ML18040B150

Forwards Application for Proposed Amend 39 to License NPF-22,revising Tech Specs to Support Cycle 2 Reload.Fee Paid

ML18040B150
Person / Time
Site:	Susquehanna
Issue date:	06/19/1986
From:	KENYON B D PENNSYLVANIA POWER & LIGHT CO.
To:	ADENSAM E Office of Nuclear Reactor Regulation
Shared Package
ML17146A415	List: ML17146A414 ML17146A416 ML17146A417 ML17146A419 ML17146A421 ML17146A422 ML17146A423 ML18040B150
References
PLA-2661, NUDOCS 8606240301
Download: ML18040B150 (56)

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REQULATYINFORMATION DISTRIBUTIQNYSTEI'l(RIDS)ACCESSIQNNBR:8606240301 DOC.DATE:Sb/06/19NOTARIZED:YESFACIL:50-388Susquehanna SteamElectricStationsUnit2.Pennsglva AUTH.NAt'tEAUTHORAFFILIATION KENYON'.D.Pennsylvania Potoer5LightCo.RECIP.NANERECIPIENT AFFILIATION ADENSAI'1p E.BWRProspectDirectorate 3[)g

SUBJECT:

Forwardsapplication forproposedAmend39toLicenseNPF-22'evising TechSpecstosupportCycle2reload.Feepaid.DISTRIBUTION CODE:ACOIDCOPIESRECEIVED:

LTR8ENCLRSIZE:TITLE:OR,Submittal:

l'eneralDistribution NOTES:icyNNSS/FCAF/PN.

LPDR2cgsTranscripts.

DOCKET¹0500038805000388RECIPIENT IDCODE/NANE BWREBBWRFOBBWRPD3PD01BWRPSB09INTERNAL:

ACRSELD/HDS4NRR/ORASRONDCOPIESLTTRENCLSSlyRECIPIENT IDCODE/NANE BWREICSBBWRPD3LACANPAQNONE BWRRSBADN/LFNBNRR/T/TSCB04COPIESLTTRE'CL2EXTERNAL:

EQ8(QBRUSKE>SNRCPDR02NOTES:33LPDRNSIC03050/>1TOTALNUNBEROFCOPIESREQUIRED:

LTTR33ENCL

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Pennsylvania Power8LightCompanyTwoNorthNinthStreet~Allentown, PA18101~215i770.5151BruceD.KenyonSeniorVicePresident-Nuclear 215/770-41S4 JUN$91986DirectorofNuclearReactorRegulation Attention:

Ms.E.Adensam,ProjectDirectorBWRPro)ectDirectorate No.3DivisionofBWRLicensing U.S.NuclearRegulatory Commission Washington, D.C.20555SUSQUEHANNA STEAMELECTRICSTATIONPROPOSEDAMENDMENT 39TOLICENSENO.NPF-22PLA-2661FILESR41-2,A7-8CDocketNo.50-388

DearMs.Adensam:

ThepurposeofthisletteristoproposechangestotheSusquehanna SESUnit2Technical Specifications insupportoftheensuingCycle2reload.Changestothefollowing Technical Specifications arerequested:

1.03/4.1.23/4.2.13/4.2.23/4.2.33/4.2.43/4.3.4.2 3/4.4.1.1.2 3/4.7.85.3.1B2.1B3/4.1.1B3/4.1.2B3/4.1.3B3/4.1.4B3/4.2.1B3/4.2.2B3/4.2.3B3/4.2.4B3/4.4.1B3/4.7..8IndexDefinitions Reactivity Anomalies AveragePlanarLinearHeatGeneration RateAPRMSetpoints MinimumCriticalPowerRatioLinearHeatGeneration RateEnd-of-Cycle Recirculation PumpTripSystemInstrumentation Recirculation Loops-SingleLoopOperation MainTurbineBypassSystemFuelAssemblies SafetyLimitsShutdownMarginReactivity Anomalies ControlRodsControlRodProgramControlsAveragePlanarLinearHeatGeneration RateAPRMSetpoints MinimumCriticalPowerRatio~e0LinearHeatGeneration RateRecirculat:ion SystemMainTurbineBypassSystem,p>i8606240301 860619PDRADOCK05000388PPDR 1~NKCrt,ptprplt1'I~hIII;IPh~i..'pP.P',t4I'<<PI~14ItI1I~rh44I!II4~4h)4*~.,~$

