1000'.0<000.0<500.0500.00F00.00Xa.300.00CD200.00100.000.0CDCDCDCDCDCDCDAlTlME<SEC>CDCDCDC)CDCD14.2.5-22 FIGURE14.2,5-13 GINNASTEAMLINE RUPTURE-FAILEDSAFETYVALVE.40000c%C).30000lLJ.20000~100000.0.50000.iooooxIDXla.30000I4J.2OOOO'.100000.0C)CICIC)C)ClC)CICImTIHE(SEC)14.2.5-23 FIGURE14.2.5-14GINNASTEAMLINE RUPTURE-FAILEDSAFETYVALVE2LoopsinService2OOO.01750.015OO.01250.01000.00500.001000.00900.00w~&00.00700.00600.00laJo500.00o)cgF00.00K300.00200.00100.000.0ooooooooooAJooiiHE(SEC)oCt14.2.5-24 FIGUREj4,2.5GINNASTEAMLINE RUPTURE-FAILEDSAFETYVALVE2LoopsinService550.00500.00o50.00I100.00F~<YEQo350.00I~o30000250.00200.00550.00500.00i50.00IOO.00350.00I300.00Z50.00oooCIoCIAloCIoTIHEtSEC)oooCIoCICICI14.2.5-25 FIGURE14,2.516GINNASTEANLIHE RUPTURE-FAILEDSAFETYVALVE2zcppx2.00001.7500'1.50001.25001.0000.75000.50000Cl.250000.0CICIClCICICIoCICDClClCDmClClCIc5CICIClClCICDIDCl250002000'1000.000.01000.0<000.0%500.0CI'ICIClCDCDClCDCDClClClClClClCITIHE(SEC)14.2.5-26 FIGURE14.2.5-QGINNASTEAMLINERUPTURE-FAILEOSAFETYVALVE500.00400.00Xa-300.00CD200.00CDLIIOO.000.0CDCD'DCDAJCDCDCDmCDClCDCDCDCDCDCDC)C)CDTIME(SEC)142.5-2>

Figure14.2.5-18GinnaSteamline Rupture4.6ftBreakwithPower-OneLoopinService2K.l0000Clcl.34)000ILJK4R.20000.100000.0.IOOOOXCl.300M.20000.100000.0ClClCIAlTlHE<SEC)8ClClCl14.2.5-28

Figure14.2.5-19GinnaSteamline Rupture4.6ftBreakwithPower-OneLoopinService22000.01750.01500.01250.01000.00750.001000.00SOO.00800.00700.00600.00Q500.00cc100.00300.00200.00100.000.0NCIEDClEDClClAlT1HE{SEC)ClCIClClCIClClCl14.2.5-29 Figure14.2.5-20GinnaSteamline Rupture4.6ftBreakwithPower-OneLoopinService2550.00500.00i50.00400.009cD350.00I300.00250.00200.00550.00500.00i50.00ccIoo.00CD350.00I300.00250.00200.00CDCDCDCDCDCD"CDmCDCDCDillCTIME(5E()14.2.5-30 Figure14.2.5-21GinnaSteamline Rupture4.6ftBreakwithPower-OneLoopinService22.50002.25002.0000I.nOOl.5000t.moI.OOOO.75000A.500000.0AA8CCCOClCDCDIll2.2500K2.0000I.7500CDlCClaI.5000I.2500I.0000x.75000.50000o.2SO0OIntactFaulted0.0OOOOO000000OCIClClCIOOCDCIOCCJCClTIME(SEC)CDOCDCDCOOCRO.CNOO14.2.5-3l FIGURE14.2.5-22GINNASTEAMLINE RUPTURE4.6ftBreakwithPower2OneLoopinService14.2.5-32 Figure14.2.5-23GinnaSteamline RuptureIFailedSafetyYalve-OneLoopinServiceC2000.0l750.0III'5000l250.0~lON.00750.00l000.00900.00600.00700.00600.00500.00CIcg100.00g3N.00200.00l00.000.0CICICICIAJCICICICIm7lHE(SEC)CICICICI14.2.5-33 Figure14.2.5-24GinnaSteamline RuptureFailedSafeyValve-OneLoopinService550.00500.00150.00IAcc100.00350.00I300.00250.00200.00550.00500.00't450.00F00.00350.009I300.00250.00200.00CDCDCICDC)CMCDCITtMKlSEC)CDC)IDCDCDqCDID14.2.5-34

FIGURE14.2.5-25GINNASTEAMLINE RUPTUREFailedSafetyValve-OneLoopinServiceLSD625002.00001.7500gL5000It,2500<.0000.75000h.saoo0.0.a8814.2.5-35 4lnCDQVlKlCDV)PlC+'Cm(X~INVlCDPlIIChCDCX7Im0OaI/lIDO'CD 14.2.6RuptureofaControlRodMechanism Housing-RCCA EjectionInorderforthisaccidenttooccur,arupture'ofthecontrolrodmechanism housingmustbepostulated creatingafullsystempressuredifferential actingonthedriveshaft.Theresultant corethermalpowerexcursion islimitedbytheDopplerreactivity effectsoftheincreased fuel'emperature andterminated byreactortripactuatedbyhighnuclearpowersignals.Afailureofacontrolrodmechanism housingsufficient to,allowacontrolrodtoberapidlyejectedfromthecoreisnotconsidered credibleforthefollowing reasons:1.Eachcontrolroddrivemechanism housingiscompletely assembled andshop-tested at4100psi.2.Themechanism housingareindividually hydrotested to3105psigas'heyareinstalled onthereactorvesselheadtotheheadadapters, andcheckedduringthehydrotest ofthecompleted reactorcoolantsystem.3.'tresslevelsinthe.mechanism arenotaffectedbysystemtransients atpower,orbythethermalmovementofthecoolantloops.Momentsinducedbythedesignearthquake canbeacceptedwithintheallowable primaryworkingstressrangespecified bytheASMECode,SectionIII,forClassAcomponents.

4.Thelatchmechanism housingandrodtravelhousingareeachasinglelengthofforgedtype-304stainless steel.Thismaterialexhibitsexcellent notchtoughness atalltemperatures thatwillbeencountered.

Thejointsbetweenthelatchmechanism housingandheadadapter,andbetweenthelatchmechanism housingandrodtravelhousing,arethreadedjointsreinforced bycanopytyperodwelds.0882L:614.2.6-1 NuclearOesinEvenifaruptureofaRCCAdrivemechanism housingispostulated, theoperation ofaplantutilizing chemicalshimissuchthattheseverityofanejectedRCCAisinherently limited.Ingeneral,thereactorisoperatedwiththeRCCA'sinsertedonlyfarenoughtopermitloadfol-low.Reactivity cha'ngescausedbycoredepletion andxenontransients arecompensated byboronchanges.Further,thelocationandgroupingofcontrolRCCAbanksareselectedduringthenucleardesigntolessentheseverityofaRCCAejectionaccident.

Therefore, shouldaRCCAbeejectedfromitsnormalpositionduringfullpoweroperation, onlyaminorreactivity excursion, atworst,couldbeexpectedtooccur.However,itmaybeoccasionally desirable tooperatewithlargerthannormalinsertions.

Forthisreason,arodinsertion limitisdefinedas,afunctionofpowerlevel.Operation withtheRCCA'sabovethislimitguarantees adequateshutdowncapability andacceptable powerdistribution.

ThepositionofallRCCA'siscontinuously indicated inthecontrolroom.AnalarmwilloccurifabankofRCCA'sapproaches itsinsertion limitorifoneRCCAdeviatesfromitsbank.Operating instructions requireborationatlowlevelalarmandemergency borationatthelow-lowalarm.ReactorProtection Thereactorprotection intheeventofarodejectionaccidenthasbeendescribed inReference 4.Theprotection forthisaccidentisprovidedbyhighneutronfluxtrip(highandlowsetting).

Theseprotection functions aredescribed indetailinSection7.2oftheFSAR.0882L:614.2.6-2

EffectsonAdjacentHousinsDisregarding theremotepossibility oftheoccurrence ofaRCCAmech-anismhousingfailure,investigations haveshownthatfailureofahousingduetoeitherlongitudinal orcircumferential crackingwouldnotcausedamagetoadjacenthousings.

However,evenifdamageispostu-lated,itwouldnotbeexpectedtoleadtoamoreseveretransient, sinceRCCA'sareinsertedinthecoreinsymmetric

patterns, andcontrolrodsimmediately adjacenttotheworstejectedrodsarenotinthecorewhenthereactoriscritical.

Damagetoanadjacenthousingcould,atworst,causethatRCCAnottofallonreceiving atripsignal;however,thisisalreadytakenintoaccountintheanalysisbyassumingastuckrodisadjacenttotheejectedrod.LimitinCriteriaThiseventisclassified asanANSCondition IVincident.

Duetotheextremely lowprobability ofaRCCAejectionaccident, somefueldamagecouldbeconsidered anacceptable consequence.

Comprehensive studies,bothofthethreshold offuelfailureandofthethreshold orsignificant conversion ofthefuelthermalenergytomechanical energy,havebeencarriedoutaspartoftheSPERTprojectbythe'IdahoNuclearCorporation.

Extensive testsofU02zirconium cladfuelrodsrepresentative ofthoseinpressurized waterreactortypecoreshavedemonstrated failurethresholds intherangeof240to257cal/gm.However,otherrodsofasightlydifferent desi.gnhaveexhibited failuresaslowas225cal/gm.Theseresultsdiffersignificantly fromtheTREATresults,whichindicated afailurethreshold of280cal/gm.Limitedresultshaveindicated thatthisthreshold decreases byabout10%withfuelburnup.Thecladfailuremechanism appearstobemeltingforzeroburnuprodsandbrittle0882L:614.2.6-3

fractureforirradiated rods.Alsoimportant istheconversion ratioofthermaltomechanical energy.Thisratiobecomesmarginally detectable above300cal/gmforunirradiated rodsand200cal/gmforirradiated rods;catastrophic failure(largefueldispersal, largepressurerise)evenforirradiated rodsdidnotoccurbelow300cal/gm.Inviewoftheaboveexperimental results,criteriaareappliedtoensurethatthereislittleornopossibility offueldispersal inthecoolant,grosslatticedistortion, orsevereshockwaves.Thesecriteriaare:a.Averagefuelpelletenthalpyatthehotspotbelow200cal/gm.b.Averagecladtemperature atthehotspotbelowthetemperature atwhichcladembrittlement maybeexpected(2700'F).

c.Peakreactorcoolantpressurelessthanthatwhichcouldcausestressestoexceedthefaultedcondition stresslimits.d.Fuelmeltingwillbelimitedtolessthantenpercentofthefuelvolumeatthehotspoteveniftheaveragefuelpelletenthalpyisbelowthelimitsofcriterion (a)above.AnalsisofEffectsandConseuencesMethodofAnalysisThecalculation oftheRCCAejectiontransient isperformed intwostages,firstanaveragecorechannelcalculation andthenahotregioncalculation.

Theaveragecorecalculation isperformed usingspatialneutronkineticsmethodstodetermine theaveragepowergeneration withtimeincluding thevarious.totalcorefeedbackeffects,i.e.,Doppler0882L:614.2.6"4

reactivity andmoderator reactivity.

Enthalpyandtemperature tran-sientsinthehotspotarethendetermined bymultiplying theaveragecoreenergygeneration bythehotchannelfactorandperforming afuelrodtransient heattransfercalculation.

Thepowerdistribution calcu-latedwithoutfeedbackispes'simistically assumedtopersistthroughout thetransient.

Adetaileddiscussion ofthemethodofanalysiscanbefou'ndinReference 4.AverageCoreAnalsisThespatialkineticscomputercode,TWINKLE(Reference 4),isusedfortheaveragecoretransient analysis.