Page2SSESPLA-2661FilesR41-2,A7-8CMs.E.AdensamAsdiscussed ina'teleconhei'dwithyourstaffonJune16,1986,andintheattached'reload summaryreport,thissubmittal doesnotcontainMinimumCriticalPowerRatio(MCPR)Technical Specification Limits.Themethodology whichwillbeusedtoderivetheselimitsisbeingprovidedatthistimeforyourreview;theactualvalueswillbesuppliedinmid-July.

Thefollowing attachments tothisletterareprovidedtoillustrate andtechnically supporteachofthechanges:Marked-up Technical Specification ChangesNoSignificant HazardsConsiderations Susquehanna SESUnit2Cycle2ReloadSummaryReportXN-NF-86-60, "Susquehanna Unit2Cycle2ReloadAnalysis,"

May,1986XN-NF-86-55, "Susquehanna Unit2Cycle2PlantTransient Analysis,"

May,1986XN-NF-86-65, "Susquehanna LOCA-ECCS AnalysisMAPLHGRResultsfor9X9fuel,"May,1986Susquehanna SESUnit2Cycle2ProposedStartupPhysicsTestsSummaryDescription, May,1986Pleasenotethatwithrespecttothermalhydraulic stability oftheExxonNuclearCompany9X9fuelbeinginsertedduringthisreload,PP&Lhasalreadysubmitted (PLA-2637, datedApril30,1986)astability testprogramwhichwillsupplement theresultspresented inthepertinent analysesattached.

Certainothersupplementary datawillalsobesubmitted totheNRCattheirrequestinaccordance withourdiscussions onMay30,1986.Itisnotourintenttotreatanyofthissupplementary information asrevisions tothisproposal.

Also,sufficient analysishasnotbeencompleted tosupportSingleLoopOperation (SLO)withthe9X9fueldesign.TheTechnical Specifications havebeenalteredaccordingly, andwewillprovideaseparatesubmittal onthisissuebasedonappropriate analysiswhenitisavailable.

Again,thissubmittal isnottobeconsidered arevisiontothisproposedamendment.

Susquehanna SESUnit2iscurrently scheduled tobeshutdownforrefueling andinspection onAugust2,1986andtorestartasearlyasOctober3,1986.Werequestthatyourapprovalbeconditioned tobecomeeffective uponstartupafterthisoutage,andwillkeepyouinformedofanyschedulechanges.

~IhIeeeIe~4)P,l,4'lte4h(h>>CepIe'feehP~'Ie~>"Ita4I4I~II4'"'heII',I'P'I'4It4eVI Page3SSESPLA-2661FilesR41-2,A7-SCMs.E.AdensamAnyquestions withrespecttothisproposedamendment shouldbedirectedtoMr.R.Sgarroat(215)770-7855.

Pursuantto10CFR170, theappropriate feeisenclosed.

Verytrulyyours,B.D.KenySeniorVicePresident-Nuclear Attachments cc:M.J.Campagnone

-USNRCR.H.Jacobs-USNRCT.M.GeruskyBureauofRadiation Protection Pennsylvania Department ofEnvironmental Resources P.O.Box2063Harrisburg, PA17120

(.~p~I~hR'q~t)v'>>.1f't1'elW0t,n XN-NF-86-55 IssueDate:5/]5/86SUSQUEHANNA UNIT2CYCLE2PLANTTRANSIENT ANALYSISPreparedby:T.H.Keheley,TeamLeaderBWRSafetyAnalysisConcur:R.EDollingh,ManagerBWRSafetyAlysisConcur:J.N.Horgan,HagerCustomerServicesEngineering Concur:G.N.Ward,ManagerReloadLicensing Approve:H.E.Williamson, ManagerLicensing

&SafetyEngineering Approve:G.L.Ritter,ManagerfuelEngineering

&Technical Servicesthk/mlnEQONNUCLEARCOMPANY,INC.,860624030

>

NUCLEARREGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICEREGARDING CONTENTSANDUSEOFTHISDOCUMENTPLEASEREADCAREFULLY Thistechnical reportwasderivedthroughresearchanddevelopment programssponsored byExxonNuclearCompany,Inc.Itisbeingsub.mittedbyExxonNucleartotheUSNRCaspartofatechnical contri-butiontofacilitate safetyanalysesbylicensees oftheUSNRCwhichutilizeExxonNudear.fabricated reloadfuelorothertechnical servicesprovidedbyExxonNuclearforlichtwaterpowerreactorsanditistrueandcorrecttothebestofExxonNuclear's knowledge, informaaon, andbegef.Theinformation contained hereinmaybeusedbytheUSNRCinitsreviewofthisreport,andbylicensees orapplicants beforetheUSNRCwhicharecustomers ofExxonNuclearintheirdemonstradon ofcompliance withtheUSNRC'sreguladons.

Withoutderogating fromtheforegoing, neitherExxonNuclearnoranypersonactingnnitsbehalf:A.Makesanywarranty, expressorimplied,withrespecttotheaccuracy, completeness, orusefulness oftheinfor.mationcontained inthisdocument, orthattheuseofanyinformation, apparatus, method,orprocessdisclosed inthisdocumentwillnotinfringeprivately ownedrights;orB.Assumesanyliabilities withrespecttotheuseof,orfordan'agesresulting fromtheuseof,anyinformation, ap.paratus,method,orprocessdisclosed inthisdocument.

XN.NF-FOO,7BB XN-NF-86-55 TABLEOFCONTENTSSectionPacae

1.0INTRODUCTION

..........................................

12.0UHMARY'~~~~~~~~~~~~~~~~~~~~~~o~~o~o~~~~~~~~~~~~~~~~2S3.0TRANSIENT ANALYSISFORTHERMALMARGIN..................

43.1DesignBasist~~~~~~~~~0~~~~~~~~~~~~~~~43.2Anticipated Transients

................................

53.2.13.2.23.2.3LoadRejection WithoutBypass........~~~~~~~~~~5LossofFeedwater Heating.............................

7Feedwater Controller Failure..........................

63.3Calculational Model....................................

83.4afetyLimit..........................................

8S4.0ANALYSESFORINCREASED COREFLOW(ICF)ANDFINALFEEDWATER TEMPERATURE REDUCTION (FFTR)...........

195.0MAXIMUMOVERPRESSURIZATION

............................

225.1D0esignBasis.....................................

225.2Pressurization Transients

.............................

225.3ClosureofAllHainSteamIsolation Valves.............

236.0RECIRCULATION PUMPRUN-UP.."............................

2

47.0REFERENCES

............................................

26APPENDIXA...:.............

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

XN-NF-86-55 ListofTablesTablePacae2.1Transient AnalysisResultsatDesignBasisConditions........................................

33.13.23.33.4ReactorDesignandPlantConditions Susquehanna Unit2...................

Significant Parameter ValuesUsedinSusquehanna Unit2Analysis..........