Thiscodesolvesthetwogroupneutrondiffusion theorykineticequationinone,twoorthreespatialdimensions (rectangular coordinates) forsixdelayedneutrongroupsandupto2000spatialpoints.Thecomputercodeincludesadetailedmultiregion, transient fuel-clad-coolant heattransfermodelforcalcu-lationofpointwise Dopplerandmoderator feedbackeffects.Inthisanalysis, thecodeisusedasaonedimensional axialkineticscode,sinceitallowsamorerealistic representation ofthespatialeffec~tsofaxialmoderator feedbackandRCCAmovement.

However,sincetheradialdimension ismissing,itisstillnecessary toemployverycon-servative methods(described inthefollowing) ofcalculating theejectedrodworthandhotchannelfactor.Furtherdescription ofTWINKLEappearsinSection14.HotSotAnalsisInthehotspotanalysis, theinitialheatfluxisequaltothenominaltimesthedesignhotchannelfactor.Duringthetransient, theheatfluxhotchannelfactorislinearlyincreased tothetransient valuein0.1second,thetimeforfullejectionoftherod.Therefore, the0882L:614.2.6-5 assumption ismadethatthehotspotsbeforeandafterejectionarecoincident.

Thisisveryconservative, sincethepeakafterejectionwilloccurinoradjacenttotheassemblywiththeejectedrod,andpriortoejectionthepowerinthisregionwillnecessarily bedepressed.

IIThehotspotanalysisisperformed usingthedetailedfuel-andcladdingtransient heattransfercomputercode,FACTRAN(Reference 2).Thiscomputercodecalculates thetransient temperature distribution inacrosssectionofametalcladU02fuelrod,andtheheatfluxatthesurfaceoftherod,usingasinputthenuclearpowerversustimeandthelocalcoolantconditions.

Thezirconium-water reactionisexplicitly represented, andallmaterialproperties arerepresented asfunctions oftemperature.

Aconservative pelletradialpowerdistribution isusedwithinthefuelrod.FACTRANusestheDittus-Boelter orJens-Lottes correlation todetermine thefilmheattransferbeforeDNB,andtheBishop-Sandburg-Tong correla-tiontodetermine thefilmboilingcoefficient afterDNB.TheBSTcorrelation isconservatively usedassumingzerobulkfluidquality.TheDNBratioisnotcalculated, insteadthecodeisforcedintoDNBbyspecifying aconservative DNBheatflux.Thegapheattran'sfer coefficient canbecalculated bythecode;however,itisadjustedinordertoforcethefullpowersteady-state temperature distribution toagreewiththefuelheattransferdesigncodes.Furtherdescription ofFACTRANappearsinSection14.SstemOverressureAnalsisBecausesafetylimitsforfueldamagespecified earlierarenotexceeded, thereislittlelikelihood offueldispersal intothecool-ant.Thepressuresurgemaythereforebecalculated onthebasisofconventional heattransferfromthefuelandpromptheatgeneration inthecoolant.0882L:614.2.6-6 Thepressuresurgeiscalculated byfirstperforming thefuelheattransfercalculation todetermine theaverageandhotspotheatfluxversustime.Usingtheseheatfluxdata,aTHINC(Section4)calcula-tionisconducted todetermine thevolumesurge.Fina'lly, thevolumesurgeissimulated inaplanttransient computercode.Thiscodecalcu-latesthepressuretransient takingintoaccountfluidtransport inthereactor:coolant systemandheattransfertothesteamgenerators.

Nocreditistakenforthepossiblepressurereduction causedbytheassumedfailureofthecontrolrodpressurehousing.Calculation ofBasicParameters Inputparameters fortheanalysisareconservatively selectedonthebasisofvaluescalculated forthistypeofcore.Themoreimportant parameters arediscussed below.Table14.2.6-1presentstheparameters usedinthisanalysis.

EjectedRodWorthsandHotChannelFactorsThevaluesforejectedrodworthsandhotchannelfactorsarecalculated usingeitherthree-dimensional staticmethodsorbyasynthesis methodemploying one-dimensional andtwo-dimensional calculations.

Standardnucleardesigncodesareusedintheanalysis.

Nocreditistakenforthefluxflattening effectsofreactivity feedback.

Thecalculation isperformed forthemaximumallowedbankinsertion atagivenpowerlevel,asdetermined bytherodinsertion limits.Adversexenondistributions areconsidered inthecalculation.

Appropriate marginsareaddedtotheejectedrodworthandhotchannelfactorstoaccountforanycalculational uncertainties, including anallowance fornuclearpowerpeakingduetodensification.

0882L:614.2.6-7 Reactivit FeedbackWeihtinFactorsThelargesttemperature rises,andhencethelargestreactivity feed-backs,occurinchannelswherethepowerishigherthanaverage.Sincetheweightofaregionisdependent onflux,theseregionshavehighweights.Thismeansthatthereactivity feedbackislargerthanthatindicated byasimplechannelanalysis.

Physicscalculations havebeencarriedoutfortemperature changeswithaflattemperature distribu-tion,andwithalargenumberofaxialandradialtemperature distribu" tions.Reactivity changeswerecomparedandeffective weighting factorsdetermined.

Theseweighting factorstaketheformofmultipliers whichwhenappliedtosinglechannelfeedbacks correctthemtoeffective wholecorefeedbacks fortheappropriate fluxshape.Inthisanalysis, sinceaone-dimensional (axial)spatialkineticsmethodisemployed, axialweighting isnotnecessary iftheinitialcondition ismadetomatchtheejectedrodconfiguration.

Inaddition, noweighting isappliedtothemoderator feedback.

Aconservative radialweighting factorisappliedtothetransient fueltemperature toobtainaneffective.

fueltempera-tureasafunctionoftimeaccounting forthemissingspatialdimen-sion.Theseweighting factorshavealsobeenshowntobeconservative comparedtothree-dimensional analysis(Reference 4).Moderator andDolerCoefficient Thecriticalboronconcentrations atthebeginning oflifeandendoflifeareadjustedinthenuclearcodeinordertoobtainmoderator densitycoefficient curveswhichareconservative comparedtoactualdesignconditions fortheplant.Asdiscussed above,noweighting factorisappliedtotheseresults.0882L:614.2.6"8

TheDopplerreactivity defectisdetermined asfunctionofpowerlevelusingaone-dimensional steady-state computercodewithaDopplerweighting factorof1.0.TheDopplerdefectusedisgiveninSection3.0.TheDopplerweighting factorwillincreaseunderaccidentconditions, asdiscussed above.DelaedNeutronFractionBCalculations oftheeffective delayedneutronfraction(P)efftypically yieldvaluesno,lessthan0.70io'tbeginning oflifeand0.50%atendoflifeforthefirstcycle.Theaccidentissensitive to5iftheejectedrodworthisequaltoorgreaterthan5asinzeropowertransients.

Inordertoallowforfuturecycles,pessimistic estimates ofPof0.495atbeginning ofcycleand0.43:oatendofcyclewereusedintheanalysis.

TriReactivit Insertion Thetripreactivity insertion assumedisgiveninTable14.2.6-1andincludestheeffect~ofonestuckRCCA.Theshutdownreactivity wassimulated bydroppingarodoftherequiredworthintothecore.Thestartofrodmotionoccurred0.5secondsafterthehighneutronfluxtrippointwasreached.Thisdelayisassumedtoconsistof0.2secondfortheinstrument channeltoproduceasignal,0.15secondforthetripbreakertoopenand0.15.secondforthecoiltoreleasetherods.Acurveoftriprodinsertion versustimewasusedwhichassumedthatinsertion tothedashpotdoesnotoccuruntil1.8secondsafterthestartoffall.Thechoiceofsuchaconservative insertion ratemeans0882L:614.2.6-9

~l"*~~(

thatthereisoveronesecondafterthetrippointisreachedbeforesignificant shutdownreactivity isinsertedintothecore.Thisisaparticularly important conservatism forhotfull-power accidents.'eactor Protection Reactorprotection forarodejectionisprovidedbyhighneutronfluxtrip(highandlowsetting).

Theseprotection functions arepartofthereactortripsystem.Nosinglefailureofthereactortripsystemwillnegatetheprotection functions requiredfortherodejectionaccident, oradversely affecttheconsequences oftheaccident.

ResultsCasesarepresented forbothbeginning andendoflifeatzeroandfullpower.1.BeinninofCcleFullPowerControlbankDwasassumedtobeinsertedtoitsinsertion limit.Theworstejectedrodworthandhotchannelfactorwereconserva-tivelycalculated tobe.40%6kand5.61respectively.

Thepeakhotspotcladaveragetemperature was2543'F~Thepeakhotspotfuelcentertemperature reachedmelting,wasconservatively assumedat4990'F.However,meltingwas'restricted tolessthan10%ofthepellet.2.BeinninofCcle,ZeroPowerForthiscondition, controlbankDwasassumedtobefullyinsertedandbanksBandCwereattheirinsertion limits.Theworstejectedrodislocatedincontrolbank,Dandhasaworthof.78%6kandahotchannelfactorof7.80.Thepeakhotspotcladtemperature reached2639F,thefuel.centertemperature was3861F.0882L:614.2.6-10

3.EndofCcleFul1PowerControlbankDwasassumedtobeinsertedtoitsinsertion limit.Theejectedrodworthandhotchannelfactorswereconservatively calculated tobe.42%6kand5.69respectively.

Thisresultedinapeakcladaveragetemperature of2246F.Thepeakhotspotfueltemperature reachedmeltingconservatively assumedat4800~F.However,meltingwasrestricted tolessthan10~ofthepellet.4.EndofCcleZeroPowerTheejectedrodworthandhotchannelfactorforthiscasewereobtainedassumingcontrolbank0tobefullyinsertedandbanksCandBattheirinsertion limits.Theresultswere.95'o'hkand9.4'Frespectively.

Thepeakcladaverageandfuelcentertemperatures were2421and3449'F.TheDopplerweighting factorforthiscaseissignificantly higherthanfortheothercasesduetotheverylargetransient hotchannelfactor.Asummaryofthecasespresented aboveisgiveninTable14.2.6-1.

Thenuclearpowerandhotspotfuelandcladte'mperature transients fortheworstcasesarepresented inFigures14.2.6-1'through14.2.6"2(beginning"of-life fullpowerandbeginning-of-life

'zeropower).Thesequenceofeventsforthesetwocasesispresented inTable14.2.6-2.

Forallcases,reactortripoccursveryearlyinthetransient, afterwhichthenuclearpowerexcursion isterminated.

Asdiscussed previously, thereactorwillremainsubcritical following reactortrip.TheejectionofanRCCAconstitutes abreakintheReactorCoolantSystem,locatedinthereactorpressurevesselhead.Theeffectsandconsequences ofloss-of-coolant accidents arediscussed inSection14.3.Following theRCCAejection, theoperatorwouldfollowthesame0882L:614.2.6"ll

emergency instructions asforanyotherlossofcoolantaccidenttorecoverfromtheevent.FissionProductReleaseItisassumedthatfissionproductsarereleasedfromthegapsofallrodsenteringDNB.Inallcasesconsidered, lessthan10%ofth0rodsenteredDNBbasedonadetailedthree-dimensional THINCanalysis.

PressureSureAdetail.ed calculation ofthepressuresurgeforanejectionworthofonedollaratbeginning oflife,hotfullpower,indicates thatthepeakpressuredoesnotexceedthatwhichwouldcausestresstoexceedthefaultedcondition stresslimits.Sincetheseverityofthepresentanalysisdoesnotexceedthe"worstcase"analysis, theaccidentforthisplantwillnotresultinanexcessive pressureriseorfurtherdamagetothereactorcoolantsystem.LatticeDeformations Alargetemperature gradientwillexistintheregionofthehotspot.Sincethefuelrodsarefreetomoveintheverticaldirection, differ-entialexpansion betweeqseparaterodscannotproducedistortion.