ResultsofPlantTransient Analysis..FWCFResultsat100%Flow............

~~~~~~~~~~~~~t~~~09~~~~~~~~~~~~~.10~~~~~~~~~~~~~~~~~~13144.14.2ResultsofSystemPlantTransient AnalysisatICFandFFTR................................

20Feedwater Controller FailureDeltaCPRResultsofICFandFFTRAnalyses..............

21

}4~~~

XN-NF-86-55 ListofFiuresFiciurePa(ac3.13.23.33.4LoadRejection WithoutBypass..........................

15LoadRejection WithoutBypass...........16Feedwater Controller Failure......................17Feedwater Controller Failure...........................

186.1ReducedFlowHCPROperating Limit..............

.......25A-3.1A-3.2A-3.3DesignBasisRadialPowerHistogram.....................

A-4DesignBasisLocalPowerDistribution (ENCXN-19x9Fuel)...............................

...A-5DesignBasisLocalPowerDistribution (GESx8Fuel)

XN-NF-86-55

1.0INTRODUCTION

ThisreportpresentstheresultsofExxonNuclearCompany's (ENC's)evaluation ofsystemtransient eventsforSusquehanna Unit,2duringCycle2operation withareloadofENC9x9BWRfuel.Thisevaluation togetherwithanevaluation ofcoretransient eventsdetermines thenecessary thermalmargin(HCPRlimits)toprotectagainsttheoccurrence ofboilingtransition duringthemostlimitinganticipated transient.

Thermalmarginsarecalculated foroperation withintheallowedregionsofthepower/flow operating mapuptothefullpower/full flowoperating condition.

Resulting Thermalmarginanalysesarealsopresented foroperation intheIncreased CoreFlow(ICF)regionofthepower/flow operating mapandforoperation withaFinalFeedwater Temperature Reduction (FFTR).Analysesarealsoreportedforoperation withtheRecirculation PumpTrip(RPT)outofserviceandwiththeturbinebypasscapability inoperable.

Anevaluation isalsomadetodemonstrate thevesselintegrity forthemostlimitingpressurization event.ThebasesfortheseanalyseshavebeenprovidedinReference

l.

l5g XN-NF-86-55

2.0 SUMMARYUsingENCmethodology

andconsidering Cycle2fuels,themostlimitingplantsystemtransient withregardtothermalmarginatratedpowerandflowconditions wasdetermined tobethegenerator LoadRejection WithoutBypass(LRWB).TheMinimumCriticalPowerRatio(HCPR)limitsforpotentially limitingplantsystemtransient eventsareshowninTable2.1forcomparison.

ThevaluesinTable2.1weredetermined assumingboundingconditions intheanalyses.

Thesetransients wereevaluated withallco-resident fueltypesmodeledandthemostlimitingcondition wasusedtodetermine thereportedMCPRs.TheControlRodWithdrawal Error(CRWE)analysisandCycle2HCPRoperating limitarereportedinReference 2.Maximumsystempressurehasbeencalculated forthecontainment isolation event,whichisarapidclosureofallmainsteamisolation valves,usingthescenarioasspecified bytheASMEPressureVesselCode.ThisanalysisshowsthatduringCycle2thesafetyvalvesofSusquehanna Unit2havesufficient capacityandperformance topreventthepressurefromreachingtheestablished transient pressuresafetylimitof110%ofdesignpressure(1.1x1250=1375psig).Theanalysisalsoassumedsixsafetyreliefvalvesoutofservice.Themaximumsystempressures predicted duringtheeventareshowninTable2.1.ResultsforRPToutof'ervice arereportedinSection3.2.1,andresultsforoperation atICFandFFTRarereportedinSection4.

XN-NF-86-55 Table2.1Transient AnalysisResultsatDesignBasisConditions*