However,thetemperature gradients acrossindividual rodsmayproduceadifferential expansion tendingtobowthemidpointoftherodstowardthehottersideoftherod.Calculations haveindicated thatthisbowingwouldresultinanegativereactivity effectatthehotspotsinceWestinghouse coresareunder-moderated, andbowingwilltendtoincreasetheunder-moderation atthehotspot.Sincethe14x14fueldesignisalsounder-moderated, thesameeffectwouldbeobserved.

Inpractice, nosignificant bowingisanticipated, sincethestructural rigidityofthecoreismorethansufficient towithstand theforcesproduced.

Boilinginthehotspotregionwouldproduceanetflowawayfromthatregion.However,theheatfromthefuelisreleasedtothe0882L:614.2.6-12

waterrelatively slowly,anditisconsidered inconceivable thatcrossflowwillbesufficient toproducesignificant latticeforces.Evenifmassiveandrapidboiling,sufficient todistortthelattice,ishypothetically postulated, thelargevoidfractioninthehotspotregionwouldproduceareduction inthetotalcoremoderator tofuelratioandalargereduction inthisratioatthehotspot.Theneteffectwouldtherefore beanegativefeedback.

Itcanbeconcluded thatnoconceivable mechanism existsforanetpositivefeedbackresulting fromlatticedeformation.

Infact,asmallnegativefeedbackmayresult.Theeffectisconservatively ignoredintheanalysis.

Conclusions Conservative analysesindicatethatthedescribed fuelandcladdinglimitsarenotexceeded.

Itisconcluded thatthereisnodangerofsuddenfueldispersal intothecoolant.Sincethepeakpressuredoesnotexceedthatwhichwouldcausestressestoexceedthefaultedcondition stresslimits,itisconcluded thatthereisnodangeroffurtherconsequential damagetothereactorcoolantsystem.Theanalyseshavedemonstrated thatthefissionproductrelease,asaresultofanumberoffuelrodsenteringDNB,islimitedtolessthan10%ofthefuelrodsinthecore.0882L:614.2.6"13

,TABLE14.2.6-1PARAMETERS USEDINTHEANALYSISOFTHERODCLUSTERCONTROL"ASSEMBLYEJECTIONACCIDENTTimeinLifeParameters Beginning Beginning EndEndPowerlevel,percentEjectedrodworth,percenthkDelayedneutronfraction, percentFeedbackreactivity weighting Tripreactivity, percenthkFbeforerodejectionqFafterrodejectionqNumberofoperational pumpsMaximumfuelpelletaveragetemperature,

'FMaximumfuelcentertemperature,

,FMaximumcladaveragetemperature,

'FMaximumfuelstoredenergy,cal/gMaximumfuelmelt,percentr102.40.491.34.02.55.61419049712543184<100.78).,491.4172.07.803422386126391450.01020.42.95.43.431.31.744.02.02.55.699.402'1372630994838344922462421160129<100.00882L:614.2.6-14 TABLE14.2.6-2TIMESEQUENCEOFEVENTSRCCAEJECTIONCasea.Beginning-of-Life,FullPowerEventInitiation ofrodejectionPowerrangehighneutronfluxsetpointreachedTimeofEachEventSeconds0.00.03PeaknuclearpoweroccursRodsbegintofallintocore0.53Peakfuelaveragetemperature occursPeakcladtemperature occursPeakheatfluxoccurs2.002.00b.Beginning-of-Life,ZeroPowerInitiation ofrodejection0.0Powerrangehighneutronfluxlowsetpointreached0.24PeaknuclearpoweroccursRodsbegintofallintocorePeakcladtemperature occursPeakheadfluxoccurs0.29/(0.742.14Peakfuelaveragetemperature occurs2.270882L:614'.6"15

~'

Figure14.2.6-1GinnaRCCAEjectionBeginning ofLife,FullPower10..0ClCICIClClClClnOOVSCCChbOClOPnClmTIHEtSKC)6000.05000.04000.03000.02000.0)000.00.oe)<en'p.Fue]A.VgClad0.0ClClCCCOClClOCDCIOCDClOClOTIW(SEC)IA9.6-16 Figure14.2.6-2GinnaRCCAEjectionBeginning ofLife,ZeroPowero10S-O.01C7CDCDCD3'8Iss-sCDeCDCDhlAJmTlHE<SEC)6000.05000.0loop.0poopopOOp0)000.000.0FuelCen~eFuelAvVg.CladCDtSKCiCDCDCD14.2.6-17

ATTACHMENT CLOCAACCIDENTANALYSISREVISEDFSARSECTIONS14.3.1/14.3.2 0882L:6C-1 A

TABLEOFCONTENTSSectionDescription Page14.3PrimarySystemPipeRuptures14.3.1-114.3.1LossofReactorCoolantFromSmallRupturedPipesorFromCracksinLargePipesWhichActuatesEmergency CoreCoolingSystem14.3.1-114.3.1References 14.3.1-714.3.2MajorReactorCoolantSystemPipeRuptures(LossofCoolantAccident) 14.3.2-114.3.2References 14.3.2-70458L:6 4/t1 LISTOFTABLESTableDescriptionPage14.3.1-)SmallBreak-TimeSequenceofEvents14.3.1"814.3.1-214.3.2-1SmallBreak-AnalysisInputandResultstLargeBreak-TimeSequenceofEvents)4.3.)"914.3.2-10 14.3.2-2LargeBreak-AnalysisInputandResults)4.3.2-11 14.3.2-3LargeBreak-Containment Data14.3'-1214.3.2"4RefloodPassandEnergyRelease14.3.2-15 14.3.2-5BrokenLoopAccumulator YiassandEnergyRelease14.3.2-16 0458L:6 LISTOFFIGURESFigureDeseriptionPage14.3.1-)a 14.3.)-lb HighHeadSafetyInjection FlowRatefLowHeadSafetyInjection FlowRate14.3.1-10 14.3.1-11 14.3.1-2HotRodAxialPowerShape14.3.1"12 14.3.1"3Depressurization Transient (6-Inch)14.3.1"13 14.3.1-4CoreMixtureHeight(6-Inch)14.3.1"14 14.3.1-5PeakCladTemperature Transient (6-Inch)14.3.1-15 14.3.1-6SteamFlowRate14.3.1-16 14.3.1-7RodFilmCoefficients14.3.1-17 14.3.1"8HotSpotFluidTemperature 14.3.1-18 14.3.)-9a Depressurization Transient (4-Inch)14'.1-1914.3.1-9b Depressurization Transient (8-Inch)14.3.1-20 14.3.1-10a CoreMixtureHeight(4-Inch)14.3.1-21 14.3.1-10b 14.3'-llaCoreMixtureHeight(8-Inch)CladTemperature Transient (4-Inch)14.3.1"22 14.3.1"23 14.3.1"1)b CladTemperature Transient (8-Inch)14.3.1-24 14.3.2-)a Fluidequality-DECLG(CD=0.8)14.3.2-17 0458L:6

LISTOFFIGURES(continued)

Description Page14.B<@'luid.

Quality-DECLG(CD=0.6)14.3.2"18 14.~>,FluidQuality-DECLG(CD=0.4)14.3.2-19 14.3'+zMassVelocity-DECLG(CD=0.8)14.3.2"20 14.3~+~MassVelocity-,DECLG(CD=0.6)~'4.3.2"21 14.3.Z>Mass,Velocity-DECLG(CD=0.4)14.3.2"22 14.3.23a,HeatTransferCoefficient

-DECLG(CD=0.8)14.3.2-23 14.3.2-au HeatTransferCoefficient

-DECLG(CD=0.6)14.3.2-24 14.3.,2-~~

HeatTransferCoefficient

-DECLG(CD=0.4)14.3.2-25 CorePressure-DECLG(CD=0.8)14.3.2"26 14.32-4'4.3.2-4c; CorePressure"DECLG(CD=0.6)CorePressure-DECLG(CD=0.4)14.3.2-27 14.3.2-28 14.3.2-5a 14.3.2"5b 14.3.2-5c BreakFlowRate-DECLG(CD=0.8)BreakFlowRate-DECLG(CD=0')BreakFlopRyte-DECLG(CD=0.4)14.3.2-29 14.3.2"30 14.3.2-31 14.3.2-6a CorePressureDrop-DECLG(CD=0.8)14.3.2"32 0458L:6 I0'/

LISTOFFIGURES(continued)

FigureDescription Page14.3.2-1b FluidQuality-DECLG(CD=0.6)14.3.2-18 14.3.2-1c 14.3.2-2a FluidQuality-DECLG(CD=0.4)MassVelocity-DECLG(CD=0.8)14.3.2"19 14.3.2-20 14.3.2-2b 14.3.2"2c MassVelocity-DECLG(CD=0.6)MassVelocity-DECLG(CD=0.4)14.3.2"21 14.3.2"22

)4.3.2-3a 14.3.2"3b HeatTransferCoefficient "DECLG(CD=0.8)14.3.2-23 lHeatTransferCoefficient

-DECLG(CD=0.6)14.3.2-24 14.3.2-3c HeatTransferCoefficient

-DECLG(CD=0.4)14.3.2-25 14.3.2-4a CorePressure-DECLG(CO=0.8)14.3.2-26 14.3.2-4b CorePressure-DECLG(CO=0.6)14.3.2-27 14.3.2"4c CorePressure-DECLG(CD=0.4)14.3.2-28 14.3.2-5a BreakFlowRate-DECLG(CD=0.8)14.3.2-29 14.3.2"5b BreakFlowRate-DECLG(CD=0.6)14.3.2-30 14.3.2-5c BreakFlowRate-DECLG(CD=0.4)14.3.2-31 14.3.2-6a CorePressureDrop-OECLG(CD=0.8)14.3.2-32 0458L:6

LISTOFFIGURES(continued)

FigureDescription Page14.3.2-6b CorePressureDrop-DECLG(CD=0.6)14.3.2-33 14.3.2-6c CorePressureDrop-DECLG(CD=0.4)14.3.2-34 14.3.2-7a PeakCladTemperature

-DECLG(CD=0.8)14.3.2-35 14.3.2-7b 14.3.2-7c PeakCladTemperature

-DECLG(CD=0.6)PeakCladTemperature

-DECLG(CD=0.4)14.3.2-36 14.3.2-37 14.3.2-8a FluidTemperature

-DECLG(CD=0.8)14.3.2-38 14.3.2-8b 14.3.2-8c FluidTemperature

-DECLG(CD=0.6)FluidTemperature

-DECLG(CD=0.4)14.3.2-39 14.3.2-40 14.3.2-9a CoreFlow(TopandBottom)-DECLG(CD=0.8)14.3.2-41 14.3.2-9b CoreFlow(TopandBottom)-DECLG(CD=0.6)14.3.2-42 14.3.2-9c CoreFlow(TopandBottom)-DECLG(CD=0.4)14.3.2-43 14.3.2-10a RefloodTransient

-CoreInletVelocity-14.3.2-44 DECLG(CD=0.8)14.3.2-10b RefloodTransient

-CoreInletVelocity-14.3.2-45 DECLG(CD=0.6)0458L:6 LISTOFFIGURES(continued)

FigureDescription Page14.3.2-10c RefloodTransient

-CoreInletVelocity-14.3.2-46 DECLG(CD=0.4)14.3.2-11a RefloodTransient

-CoreandDowncomer WaterLevels-DECLG(CD=0.8)14.3.2-47 14.3.2"lib RefloodTransient

-CoreandDowncomer WaterLevels-DECLG(CD=0.6)14.3.2-48 14.3.2-11c Reflo'odTransient

-CoreandDowncomer WaterLevels-DECLG(CD=0')14.3.2"49 14.3.2-12a Accumulator

.Flow(Blowdown)-

DECLG(CD=0.8)14.3.2-50 14.3.2-12b Accumulator Flow(Blowdown)-

DECLG(CD=0.6)14.3.2-51 14.3.2"12c Accumulator Flow(Blowdown)-

DECLG(CD=0.4)14.3,2-52 14.3.2-13a PumpedECCSFlow(Reflood)-(CD=0.8)14.3.2-53 14.3.2-13b PumpedECCSFlow(Reflood)

-(CD=0.6)14.3.2-54 14.3.2"13c PumpedECCSFlow(Reflood)-(CD=0.4)14.3.2-55 14.3.2-14a Containment Pressure-DECLG(CD=0.8)14.3.2-56 14.3.2-14b Containment Pressure-DECLG(CD=0.6)14.3.2-57 Yi0458L:6 LISTOFFIGURES(continued)

FigureDescriPtion Page14.3'-14cContainment Pressure-DECLG(CD=0.4)14.3.2"58 14'.2-15CorePowerTransient

-DECLG(CD=0.4)14.3.2-59

,14.3.2"16 BreakEnergyReleasedtoContainment-DECLG(CD=0.4)14.3.2-60 14.3.2-17 Containment WallCondensing HeatTransferCoefficient

-DECLG(CD=0.4)14.3.2"61 Vii0458L:6

14.3PRIMARYSYSTEMPIPERUPTURES14.3.1LossOfReactorCoolantFromSmallRupturedPipesOrFromCracksinLargePipesWhichActuatesEmergency CoreCoolingSystemIdentification ofCausesandAccidentDescritionAlossofcoolantaccidentisdefinedasaruptureofthereactorcoolantsystempipingorofanylineconnected tothesystemuptothefirstclosedvalve.Rupturesofsmallcrosssecti'onwillcauselossofthecoolantataratewhichcanbeaccommodated bythechargingpumpswhichwouldmaintainanoperational waterlevelinthepressurizer permitting theoperatortoexecuteanorderlyshutdown.

Amoderatequantityofcoolantcontaining suchradioactive impurities aswouldnormallybepresentinthecoolant,wouldbereleasedtothecohtainment.

Themaximumbreaksi.zeforwhichthenormalmakeupsystemcanmaintainthepressurizer levelisobtainedbycomparing

.hecalculated flowfromthereactorcoolantsystemthroughthepostulated breakagainstthechargingpumpmakeupflowatnormalreactorcoolantsystempressurei.e.,2250psia.Amakeupflowratefromonechargingpumpistypically adequatetosustainpressurizer pressureat2250psiaforabreakthrougha3/8in.diameterhole.Thisbreakresultsinalossofapproximately 17.5lb/sec.Shouldalargerbreakoccur,depressurization ofthereactorcoolantsystemcausesfluidtoflowtothereactorcoolantsystemfromthepressurizer resulting inapressureandleveldecreaseinthepressurizer.

Reactortripoccurswhenthepressurizer lowpressuretripsetpointisreached.Theconsequences oftheaccidentarelimitedintwoways:1.Reactortripandboratedwaterinjection complement voidformation incausingrapidreduction ofnuclearpowertoaresiduallevelcorresponding tothedelayedfissionandfissionproductdecay.0458L:614.3.1-1 2.Injection ofboratedwaterensuressufficient floodingofthecoretopreventexcessive cladtemperatures.

Beforethebreakoccurs,theplantisinanequilibrium condition, i.e.,theheatgenerated inthecoreisbeingremovedviathesecondary system.Duringblowdown, heatfromdecay,hotinternals andthevesselcontinues tobetransferred tothereactorcoolantsystem.Theheattransferbetweenthereactorcoolantsystemandthesecondary systemmaybeineitherdirection depending ontherelativetemperatures.

Inthecaseofcontinued heat,additiontothesecondary, systempressureincreases andsteamdumpmayoccur.Makeuptothesecondary sideisautomatically providedbytheauxiliary feedwater pumps.Thesafetyinjection signalstopsnormalfeedwater flowbyclosingthemainfeedwater lineisolation valvesandinitiates emergency feedwater flowbystartingauxiliary feedwater pumps.Thesecondary flowaidsinthereduction ofreactorcoolantsystempressure.

WhentheRCSdepressurizes to715psia,theaccumulators begintoinjectwaterintothereactorcoolantloops.Thereactorcoolantpumpsareassumedtobetrippedattheinitiation oftheaccidentandeffects'f pumpcoastdown areincludedintheblowdownanalyses.

AnalsisofEffectsandConseuencesMethodofAnalysisForsmallbreakslessthan1.0fttheWFLASHdigitalcomputercode2References 1,2and3,isemployedtocalculate thetransient depressurization ofthereactorcoolantsystemaswellastodescribethemassandenthalpyofflowthroughthebreak.Theanalysiswasperformed foranassumedsteamgenerator tubeplugginglevelof12,".andareactorcoolantsystemloop'lowrate of84,000gpm.SmallBreakLOCAAnalsisUsinWFLASHTheWFLASHprogramusedintheanalysisofthesmallbreaklossofcoolantaccidentisanextension oftheFLASH-4code,Reference 3,0458L:614.3.1-2

developed attheWestinghouse BettisAtomicPowerLaboratory.

TheWFLASHprogrampermitsadetailedspatialrepresentation ofthereactorcoolantsystem.Thereactorcoolantsystemisnodalized intovolumesinterconnected byflowpaths.

Boththebrokenloopandtheintactlooparemodeledexplicitly fortwoloopplants.Thetransient behaviorofthesystemisdetermined fromthegoverning conservation equations ofmass,energy,andmomentumappliedthroughout thesystem.Adetaileddescription ofWFLASHisgiveninReference 1and2.TheuseofWFLASHintheanalysisinvolves, amongotherthings,therepresentation ofthereactorcoreasaheatedcontrolvolumewiththeassociated bubblerisemodeltopermitatransient mixtureheightcalculation.

Themulti-node capability oftheprogramenablesanexplicitanddetailedspatialrepresentation ofvarioussystemcomponents.

Inparticular, itenablesapropercalculation othebehavioroftheloopsealduringaloss-of-coolant transient.

Safetyinjection flowratetothereactorcoolantsystemasafunctionofthesystempressureisusedaspartoftheinput.TheSafetyInjection (SI)systemwasassumedtobedelivering totheRCS,25secondsafterthegeneration ofasafetyinjection signal.Fortheseanalyses, theSIdeliveryconsiders pumpedinjection flowwhichisdepictedinFigures14.3.1-1a and14.3.1-lb asafunctionofRCSpressure.

Figure14.3.1-1a represents injection flowfromoneHHSIpumpbasedonperformance curvesdegraded5ofromthedesignhead.Figure14.3.1-1b represents injection flowfromoneLHSIpump.The25seconddelayincludestimerequiredfordieselstartupandloadingofthesafetyinjection pumpsontotheemergency buses.AlsominimumSafeguards Emergency CoreCoolingSystemcapability andoperability hasbeenassumedintheseanalyses.

I0458L:614.3.1-3 Peakcladtemperature analysesareperformed withtheLOCTAIYcode,References 2and4.InputforthecodeisobtainedfromtheWFLASHcodewhichdetermines theRCSpressure,'uel rodpowerhistory,steamflowpasttheuncovered partofthecoreandmixtureheighthistory.Figure14.3.1-2presentstheaxialpowershapeutilizedtoperformthesmallbreakanalysispresented here.Thispowershapewaschosenbecauseitprovidesanappropriate distribution ofpowerversuscoreheightandalsolinearpowerismaximized intheupperregionsofthereactorcore(.10ft.to12ft.).Thispowershapeisskewedtothetopofthecorewiththepeaklinearpoweroccurring atthe10ft.coreelevation.

Thelinearpowerforthispowershapeabove10ft.essentially matchestheshapeofthegenericoperation F~envelopefornormalplantoperation andhencelinearpowerismaximized forthe10ft.coreelevation andabove.Thisislimitingforsmallbreakanalysisbecauseoftheuncoveryprocessforsmallbreak.Asthecoreuncovers, thecladdingintheupperelevation ofthecoreheatsupandissensitive tothelinearpoweratthatelevation.

OThecladdingtemperatures inthelowerelevations ofthecore,belowthetwophasemixtureheight,remainslow.Thepeakcladtemperature occursabove10ft.ResultsofSmallBreakAnalsisThissectionpresentsresultsofthelimitingbreaksizeintermsofhighestpeakcladtemperature.

Theworstbreaksize(smallbreak)isa6in.diameterbreak.Thedepressurization transient forthisbreakisshowninFigure14.3.1-3.Theextenttowhichthecoreisuncovered isshowninFigure14.3.1-4.Ouringtheearlierpartofthesmallbreaktransient, theeffectofthebreakflowisnotstrongenoughtoovercometheflowmaintained bythereactorcoolantpumpsthroughthecoreastheyarecoastingdown0458L614.3.1-4 0

following reactortrip.Therefore, upwardflowthroughthecoreismaintained.

Theresultant heattransfercoolsthefuelrodandcladto'Iverynearthecoolanttemperatures aslongasthecoreremainscoveredbyatwophasemixture.Themaximumhotspotcladtemperature calculated duringthetransient is1092'Fincluding theeffectsoffueldensification asdescribed inReference 5.Thepeakcladtemperature transient isshowninFigure14.3.1-5fortheworstbreaksize,i.e.,thebreakwiththehighestpeakcladtemperature.

ThesteamflowratefortheworstbreakisshownonFigure14.3.1-6.

Whenthemixtureleveldropsbelowthetopofthecore,thesteamflowcomputedinWFLASHprovidescoolingtotheupperportionofihecore.Therodfilmcoefficients forthisphaseofthetransient aregiveninFigure14.3.1-7.

Thehotspotfluidtemperature fortheworstbreakisshowninFigure14.3.1-8.Thereactorscramtimeisequaltothereactortripsignaltimeplus4.4secondsforsignaltransmission androdinsertion.

Duringthisperiod,thereactorisconservatively assumedtooperateatratedpower.Additional BreakSizesAdditional breaksizeswereanalyzed.

Figures14.3.1-9aand14.3.1-9bpresenttheRCSpressuretransient forthe4and8in.breaksrespectively andFigures14.3.1-10aand14.3.1-10bpresentthevolumehistory(mixtureheight)plotsforthesebreaks.Thepeakcladtemperatures forthesecasesarelessthan-thepeakcladtemperature ofthe6in.break.Thepeakcladtemperatures forthesecasesaregiveninFigures14.3.1-llaand14.3.1-11b.Conclusions Analysespresented inthissectionshowthatthehighheadandlowheadportionsoftheemergency corecoolingsystem,togetherwith0458L:614.3.1-5 accumulators, providesufficient corefloodingtokeepthecalculated peakcladtemperatures belowrequiredlimitsof10CFR50.46.Hence,adequateprotection isaffordedbytheemergency corecoolingsytemintheeventofasmallbreaklossofcoolantaccident.

Table14.3.1-1presentstheresultsoftheseanalyses.

0458L:614.3.1-6 REFERENCES

-Section14.3.11.Esposito, V.J.,Kesavan,D.,Maul,B.A.,"WFLASH-A FORTRANIVComputerProgramforSimulation ofTransients inaMulti-Loop PWR",WCAP-8261, Rev.1,July,'974.