Transient CPRMCPRENC9x9GEBx8LoadRejection WithoutBypass0.17/1.23 0.16/1.22 Feedwater Controller FailureLossofFeedwater Heating0.15/1.21 NA/1.140.14/1.20 NA/1.14MaximumPressuresiTransient VesselDomeVesselLowerPlenumSteamLineMSIVClosure130113151305*104%power/100%

flow.**BasedonasafetylimitMCPRof1.06.

XN-NF-86-55

3.0 TRANSIENT

ANALYSISFORTHERMALMARGIN3.1DesinBasisConsistent withtheFSARplanttransient

analysis, thermalmarginoperating MCPRlimitsaredetermined basedonthe104%power/100%

flowoperating point.Thisthermalmarginoperating MCPRlimitisthenmodifiedasafunctionofpowerandflowasrequiredtoprotectagainstboilingtransition resulting fromtransients occurring fromallowedconditions onthepower/flow operating map.Theplantconditions forthe104%power/100%

flowpointareasshowninTable3.1.ThemostlimitingpointinCycle2hasbeendetermined tobeatendoffullpowercapability whencontrolrodsarefullywithdrawn fromthecore.Thethermalmarginlimitestablished forendoffullpowerconditions isconservative forcaseswherecontrolrodsarepartially inserted.

Follow-ingrequirements established inthePlant.Operating Licenseandassociated Technical Specifications, observance ofaMCPRlimitof1.23for9x9fueland1.22for8x8fuelorgreaterconservatively protectsagainstboilingtransi-tionduringanticipated plantsystemstransients fromdesignbasisconditions forSusquehanna Unit2Cycle2.Thecalculational modelsusedtodetermine thermalmarginincludeENC'splanttransient andcorethermal-hydraulic codesasdescribed inprevious(1,4-7)documentation

'Fuelpellet-to-clad gapconductances usedintheanalysesarebasedoncalculations withRODEX2).Table3.2summarizes thevaluesusedforimportant parameters toprovideaboundinganalysis.

Recirculation PumpTrip(RPT)coastdown wasinputbasedonmeasuredSusquehanna Unit2startuptestdata.Toconfirmtheneutronics as'required bytheSERissuedforthesupplements ofReference ItheSusquehanna systemtransient modelwasbenchmarked toappropriate Susquehanna Unit2startuptestdata.Alltransients wereanalyzedonaboundingbasisusingtheCOTRANSAhotchanneldeltaCPRmodelasdescribed inReference

9.

XN-NF-86-55 3.2nticiatedTransients ENCconsiders eightcategories ofpotential systemtransient occurrences forJetPumpBWRsinXN-NF-79-71(

'.Thelossoffeedwater heating(g)transient hasbeenanalyzedonagenericbasisasreportedinReference 10.Resultsshownforthistransient arefromtheENCgenericanalysis.

Thetwomostlimitingtransients aredescribed herein.detailtoshowthethermalmarginforCycle2ofSusquehanna Unit2.Thesetransients are:LoadRejection WithoutBypass(LRWB)Feedwater Controller Failure(FWCF)Asummaryofthetransient analysesisshowninTable3.3.Otherplanttransient eventsareinherently nonlimiting orclearlyboundedbyoneoftheaboveevents.3.2.1LoadRejection WithoutBypassThiseventisthemostlimitingoftheclassoftransients characterized byrapidvesselpressurization.

Thegenerator loadrejection causesaturbinecontrolvalvetrip,whichinitiates areactorscramandRPT.Thecompression waveproducedbythefastcontrolvalveclosuretravelsthroughthesteamlinesintothevesselandcreatesthevesselpressurization.

Turbinebypassflow,whichcouldmitigatethepressurization effect,isnotallowed.Theexcursion ofcorepowerduetovoidcollapseisprimarily terminated byreactorscramandvoidgrowthduetoRPT.Figures3.1and3.2depictthetimevarianceofcriticalreactorandplantparameters duringtheloadrejection transient calculation withboundingassumptions.

Theboundingassumptions areconsistent withENC'sCOTRANSAcodeuncertainties analysismethodology asreportedinXN-NF-79-71(P)

Rev.2,Supplements 1-3andapprovedbyNRC.Theboundingassumptions include:

XN-NF-86-55 Technical Specification minimumcontrolrodspeedTechnical Specification maximumscramdelaytimeintegralpowerincreased by10%Atdesignbasisconditions (104%power/100%

flow)thisresultsinadeltaCPRof0.17fortheloadrejection withoutbypasswhenRPTisoperableforENC9x9fuel.Thecorresponding deltaCPRforGE8x8fuelis0.16.Theloadrejection wasthenanalyzedassumingthesameboundingconditions but.withbothRPTand.bypassinoperable.

ThisresultedindeltaCPRsof0.31forbothENC9x9andGE8x8fuel.3.2.2Feedwater Controller FailureFailureofthefeedwater controlsystemispostulated toleadtoamaximumincreaseinfeedwater flowintothevessel.Astheexcessive feedwater flowsubcoolstherecirculating waterreturning tothereactorcore,thecorepowerwillriseandattainanewequilibrium ifnootheractionistaken.Eventually, theinventory ofwaterinthedowncomer willriseuntilthehighlevelvesseltripsettingisexceeded.

Toprotectagainstspillover ofsubcooled watertotheturbine,theturbinetrips,closingtheturbinestopvalvesandinitiating areactorscram.Thecompression wavethatiscreated,thoughmitigated bybypassflow,pressurizes thecoreandcausesapowerexcursion.

Thepowerincreaseisterminated byreactorscram,RPT,andpressurerelieffromthebypassvalvesopening.Theevaluation ofthefloweventatdesignbasisconditions wasperformed withboundingvaluesandresultedinadeltaCPRof0.15forENC9x9fueland0.14forGE8x8fuel.Figures3.3and3.4presentkeyvariables forthisfeedwater controller failureevent.Thiseventwasalsoexaminedforreducedpowerconditions atfullflow.TheresultsfortheFWCFtransients XN-NF-86-55 fromreducedpowerconditions areshowninTable3.4.Thecalculated resultsshowthatFWCFdeltaCPRsvarywithdecreasing poweratfullflowconditions.

ThehighestdeltaCPRswerecalculated atthe40%power/100%

flowconditions.

Thistransient eventatfullpowerandfullflowconditions wasalsoanalyzedassumingboundingconditions andfailureofthebypassvalvestoopen.Thisresulted'inadeltaCPRof0.18forENC9x9fueland0.17forGE8x8fuel.3.2.3LossofFeedwater HeatingThelossoffeedwater heatingleadstoagradualincreaseinthesubcooling ofthewaterinthereactorlowerplenum.Reactorpowerslowlyrisestothethermalpowermonitorsystemtripsetpoint.

Thegradualpowerchangeallowsfuelthermalresponsetomaintainpacewiththeincreaseinneutronflux.ENChasanalyzedthelossoffeedwater heatingeventonagenericbasisasdescribed inReference 10.BasedonthegenericanalysisandtheCycle2safetylimitof1.06,theMCPRlimit.forSusquehanna Unit2Cycle2willbe1.14forboththeENC9x9fuelandtheGE8x8fuelforthelossoffeedwater heatingevent.Thebypassvalvesdo'otsignificantly affectthelossoffeedwater

.heatingresults.Thus,thisMCPRlimitisapplicable whetherthebypassvalvesareoperableornot.3.3Calculational ModelTheplanttransient codeusedtoevaluatethegenerator loadrejection andfeedwater flowincreasewasENC'scodeCOTRANSA~

~.Theaxialone-dimensional neutronics modelpredicted reactorpowershiftstowardthecoremiddleandtopaspressurization occurred.