2.Skwarek,R.J.,Johnson,W.J.,andMeyer,P.E.,"Westinghouse Emergency CoreCoolingSystemSmallBreakOctober1975Model,"WCAP-8970-P-A (Proprietary) andWCAP-8971-A (Non-Proprietary)

January1979.3.Porsching, T.A.,Murphy,J.H.,Redfield, J.A.,andDavis,V.C.,"FLASH-4:

AFullyImplicitFORTRAN-IV ProgramfortheDigitalSimulation ofTransients inaReactorPlant",WAPD-TM-84; BettisAtomicPowerLaboratory, March,1969.4.Bordelon, F.M.,etal.,"LOCTA-IV Program:Loss-of-Coolant Transient Analysis",

WCAP-8301 (Proprietary Version),

WCAP-8305 (Non-Proprietary Version),

June1974.5.Hellman,J.M.,"FuelDensification Experimental Results'andModelforReactorAppl.ication",

WCAP-8219, gctober,1973.0458L:614.3.1-7 TABLE14.3.1"1SMALLBREAKTIMESEQUENCEOFEVENTSEvent'4in.6in.8in.Start0.00.00.0ReactorTripSignal(Sec.)12.510.09.5TopofCoreUncovered (Sec.)165.74.69.Accumulator Injection Begins(Sec.)323.138.75.PCTOccurs(Sec.)333.7121.492.0TopofCoreCovered(Sec.)374.168.101.0458L:614.3.1-8 TABLE14.3.1-2SMALLBREAKANALYSISINPUT.AND RESULTSResults4in.6in.8in.PeakCladTemp.'F9761092758PeakCladLocationFt.11.75,10.7510.75LocalZr/H20Rxn(max)X0.06780.06890.0675LocalZr/H20LocationFt.11.7510,.7510.75Total'r/H20 Rxn%%d<0.3<0.3<0~3HotRodBurstTimesecnoburstnoburst.noburstHotRodBurstLocationFt.noburstnoburstnoburstCa1cula.ionCorePowerMMt1004of1520PeakLinearPowerkw/ft102~ofSeeFigure14.3.1-2PeakingFactor(AtLicenseRating)Accumulator h'aterVolumeFt.3SeeFigure'4.3.1-2 11000458L:614.3.1-9

16001400FIGURE14.3.1-1aHIGHHEADSAFETYINJECTION FLOWRATEOnePunpinOperation 1200-1000800600400200102030S'.I.FLOW(lb/sec)4050600458L:614.3.1-.10 160140FIGURE14.3.1-1b LOWHEADSAFETYINJECTION FLOWRATEOnePumpinOperation 12010080604020IL0204060.80100120140160180200220S.I.FLOW(1b/sec)0458L:614.3.1-11 14FIGURE14.3.1-2HOTRODAXIALPOWERSHAPE1210U6C)CD02COREHEIGHT(Ft.)0458L:614.3.1-12 FIGURE14.3.1-3DEPRESSURIZATION

~iNSIENT(6INCH)O'Oool00'OOL00'005ICCIoo'oos~00'OowIvhXa.Vl4JCC~/1IClkIL00'OOK00'OOZ00'OOl0'0OoCICIoYtSII38flSS3tl4 S)ICI0458L:614.3.1-13 P

FIGURE14:3.1-4COREMIXTUREHEIGHT(6INCH)OOOOt00'00800'ool00'008IlalIIICClsJ~illoo'oos*I00'ooeZ4laIXaelab'X~eJlalCL'O~C00'00800'OOZ00'ool0'00458L:6CI<lglIH'Jl3H380314.3.1-14 C7CI

FIGVRE14.3.1-5PEAKCLADTEHPERATVRE TRANSIENT (6INCH)0'oool00'0060000800'OOCClC~I00009CIIVlgCCggEQIoOELCOOZELX4JZCsatCCL"oooos<00'00100'OOE00'OOZ00'Oot0'0oCl8tQ<3$33ei30)CI008!OH'4H3L'OAT OT1)CIoCICI0458L:614.3.1-15 FIGURE14;3.1-6STEAMFLOWRATE0'000100'OotIIACCIIw00'os<I~JuVlVlZsoXU4a00'Oot00'00C00'00200'0010'08888)35/illAolgHr315CI8888888S.0458L:614.3.1-16

EDVl00I~~Ch600.00$00.00F00.00~300.00g200.00RGE6INCIISNAtlBRCAKTRANSIENT HEATTRARS.COEF.NOT ASSTBURSTe 10.00FTllP(AX~I0.75FTT~l60.000~50.000i0.00030.000X'0.0006.0000$.0000~.00003.00002.0000XlEDC3mOmnImI.OOOOooCICICICICICICICITIH([SEC)DooMICIDooClCIoCI

.T000.0RGE0IREHSHALLBREAKTRAllSIEllT FLUIDTEMPERATURE BURST~l0.00FTl)PEAK,IO.TSFll~l2500.02000.0IIS00.0oI000.0ClCTlX7mCJCrJImm500.000.0ClCIClCloAlCImClClCloInTIHElSEEIClClooEOClCloo C7Vl00I~~C7l8000.0ZS00.0RGfIIkUFLASHSHALLSRfAKTRAlISIfNI RCSPRfSSURf(PSIACL%00.0~CEA~CIS00.0-A7IlKlfoalGDIlOI000.00.0ClCIClClClAjoOoClooVICloCl CJlCOI~~2800.02500.0AfE8IHVfLEASHSHALLBREA'RAHSIEHJ.

RESIRE(CUREIPSIX2000.0CCVlMO:CLI500.0I000.0IC)C7mmt/lIjl~-emQIIIIm500.00n0.0ooooooP%oooTIHE(SEC)oaooEOooooo

11.00012.500AGEIIIIVFEASHSHAlBREAK1RAIISIEH1'OAE HI,IGHIIFFI10.000XCIu750QQnC3m5.00002.5000>CJmmWC)~DC)Ql0.0CICIClTIHE(SECICl33ClQl

EDVlCOI~~CJlIA.QQQI8.500RGE8IHVELASHSHALLBREAKTRANSIEIIT EQREHEIGHTIFTIGOItOtOIlal0.000Ku75QQQ5.0000Tl>CAm~~4lIOoocr2.50000.0OOOCIOOOoAlOOOOOOTIHE(SEE)OOOOO07OO~DEtlOOO

%00.0RGfIIIICH'SHALLBRfAaTRAIIS1fIIT CLAOAVGTfHP~HOTR00BURST~'10~00TTI)PfAR~11~75TTI+IR.SOO.Oo2000.0oZ1500.00CI1000.00.0ooooooooomoooT1HfIS'fCIo

C)CJlCQI~~Ch3000.0RCE8INCHSHALLBREAKTRANSIENT CLAOAVG.TEHP.HOT ROOBOASTS10.00ETI)PEAK~I0.15FTI~Iw+QQ.Qwcl2000.0ZI500.0mClmAlmt/0mI000.0500.000.0ClClClClClCDClAl8ClClClnTIHEISEC)CIClClClCl 14.3.2MajorReactorCoolantSystemPipeRuptures(LossofCoolantAccident)

Theanalysisspecified by10CFR50.46,"Acceptance Criteriafor,Emergency CoreCoolingSystemsforLightVaterPowerReactors",

Reference 1,ispresented inthissection.TheresultsofthelossofcoolantaccidentanalysisareshowninTable14.3.2-2andshowcompliance withtheAcceptance Criteria.

Theanalytical techniques usedareincompliance withAppendixKof10CFR50,andaredescribed inlistedreferences.

Theresultsforthesmallbreakloss-of-coolant acciden.arepresented inSection14.3.1andareinconformance with10CFR50.46andAppendixKof10CFR50.Shouldamajorbreakoccur,depressurization ofthereactorcoolantsystemresultsinapressuredecreaseinthepressurizer.

Reactortripsignaloccurswhenthepressurizer lowpressuretripsetpointisreached.Asafetyinjection systemsignalisactuatedwhentheappropriate setpointisreached.Thesecountermeasures willlimittheconsequences oftheaccidentintwoways:1.Reactortripandboratedwaterinjection complement voidformation incausingrapidreduction ofpower.oaresiduallevelcorresponding tofissionproductdecayheat.12.Injection ofboratedwaterprovidesheattransferfromthecoreandpreventsexcessive cladtemperatures.

1Atthebeginning oftheblowdownphase,theentirereactorcoolantsystemcontainssubcooled liquidwhichtransfers heafromthecorebyforcedconvection withsomefullydeveloped nucleateboiling.Afterthebreakdevelops, thetimetodeparture fromnucleateboilingiscalculated, consistent withAppendixKof10CFR50.Thereafter, the0458L:6l4.3.2-l coreheattransferisunstable, withbothnucleateboilingandfilmboilingoccurring.

Asthecorebecomesuncovered, bothturbulent andlaminarforcedconvection andradiation areconsidered ascoreheattransfermechanisms.

Whenthereactorcoolantsystempressurefallsbelow715psiatheaccumulators begintoinjectboratedwater.Theconservative assumption ismadethataccumulator waterinjectedbypassesthecoreandgoesoutthroughthebreakuntilthetermination ofbypass.Thisconservatism isagainconsistent withAppendixKof10CFR50.CorePowerTransient DurinBlowdownThecorepower.ransient duringblowdownforlargebreaksisevaluated usingtheSATAN-VIcomputercode.Thiscodeisdiscussed indetailinWCAP-8306, Reference 3.ThermalAnalsisPerformance CriteriaforEmerencCoreCoolinSstemThereactorisdesignedtowithstand thermaleffectscausedbyalossofcoolantaccidentincluding thedoubleendedseverance ofthelargestreactorcoolingsystemcoldlegpipe.Thereactorcoreandinternals togetherwiththeemergency corecoolingsytemaredesignedsothatthereactorcanbesafelyshutdownandtheessential heattransfergeometryofthecorepreserved following theaccident.

Theemergency corecoolingsystem,evenwhenoperating duringheinjection modewiththemostseveresinglefailure,isdesignedtomeettheAcceptance Criteria.

0458L:614.3.2-2

MethodofThermalAnalsisThedescrsptson ofthevariousaspectsoftheLOCAanalysisisgiveninthelistedreferences.

Thisdocumentdescribes themajorphenomena modeled,theinterfaces amongthecomputercodesandfeaturesofthecodeswhichmaintaincompliance withtheAcceptance Criteria.

TheSATAN-VI,

WREFLOOD, andLOCTA-IVcodesusedinthisanalysisaredescri.bed indetailinWCAP-8306, Reference 3,WCAP-8171, Reference 5andWCAP-8305, Reference 4,respectively.

Thecontainment parameters usedinthecontainment analysiscodetodetermine theECCSbackpressure arepresented inTable14.3.2-3.

Thecontainment pressureanalysiscode(COCO)isde'scribed inWCAP-8326, Reference 6.Thelargebreakanalysiswasperformed withtheNRCApproved1981VersionoftheEvaluation Model,Reference 24,whichincludesmodifications delineated inWCAP-9220-P-A andWCAP-9221-A (1981),andcomplieswithAppendixKof10CFR50.46.Theanalysiswasperformed foranassumedsteamgenerator tubeplugginglevelof12andareactorcoolantsystemloopflowrateof84,000gpm.ResultsTable14.3.2-2presentsthepeakcladtemperatures andhotspotmetalreactionforalargebreakoverarangeofdischarge coefficients orbreaksizes.Thisrangeofdischarge coefficients wasdetermined toincludethelimitingcaseofpeakcladtemperature fromthesensitivity studies.Theanalysisofthelossofcoolantaccidentisperformed at102,.ofratedcorepower.Thepeaklinearpower,andcorepowerusedintheanalysesaregiveninTable14.3.2-2.

Theequivalent coreparameter atthelicenseapplication powerlevelarealsoshowninTable14.3.2-2.

Sincethereismarginbetweenthevalueofthe-peaklinearpowerdensityusedinthisanalysisandthevalueexpectedinoperation, alowpeakcladtemperature wouldbeobtainedbyusingthepeaklinearpowerdensityexpectedduringoperation.

0458L:614.3.2-3 Fortheresultsdiscussed below,thehotspotisdefinedtobethelocationofmaximumpeakcladtemperature.

ThislocationisgiveninTable14.3.2-2foreachdischarge coefficient orbreaksizeanalyzed.

Tables14.3.2-4and14.3.2-5presentrefloodmassandenergyreleasestothecontainment andthe'roken loopaccumulator massandenergyreleasetothecontainment, respectively.

Figures14.3.2-1through14.3.2-16 presentthetransients fortheprincipal parameters forthedischarge coefficients analyzed.

Thefollowing itemsarenoted:Figures14.3.2-1a guality,massvelocity, andcladheattransfercoef-through14.3.2-3c ficientforthehotspotandburstlocations.

Figures14'.2-4athrough14.3.2-6c Corepressure, breakflow,andcorepressuredrop.Thebreakflowisthesumoftheflowrates frombothendsoftheguillotine break.Thecorepressuredropistakenasthepressurejustbeforethecoreinlettothepressurejustbeyondthecoreoutlet.Figures14.3.2-7a Cladtemperature, fluidtemperature, andcoreflow.through14.3.2-9c Thecladandfluidtemperatures areforthehotspotandburstlocations.

Figures14.3.2-10a RefloodTransient

-CoreInletVelocitythrough14.3.2-10c IFigures14.3.2"11a RefloodTransient

-CoreandDowncomer WaterLevelsthrough14.3.2.11cFigures14.3.2-12a Emergency corecoolingsystemflowrates, forboththrough14.3.3-13a accumulator andpumpedsafetyinjection.

045SL6l4.3.2-4 Figures14.3.2-14a Containment pressure.

through14.3.2"14c Figure14.3.2-15 CorepowerFigures14.3.2-16 Breakenergyreleaseduringblowdownandthecon-and14.3.'2-17tainmentwallcondensing heattransfercoefficient fortheworstbreak.Thecladtemperature analysisisbasedonatotalpeakingfactorof2.32.Thehotspotmetal-water reactionreachedis2.1%whichiswellbelowtheembrittlement limitof17~asrequiredby10CFR50.46.Inaddition, thetotalcoremetal-water reactionislessthan0.3~forallbreaksascomparedwiththe1Xcriterion of10CFR50.46.TheresultsofECCSevaluations andsensitivity studiesarereportedinReferences 2,7,8,9,26,12,13,16,18,20and24.Theseresultsarereportedonagenericandplantspecificbasis.Conclusions Forbreaksuptoandincluding thedoubleendedseverance ofareactorcoolantpipe,theemergency corecoolingsystemwillmeettheacceptance criteriaaspresented in10CFR50.46.Thatis:1.Thecalculated peakfuelelementcladtemperature isbelowtherequirement of2200~F.)2.Theamountoffuelelementcladdingthatreactschemically withwaterorsteamdoesnotexceed1percentofthetotalamountofZircaloyinthereactor.0458L:6l4;3.2-5 i3.Thecladtemperature transient isterminated atatimewhenthecoregeometryisstillamenabletocooling.Thelocalized claddingoxidation limitof17percentisnotexceededduringorafterquenching.

4.Thecoreremainsamenabletocoolingduringandafterthebreak.5.Thecoretemperature isreducedanddecayheatisremovedforanextendedperiodoftimeasrequiredbythelong-lived radioactivity remaining

.inthecore.ThetimesequenceofeventsforallbreaksanalyzedisshowninTable14.3.2-1.

Basedontheeffectofupperplenuminjection forWestinghouse designed2loopplants,a21Fincreaseinpeakcladtemperature resultsfromassuming14xl4OFAfuelforR.E.GinnaUnit1.Themethodology employedtodevelopthispenaltywasidentical tothatperformed forpreviousLOCAanalysesperformed forthe.plant,Reference 19,and27.Utilizing thepresentWestinghouse ECCSevaluation models,References 14,15,16and24,toanalyzeapostulated LOCAinR.E.GinnaUnit1,resultsinafinalpeakcladtemperature of1854Fincluding theUPIpenalty.Itcanbeseenfromtheresultscontained hereinthatthisECCSanalysisforR.E.Ginnaremainsincompliance with10CFR50.46.0458L:614.3.2.-6 REFERENCES

-Section14.3.21."Acceptance CriteriaforEmergency CoreCoolingSystemsforLightWaterCooledNuclearPowerReactors:

10CFR50.46andAppendixKof10CFR50.46,"FederalRegister, Volume39,Number3,January4,1974.2.Bordelon, F.M.,Massie,H.W.,andZordan,T.A.,"Westinghouse ECCSEvaluation Model-Summary,"

WCAP-8339, July1974.3.Bordelon, F.M.,etal.,"SATAN-VI Program:Comprehensive Space-Time Dependent AnalysisofLoss-of-Coolant",

WCAP-8302 (Proprietary Version),

WCAP-8306 (Non-Proprietary Version),

June1974.4.Bordelon, F.M.,etal.,"LOCTA-IV Program:Loss-of-Coolant Transient Analysis",

WCAP-8301 (Proprietary Version),

WCAP-8305 (Non-Proprietary Version),

June1974.5.Kelly,R.D.,etal.,"Calculational ModelforCoreReflooding afteraLoss-of-Coolant Accident(WREFLOOD Code)".WCAP-8170 (Proprietary Version),

WCAP-8171 (Non-Proprietary Version),

June1974.6.Bordelon, F.M.,andMurphy,E.T.,"Containment PressureAnalysisCode(COCO)",WCAP-8327 (Proprietary Version),

WCAP-8326 (Non-Proprietary Version),

June1974.7.Bordelon, F.M.,etal.,"TheWestinghouse ECCSEvaluation Model:Supplementary Information",

WCAP-8471 (Proprietary Version),

WCAP-8472 (Non-Proprietary Version),

January1975.8.Salvatori, R.,"Westinghouse ECCS-PlantSensitivity Studies",

WCAP-8340 (Proprietary Version),

WCAP-8356 (Non-proprietary Version),

July1974.0458L:614.3.2-7

REFERENCES

-Section14.3.2(cont)9.Delsignore, T.,etal.,"Westinghouse ECCSTwo-LoopSensitivity Studies(14x14)"WCAP-8854 (Non-Proprietary Version),

September 1976.10."Westinghouse ECCSEvaluation Model,October,1975Versions",

WCAP-8622 (Proprietary Version),

WCAP-8623 (Non-Proprietary Version),

November1975.11.LetterfromC.Eicheldinger ofWestinghouse ElectricCorporation toD.B.VassalooftheNuclearRegulatory Commission, letternumberNS-CE-924, January23,1976.12.Kelly,R.D.,Thompson, C.M.,etal.,"Westinghouse Emergency CoreCoolingSystemEvaluation ModelforAnalyzing LargeLOCAsDuringOperation withOneLoopoutofServiceforPlantswithoutLoopIsolation Valves",WCAP-9166,

February, 1978.13.Eicheldinger, C.,"Westinghouse ECCSEvaluation Model,February1978Version",

WCAP-9220 (Proprietary Version),

WCAP-9221 (Non-Proprietary Version),

February, 1978.14.LetterfromT.M.AndersonofWestinghouse ElectricCorporation toJohnStolzoftheNuclearRegulatory Commission, letternumberNS-TMA-8130, June1978.15.LetterfromT.M.AndersonofWestinghouse ElectricCorporation toJohnStolzoftheNuclearRegulatory Commission, letternumberNS-TMA-1834, June20,1978.16."SafetyEvaluation ReportonECCSEvaluation ModelforWestinghouse Two-LoopPlants",November, 1977.0458L:6l4.3,2-8 REFERENCES

-Section14.3.2(cont)17.LetterfromS.BursteinofWisconsin ElectricPowerCo.toE.G.CasisoftheNuclearRegulatory Commission, January17,1978.18.LetterfromR.L.KellyofWestinghouse ElectricCorporation toT.R.WilsonofWisconsin ElectricPowerCo.,letternumberWEP-78-2, February24,1978.19."NRCQuestions Regarding the1/16/78Submittal byWestinghouse DesignedTwo-LoopPlantOperators",

February1,1978.20.RefertoReference 18.21'SafetyEvaluation ReportonInterimECCSEvaluation modelforWestinghouse Two-LoopPlants",March1978.22.LetterfromS.BursteinofWisconsin ElectricPowerCo.toE.G.CaseoftheNuclearRegulatory Commission, March16,1978.23.LetterfromS.BursteinofWisconsin ElectricPowerCo.toE.G.CaseoftheNuclearRegulatory Commission, April6,1978.24.Eicheldinger, C.,"Westinghouse ECCSEvaluation Model,1981Version",

WCAP-9220-P-A (Proprietary Version)andWCAP-9221-A (Non-Proprietary),

Revision1,1981.25.Johnson,W.J.andThompson, C.M,"Westinghouse Emergency CoreCoolingSystemEvaluation Model-ModifiedOctober1975Version",

WCAP-9168 (Proprietary) andWCAP-9169 (Non-Proprietary),

1977.26."Westinghouse ECCSEvaluation ModelSensitivity Studies",

WCAP-8341 (Proprietary) andWCAP-8342 (Non-Proprietary),

1974.27.LetterfromR.A.Wiesemann, (Westinghouse) toD.Eisenhut(NRC),December11,1979.0458L:6l4.3.2-9

0TABLE14.3.2-1LARGEBREAKTINESEQUENCEOFEVENTSDECLG(CD08)DECLGCD0'6DECLGCDSecSTARTReactorTripSignalS.I.SignalAcc.Injection EndofBlowdownPvmpInjection BottomofCoreRecoveryAcc.Empty0.00.5810.474.5916.84825.4731.95851.3550.00.5890.545.7814.41625.5432.99053.8680.00.6020.658.2423.45425.6538.78556.100458L:614.3.2-10 TABLE14.3.2-2LARGEBREAKANALYSISINPUTANDRESULTSResultsDECLG(CD0'8DECLGCD0'6DECLG(CD=0.4)PeakCladTempFPeakCladTemp.LocationFt.Local.Zr/H20Rxn(max)LocalZr/H20LocationFt.TotalZr/H20RxnHotRodBurstTimesecHotRodBurstLocationFt~17517.51.57.5<0.364.86.7517307.51.47'<0.365.87.01833"7.52.17.5<0.353.06.0Calculation CorePowerMWt102~ofOPeakLinearPowerkw/ft102ofPeakingFactor(AtDesignRating)Accumulator WaterVolume(CubicFootperTank)Accumulator Pressure(psia)NumberofSafetyInjection PumpsOperating SteamGenerator TubesPlugged152013.4852.3211007152Fuelregion+cycleanalyzedCycleRegionR.E.GinnaR.E.GinnatoSpecifyWestinghouse OFARegion"A21FPCTpenaltymustbeaddedtotheanalysisvaluetoaccountforUPIinjection penalty.0458L:614.3,2-1,l TABLE14.3.2-3LARGEBREAKCONTAINMENT DATA(DRYCONTAINMENT)

NetFreeVolume1.066x10ftInitialConditions PressureTemperature RWSTTemperature ServiceWaterTemperature OutsideTemperature 14.7psia90oF60~F35oF10oFSpraySystemNumberofPumpsOperating RunoutFlowRateActuation Time21800gpmeach10secsSafeguards FanCoolersNumberofFanCoolersOperating FastestPostAccidentInitiation ofFanCoolers30secs0458L:6l4.3.2-l2