Thiswasaccounted forexplicitly indetermining thermalmarginchangesinthetransient.

Thelossoffeedwater heatingevent XN-NF-86-55 wasevaluated generically becauserapidpressurization andvoidcollapsedonotoccurinthisevent.AppendixAoftheSusquehanna Unit1Cycle2analysisdelineates thechangesmadetoCOTRANSAtomergethePTSBWR3codewiththeCOTRANSAcode,torefinenumerical techniques andtoimproveinput.AppendixAofReferen'ce 9describes therefinement madetothehotchannelmodeltocalculate thedeltaCPR'sduringthetransient.

AppendixBofReference 3delineates theplantrelatedchangesmadetothesecodesfortheSusquehanna Units1and2analyses.

3.4SafetLimitThesafetylimitistheminimumvalueofthecriticalpowerratio(CPR)atwhichthefuelcouldbeoperatedwheretheexpectednumberofrodsinboilingtransition wouldnotexceed0.1%ofthefuelrodsinthecore.ThesafetylimitistheHCPRwhichwouldbepermitted tooccurduringthelimitinganticipated operational occurrenc'e.

Thesafetylimitforallfueltypesin'usquehanna Unit2Cycle2wasdetermined bythemethodology presented inReference 4tohaveavalueof1.06.Theinputparameters anduncertainties usedtoestablish thesafetylimitarepresented inAppendixAofthisreport.

XN-NF-86-55 Table3.1ReactorDesignandPlantCo'nditions Susquehanna Unit2ReactorThermalPower(104%)TotalCoreFlow(100%)CoreIn-Channel FlowCoreBypassFlowCoreInletEnthalpyVesselPressures SteamDomeUpperPlenumCoreLowerPlenumTurbinePressureFeedwater/Steam FlowFeedwater EnthalpyRecirculation PumpFlow(perpump)3439HWt100.0Hlb/hr89.7Mlb/hr10.3Hlb/hr518.0Btu/ibm1031psia1049psia1058psia1067psia974.7psia14.15Hlb/hr360.8Btu/ibm15.7Hlb/hr 10XN-NF-86-55 Table3.2Significant Parameter ValuesUsedinAnalysisSusquehanna Unit2HighNeutronFluxTripControlRodInsertion TimeControlRodWorthVoidReactivity FeedbackTimetoDeenergized PilotScramSolenoidValvesTimetoSenseFastTurbineControlValveClosureTimefromHighNeutronFluxTimetoControlRodNotionTurbineStopValveStrokeTimeTurbineStopValvePositionTripTurbineControlValveStrokeTime(Total)Fuel/Cladding GapConductance CoreAverage(Constant)

Safety/Relief ValvePerformance SettingsReliefValveCapacityPilotOperatedValveDelay/Stroke 125.3%3.5sec/90%insertednominalnominal200msec(maximum) 30msec290msec100msec90%open70msec443.8Btu/hr-ft2-F Technical Specifications 225.4ibm/sec(1110psig)400/150msec XN-NF-86-55 Table3.2Significant Parameter ValuesUsedinAnalysis(Cont.)Susquehanna Unit2HSIVStrokeTimeHSIVPositionTripSetpointTurbineBypassValvePerformance TotalCapacityDelaytoOpening(80%open)FractionofEnergyGenerated inFuelVesselWaterLevel(aboveSeparator Skirt)HighLevelTripNormalLowLevel.TripHaximumFeedwater RunoutFlowThreePumpsRecirculation PumpTripSetpoint3.0sec90%open936.11ibm/sec300msec0.96558.7in36.5in8in4118ibm/sec1170psigVesselPressure 12XN-NF-86-55 Table3.2Significant Parameter ValuesUsedinAnalysis(Cont.)Susquehanna Unit2ControlCharacteristics SensorTimeConstants PressureOthersFeedwater ControlModeFeedwater MasterController Proportional GainResetRateFeedwater 100%MismatchWaterLevelErrorSteamFlowEquiv.FlowControlModePressureRegulator Settings'eadLagGain500msec250msecThree-Element 50.0(%/%)(%/ft)1.70(%/sec/ft) 48in100%Manual3.0sec7.0sec3.33%/psid 13XN-NF-86-55 Table3.3ResultsofSystemPlantTransient AnalysesEventMaximumNeutronFlux%RatedMaximumMaximumCoreAverageSystemHeatFluxPressure%Rated~sia6CPRLoadRejection Without8ypass274114.31213.17Feedwater Controller Failure245114.71180.15MSIVClosurewithFluxScram368130.71330Note:Alleventsareboundingcaseat104%power/100%