TABLE14.3.2-3(Cont.)STRUCTURAL HEATSINKDATADescritiveSurfaceAreaExposedtoContainment AtmoshereLayerThickness

~Laerinsulated portionofdomeandcontainment wall361811"1/4"3/8ll2I6llInsulation SteelConcreteuninsulated portionofdome12474ft.23/8"2I6llSteelConcretebasementfloor7955ft.2'/8"2IConcrete'teel Concretewallsofsumpinbasementfloor2342ft.5I3/8"3I6llConcreteSteelConcretefloorofsump297ft.2I3/8"2IConcreteSteelConcreteinsideofrefueling cavity5200ft.1/4'I2I6llSteelConcretebottomofrefueling cavity1200ft.1/4ll2I6llSteelConcreteareaonoutsideofrefueling 6900ft.2cavitywalls2I6llConcreteareainsideofloopandsteamgenerator compartment 14900ft.2I6llConcretefloorareaintermediate level6170ft.Concreteoperating floor1-1/2"thickI"beam6540ft.3151ft.2'-1/2"ConcreteSteel0458L:614.3.2-13

DescritiveSurfaceTABLE14.3.2-3(Cont.)STRUCTURAL HEATSINKDATAAreaExposedtoContainment AtmoshereLayerThickness

~Laer1"thickI-beam1/2"thickI-beamcylindrical supportsforS.G.8MCP's5016ft.8138ft.430ft.Steel1/2IISteel1/2"'teelplantcranerectangular supportcolumns5756ft.3/4"Steelbeamsusedforcranestructure 6023ft,Steelstructure onoperating floor2622ft.2'oncrete fromFSAR:grating,stairsmisc.steels7000ft.0104Steel0458L:6')4.3.2-14 0

TABLE14.3.2-4REFLOODMASSANDENERGYRELEASETimesecMassFlowlb/secEnerFlowBTU/sec38.785'9.040.045.059.069.079.089.099.0119.0139.0189.0399.00~00.008310.0089431.42692.496140.052173.728187.574193.280198.273202.387211.909240.9900.010.66011.47940535.91106286.56 118474.87 125391.45 126392.58 125252.45 121543'9117805.11 109040.96 92584.280458L:614.3.2-l5 TABLE14.3.2-5BROKENLOOPACCUMULATOR MASSANDENERGYRELEASETimesecMassFlowlb/secEnerFlowBTU/sec1.0102.0103.0104.0105.0106'107.0108.0109.01010.01011.01012.01013.01014F01015.01016.01017.01018.01019.01020.0102549.2622435.2932336.0902248.9812170.6862099.7242035.2371976.3881922.5601873.2131827.7661785.5491746.2011709.4241675.0791643'181613.4961585.6601559.4141534.659152779.845 145949.518 140004.203 134783.662 130091.404 125838.537 121973.761 118446.914 115220.959 112263.515 109539.815 107009.731 104651.550 102447.507 100389.155 98479.727 96698.419 95030.182 93457.249 91973.651 0458L:614.3.2-l6 I.1000I.ZSOORCKIaXIaOFAtO~0.0350PSISIR/C/L~ISKRIKSS.C.IZPKRCKNFIUQKPLVCCINCIIHOOKLt~310FOOVALITYOFFLUIOSVRSIeC15FTIlPKAR~1~50Fll~I~zI.00000.15000.5000O.ZSOO0.0CIC7aCSClCIC700gOOOOOR5QS~~~~~~00OOOO~88im83P.ruaenoranIIHKlSE'tl~fg~gg~pFigureill.3.2-1a FluidQuality-DECLG(CD=0.8)

I.loooI.F500ACEI~XIiOFACO~0~C350PSI5/1/C/L~1SESIESS.C.ItPEICESTruSKPLUCCI~CSlMODEL2.3ZOfoouALITYoffLUID~USST~l,ooFTIIPEAK~I50FTt~14JLtaWI.OOOO07500Io~0-5000O.ZXN0.0IINIwmlmm000OD~mmmIINNII5I5!fsSP.8Q8TIHE(SEC)88888888~I~PI8$NFigure14.3.2-1b FluidQuality-DECLG(CD=0.6)

I.F000I.ZVOICEI1XIlOFACOO.a350PSISIR/CIL1iSERIESS.C.IZPERCENTTVBEPLVCCINCOIHOOELR.jtFOOVALITTOFFLVIO1VRST~C~00FTlIPEANt7+$0FTI~lRWEJ4C!.00003O.NOI0o5000O.85000.0k.mIN)kmooooooooII'kmmmmP.

I555555P.SSod@88888888~8888888TIHE(SEEIFi"ure1l.3.2-1cFluid(}ua]ity-DECLG(CD=0.>I) 500.00~00.00RCEIIXIIOCACDO.ll50PSIS/R/C/EiISERIESS.C.IZPERCERTTUQEPLUCCINCIIHODEI.Z.lZOFOHASSVELOCIIYeuRSr.C.)5rTIIPEAr.7.50FII~I100.000000000800000~~~~~~~0000~z883mm(MIIp~8IZIRCONIIMEISECIFigure1II.3.2-2a t'lassVelocity-DECLG(CD=0.8)

RCEIIXI4OFACO~0.C3SOPSISIRIC/LI~SERIESS.C,IZtERCENTTUSEPLUCCIllC OIHOOELtolt0FOHASSYELOCITT~URST1QQFTIlPEAK~7.$0FTl~l3mICSIIIm0O00OOOOINIIImmmmm.TIHElSECl88888888-IIII55HNFigure14.3.2-2b HassVelocity-DECLG(CD=0.6)

RCEIAXIIOFACO>0.4350PSIS/R/C/L~lSERIESS.C.I2PERCENTTUBEPLUGCINCBlHOOEL2.32FOHASSVELOCITYBURStC.OOFTIIPEAKT~SOFTI~ICIIIIIPmmmIIIMNNNI.m<NP'R~SSRRBTIHE(SEC)Figure14.3.2-2c t1assVelocity-DECLG(CD=0.4)

a600.00500.00100.00300.00Z00.00RGEliXliOFACO*0.B350PSIS/R/C/I.11SERIESS.G.IZPERCERTTUBEPLUGGIHGBlHOOEL2.320FOHEATTRANS.COEFFICIENT

.BURST~Cl5FTIIPEAKED3.50FIl~I10.00030.00020.0006.00005.00001.00003.00002.0000I.ooooCI88TIHE(SEC)Fi"ure14.3.2-3a HeatTransferCoefficient

-DECLG(CD=0.0)

IlltII600.00500.00000.00zX0,00%0.00RCE11XtlOFAEoio.C350PSlS/R/E/Ll~SERlESS.C.tZPERCERTTUBEPLUCCIRCelHOOELCe)toFOHEATTRANS~COEFFtctERT BuRSt.T.OOFTT)PEAR~F~SoFTt~>40.000g,QQQx%.000IKC.00005.00001.0000).00008.0000l.0000ClCICI88TlHElSECI88IVCICIFigure14.3.2-3b fleatTransferCoefficient

-DECLG(CD=0.6) 0 IIIC00.00500.00100.00300.00200.00ROEl~XliOFACO~0.i350tSlS/RIC/LiiSERlESSC.l2PERCENTTUSEPLUGClkCIlKOOEL232FONEATTRANS.COEFFlClENT BURST.C.OOFTt)tEAKT.50FTE~I=.I':lli0.00030.00020.000IMi.00003.00002.0000l.000088TlHE(SEC)eicure14.3.2-3c lleatTransferCoefficient

-DECLG(CD=0.4) t500.0RCEOEAIt+T-l'I50PSI.CO~0.II'SllCONFICURATION CONTROLECCSHOOELSCO+OolOECLCFRY"IltPRESSURECORE.SOTTOHlITOP~l~l&F000.0IS00.0lalILI000.0.0.0C)TIHEISEC}Figure14.3.2-4a CorePressure-DECLG(CD=0.0)

t500.0RCEOFAIt+T.P.350PSICO%0.CItlICONFICURATIOII CORTROLECC5NOOEL'5CO~O+COECLCPXY~laltPRE'SSURE COREIQTTOKIITOP~I~It000.0ISOO0MI000.00.0C7CITIKEISEC)Figure1II.3.2-4b CorePressure-DECLG(CD=0.6)

Z500.0RCEOFAIZ+E.P.350PSICONO~1l90IKOllflCURAIIOll CO)IIROLEKCSHOOELSCO*0.IOECLGFRV*I.IZPRESSURECOREOOITOH))IOP~I~)Z000.0I500.0I000.00.0oCPCl0'IIHEISEC)Figurel~.3.2-IIc CorePressure-DECLG(CD=O.II)

I.OOEN5ACEOfh,lt+T.t.3SOFSICO<0.0ISllCOHCICUAATIOll COIITAQLECCSHOOELSCO44OECLCFRYIItPAEArFLOVvl.OOE+OlC.OOENNleCCED1.00EMt.OOERs0.0TIIIE(SECIFigure1II.3.2-5a BreakFlowBate-DECLG(CD.=-0.8) l I.OOEK5hCEOFAIZ+T.t.HOPSICORO.CISOICORI'ICUhATION CONThOLfCC$HOOELSCDIO.COECLPTXYII

~IZ~hfAXfLOUEIO.OOE+01C.OOEttNKWOhi.OOEK)it.OOENH0.0C7~4~nC7TIHEISECIFigure14.3.2-5b BreakFlewRate-DECLG(CD=0.6) 5.00C<iRC(OFA12+1.P.350P51CO*0.i1981COHPICURAflOk COHlROLECC'SHOOCLSCO*0.4OECLGFxY=l.12SR(AKFLOV3.00EK)i3aC72.00C~0i1.00CRa0.0oOCIT1xC1SCC1C7VlFigureill.3.2-5c BreakFlowRate-DECLG=(CD

=0.4) 70.000RCEOFAlt+T.t.350PS(CO~0I1'0~ICONFlCURATION CORTROLECCSHOOELSCOiO.IOECLCFXY>laitCOREtR.OROP~t5.0000.0RS.OOO-10.000aTi~E~SECiF'igure14.3.2-6a CorePressureDrop-DECLG(CD=08) lo.nRCEOFAIt+-T.P.350PSICORO.CISlICOIIFICURATIOIICONTROLECCSHODELSCO+0COECLCf<>+I~ItCOREPR.DROP4.~t5.000CI~COWClEl0.0<S.000-70.000ClC7oTIHEISECIFigure14.3.2-6b CorePressureDrop-DECLG(CD=0.6)

10.000ACEOfAl2+l.P.350PSlCO*0.i198ICORFICURAflOII CORJROLECCSHOOEESCO*0.IOECLGFXY*l~lZCOBEPR.OAOPCLi5.000Ch0.0-50.000-10.000CIOllHEISEC)C7vsFigureill.3.2-6c CorePressureDrop-DECLG(CD=0.>l)

Z500.0RGEIIXIIOFACO~0.0350PSI5/R/C/Ll1SERIESS.G.IZPERCENTTUSEPLUGGINGOlHOOELZ~IZOFOCLAOAVG.TEHP.HOT ROOQURSTC.15FTlI.PEAR750Fll~1Vl~Z000.0ClISN.OXCLZIt7o1000.00.0CICI88TIHE(SEC)Figure1II.3.2-7a Peal<CladTemperature

-DECLG(CD=0.8)

2500.0RCEIIX11OFACD~0.C350PSIS/R/C/L1~SERIKSS.C.IZPERCENT-TUSK PLUCCINC~IHODELi~Tt0FOCLADAYC.TEHP,NOT RODBURST7.00FTIIPEAR~I~50FTl~I~%00.0ClC3CgI500.0XE.XlJoIOOO.OlJ0.0C)C)88TIHEtSECIFigure14.3.2-7b Peal<CladTemperature

-DECLG(CD=0.6)

RCKTlXTiOT'ACOIOI350.PSISIRICILSERllSS.C.12PERCENTTUSEPLUCClNCITHOOELl3ZPOCLAOAVC~TKHP.NOTROOQURSTC~00Fll1PfAK~T$0Tl~l~2000.0oCICC1500.0XE.XIN~Jo1000.0~N500.000.0o88TIHE(SECIFigure1ll.3.2-7c Peal<CladTenperature

-DECLG(CD=0.>l)

Z000.0a1750.0RCEIiXIiOfACD~0.I350tSIS/R/C/L~iSERIES5C.IZtERCENTTUbE.tLUCCINC blHODELZ~3ZOf0flUIDTEHPERATURE bURST~C~7$fTI1tEAK~7,$0fTl~11500.0IZ50.0I1000.0CLxI750.00~a0.0CIC788'TIHElSEC)8Figure1II.3.2-8a FluidTemperature

-DECLG(CD=0.8)

F000.0IT50.0ACEIlXI1OFACO~0,C350PSIS/R/C/L41SERIESS.C.ItPEREERTTUSEPLUCCINCIIHOOELt3toFOFLUIOTEHPEAATURE OURST~7~00FTI)PEAK~1.50FTI~I~Col500.0IZ50.0~g1000.04xCI~F50.00t50.000.0C)88TIHEISECI8Figure14.3.2-8b FluidTemperature

-DECLG(CD=0.6) 0 2000.0-IT50.0RCKI~XI~OFACOiO.l350PSISIRICIL~lSKRIKSS.C.ItPKRCKMTTUSKPLUCCIKCRIHODKLt,32fDFLUIDTKHPKRATURK BURST,$.00FTI)PKAr.7.50fTI~>4ClJo1500.D1250.0Ea:1000.0Io750.00250.000.0oooanTIHK(SKCICI~nCICICIC7anfaaFiure14.3.2-0c FluidTemperature

-DECLG(CD=0.4) 7000.0NCEOFA12+T.t,350tSLCO<0.11SIlCONFlCVNATlON COXTNOLECCSHOOELSCoO.IOECLCFXr1.12l-FLOVNATE CONESOTTOH1ITOP~tilLIWnI%5000.02500.0~V0.0<500.0-7000.0~4TTHE(SEC)Figure111.3.2-9a CoreFlow(TopandBottom)-DECLG(CD=0.0) 7000.0RCEOfAIt+Tit.350tSICD~0CISIICOIIfICURATION CONTROLECCHODCLSCD>0~COCCLCfXYDI~IZ1-fLOVRATK CORESOTTOH-IITOf~IIILJWVlDl5000.02500.0ICI0.0<500.0A00.0-7000.0CIClTIHEISEC)Figure14.3.2-9b CoreFlow(TopandBottom)-DECLG(CD=0.6)

f000.0RGfOfAl2+T.P.350PSlCDi0~1l90lCONflGURATIOII COIITROLECCSMODELSCD>0.4DECLGfXY>l.l2l-fLOVRATE C3REBOTTOMIITOPiI~I5000.02500.0a0.0-2500.0-5000.0-1000.0TIMEISEC)Figure14.3.2-9c CoreFlow(TopandBottom)-DECLG(CD=0.4)

l.T500RGC)1Xl1DFACD'.0350PSlS/R/C/LSCRlCSS~G.l2PKRCCNTTUIlPLUGGlHGRlHODCL2+320FOFLOODRATCIIN/SEC) l.5000l.2500ul.NNW0.7500CCCIo0.5000O.ZXO0.0ChCICI888TlHC(SEC)Figure1'.3.2-10a RefloodTransient

-CoreInletVelocity-DECK(CD=0.B)

2.0000l.7500ACEI~XllOFACD~0.C)50PSIS/A/C/L4~SEAIESSC.I2PERCENTTUSEPLUGCINGAIHODEL2.)20FOFLOODRATEIIII/SECT I.5000I.2500uI.0000R0.F5004)IOO(y)O0.50000.25000.0CICI8TIHE(SEC)8888Figure14.3.2-10b RefloodTransient

-CoreInletVelocity-DECLQ(CD=0.6) 2.00001.1500RCKI4XIIOFACOz0.a350PSISIRICIL+45fRIKSS.C~12PER(EKffVBLPLUCCIHC81HOOEL2.9XOf0fL000RAffIIH/SECI1.50001.25001.0000i@II-0.1500Cl0.50900.25000.0ClClClClClClCIatIHCISECIClClClFigure14.3.2-10c RefloodTransient

-CoreinletYelocity-DECK(CD=0.)CD=0.4)I

Zo.ooot1+500ACElaXllOFACo~O.l)50tSlS/A/C/LioSEalESs.c.lZrEaCElttulEPLucclkcllHOOELZ.)ZOFOVAtEIILEVELlFt) l5.000lZ.500l0.000Ila.~'.5000+5.ooooZ.50000.088tlHElSECl."i"ureill.3.2-11a BefloodTransient

-CoreandDowncomer llaterLevelsDECLG(CD=0.8)88 Ile500iGEIiXIIOFACD~O.C3SOPSIS/iiCIL4lSERIESS.G.IZPEACENIlUBEPLUGGINGIINOOELZ+320EoUAIEiLEVELlflI I$.000fl.Soo~r.S000I0.000MICO~S.OOOO0.0888lIHEISEC)888I'i'Ure1II.3.2-11b RefloodTransient

-CoreandDowncoloer I.'aterLevels.DECLC(CD=0.6)

Zo.noo.II.%GORGKIIXIIOFA(0-0~lI$0P5I5/R/C/I.S(RIES'S.G.IZPORC(RITUB(PLUGGING8IMOO(tZ.gloFoUGLIERI.EVE.llrll li.000I2.500I0.000I~/.coco~s;00002.50000.0CIOCIC7C)C)IIHEISKGICIC)C7Fi-"ure14.3.2-11c PefloadTransient

-CoreandDmlncomer platerLevelsnCLG(CD=0.4)

F00.0hCEOfAIZ+f.t,350tSICO>O.I1SllCOhfICVhATION COH'lhOLECCSHODELSCD~O+ODECLCfXY>laIRACCVH.fLOllCl>I500.0~JEJl000.0500.00~0.0TIHE(SEC)~VFigure14.3.2-12a Accumulator Flow(Blowdown)

-DECLG(CD=0.8)

%00.0In0.0RCEOFAIZ+T.P.350PSlCOi0~C19I1COIIFlCURA110II COIIlROLECCSHOOELSCO+0~COEFXY+loltACCUHoFLOW~n1500.0IR50.0X~JlJ1000.0F50.000.0CIoFIMEISEC]n.Figure14.3.2-12b Accumulator Flaw(Blawdawn)

-DECLG(CD=0.6) 2000.0I350.0RCEOFAl2+T.P.350PSICOi0~iI9llCOHFIGURhlIOll COHIROLECCSMODELSCO~0.4OECLCI*I~I2ACCVM.FLOM1500.0CSI2SO.OXLJEPl000.0190.00ZSO.OO0.0CI~nTIMEISEC)CICIAJCICICIIrlFigure1II.3.212cAccuAccunulator Flow(Blowdown)

-DECLG(CD=O.II) 8.00007.0000RGE14X14OFACOBRK0.8N2INJECTION 44SERIESS.G.12PERCENTTUBEPLUGING81MO[)ELSAFETYINJECTION FLOW~6.0000.LalI-5.0000~4.0000n-3.000"I-Lal-2.00001.00000.0CDCDnCDnCDCDCDnAJCDCDnCDCDnCDCDCDCDTINE(SEC'ONOS)

Figure1Il.3.2-13a PumpedECCSFlaw(Reflood)-(CD=0.8)-

8.00007.0000RGE14X14OFACOBRK0.6N2INJECTION 44SERIESS.G.12PERCENTTuBEPLuGINGe1MOOELSAFETYINJECTION FLOM~6.0000M~s.oooo~4.0000-3.0000Lsf-2.0000I-LaJ1.00000.0CDCDCDCDCDCDCDCDCDCUCDCDCDCY)CDCDCDCDIlLS3CDCDCDTIME(SECONOS)

Figure14.3.2-13b PunpedECCSFlow(Refload)-(CD=0.6) 8.00007.GOOORGE14X14OFACDBRK0.4N2INJECTION SERIESS.G.12PERCENTTUBEPLUGING81MODELSAFETYINJECTION FLOW-6.0000Laf7-5.0000~4.00003.0000Ld-2.00000-LJ1.00000.0CDCDCDCDCDCDCDCDCDAJCDCDCDCDCDCDCDCDCDCDCDCDTINE(SECONDS)

Figure14.3.2-13c PumpedECCSFlow(Reflood)-(CD=0.4)

50.000P40.000RGE14X14OFACDBRK0.8N2INJECTION 44SERIESS.G.12PERCENTTUBEPLUGING.81 MODELCONTAINMENT PRESSURE30.000-20.00010.0000.0CDCDCDCDCDCDCDCDCDCDCUCDCDCDCDCDCDCDCDCDCDCDCDTIME(SECONDS)

Figure14.3.2-14a Containment Pressure-DECLG(CD=0.8)

50.000RGE14X14OFACOBRK0.6H2INJECTION 44SERIESS.G.12PERCENTTUBEPLVGING81MOOELCONTAINHEHT PRESSURE40.00030.0000.0CDCDCDCDCDCDCDCDCDAJCDCDCDCDCDCDCDCDCD4CDCDTINE(SECONDS)

Fi'ure1~I.3.2-14b Containment Pressure-DECLG(CD=0.6) 50.000RGE14X14OFACDBRK0.4N2INJECTION 44SERIESS.G.12PERCENTTUBEPLUGING81MODELCONTAINMENT PRESSURE40.00030.0004lMICJlCo-20.000C/lE/lLalCL10.0000.0CDCDCDCDCDCDCDCDCDCDCUCDCDCDCDmCDCDC)CD4SCDCDCDCDLATIME(SECONDS)

Fi"urlII.3.2-1IIc Containment Pressure-DECLG(CD=O.II) l.7500RGEOF/IZ+T.l'.350PSICO~0.I1901COIIFICVRATIOII COATAOLECCSHOOEL5C0$0.1OECL'CFXY~I.IZPOVER~,1.50001.25001.00000.75000.5000O.Z5000.0C7OoHC)TIHEl5EC)FigureI4.3.2-15 CorePowerTransient

-DECLG(CD=0.4)

S.OOEK)1RGEOFA12+T.P.lSOPSICOI0.11981CORFIGURATION COKTROLE(CSHOOELSCO*0.iOECLGFXV*I.12BREAKERERGT1.0OEKllII.OCEAN)1 2.0OEK)TI.OOE&7O.OOCIOTJHEISEC)CICICIIIIFigure14.3.2-16 Dreal'nergy ReleasedtoContainment

-DECLG(CD=0.4)

I-LJLa.LaLalCD1250.01000.0RGEIOX10OFACDBRK0.4N2INJECTION 04SERIESS.G.12PERCENTTUBEPLUGING81MODELCONDENSING HEATTRANSFERCOEFFICIENT

~750.00telI-K500.00CDIjlLalCDCD250.000.0CDCDCDCDCDCDCDCDCDCDAJCDCDClCDmCDCDCDCDCDCDCDCATINE(SECONOS)

Figure14.3.2-17 Containment

',lailCondensing IleatTransferCoefficient-DECLG(CD=0.4)

II~IIIN