flow.

14XN-NF-86-55 Table3.4Feedwater Controller FailureAnalysisResultsat100%Flow%PowerDeltaCPRCESx8ENC9x9100.14.15.80.22.2465.23.2540.26.29 30252HEA3.RECVESFLUXRCULATIELSTENFLOWFLOW20124512~g350'8.00.20.50'1'1'1'1'TINE.SEC2.02.22'Figure3.1LoadRejection WithoutBypass 172.VESELWATLEVEL(TN)12106COlr~<g750)OlEA5VlQJ54cLR500)Ol))250'0.20.50.71'1'1'TINE.SEC1~72'2'2'Figure3.2LoadRejection WithoutBypass 30252HEAFLUX3.RECRCULATIHFLOW4.VESELSTEFLOW20~15o105012820TINE.SECFigure3.3Feedwater Contro1ler fai1ure28 2.VESELMhTLEVEL(IN)121080rQlQlQla60Qlrt$0I00Ialal40~~N2010p12162024TIME.SEC2836>CIICOCJlICJl4PcJlFigure3.4Feedwater Controller Failure l'

19XN-NF-86-55

4.0 ANALYSESWITHINCREASED

COREFLOWICFANDFINALFEEDWATER TEMPERATURE REDUCTION FFTRAspartoftheSusquehanna Unit2licensing

analysis, ENCevaluated transients foroperation intheIncreased CoreFlow(ICF)operating regionupto108%ofratedflow.Transient analyseswerealsoperformed forafeedwater temperature reduction ofupto65degreesFatbothnominalflowandincreased coreflowconditions attheendoftheoperating cycle.This65degreeFtemperature reduction wasconservatively heldconstantatallpowerlevelsevaluated.

Asummaryofthetransient analysesisshowninTable4.1.Comparison oftheresultsinTable2.1and4.1indicatethatICFhadnosignificant effectontheLRWBdeltaCPRresultsandFFTRcondition slightlyreducedtheimpactofthisdocument.

Thecorresponding maximumoverpressurization eventisdiscussed inSection5.0andthepumprun-upanalysisisreportedinSection6.0.Theeffectsofthefinalfeedwater temperature reduction wereevaluated byanalyzing theFWCFtransient overtheallowedpowerrangeforbothnominalfeedwater temperature anda65degreeFfinalfeedwater temperature reduction.

Calculations wereperformed forboththe100%coreflowandforthe108%coreflowconditions.

Theresultsofthesecalculations areshowninTable4.2.Thecalculated FWCFtransient deltaCPRgenerally increases withdecreasing poweratbothflowconditions, andanincreased MCPRlimitisindicated forlowpoweroperating conditions.

Thus,forincreased coreflowoperation, increased MCPRlimitsareindicated.

Afurther,butsmall,deltaCPRincreaseisgenerally indicated tooperatewithreducedfeedwater temperature forbothratedcoreflowandincreased coreflow.

20XN-NF-86-55 Table4.1ResultsofSystemPlantTransient AnalysisatICFandatFFTRLoadRe'ection WithoutBass104/100(FFTR)104/108104/108(FFTR)MaximumNeutronic Flux(%rated)253241222MinimumCoreAverage(%rated)112.9112.1110.8MaximumSystemPressure(psia)119112101187DeltaCPR0.150.170.15ASMEOverressureMSIVClosuresiVesselDomeVesselLowerPlenumSteamLine104/100(FFTR)104/108104/108(FFTR)1264'12901257127913071274126512961259 21XN-NF-86-55 Table4.2Feedwater Controller FailureDeltaCPRResultsofICFandFFTRAnalysesNominalFeedwater Tem.FFTR~/'E,ENC9x9GE8XSENC9X9100/10080/10065/10040/100100/10880/10865/10840/1080.140.220.230.260.150.200.230.270.150.240.250.290.160.220.250.300.160.200.240.260.160.200.240.260.170.220.260.290.170.220.260.30*65'Freduction inFeedwater Temperature.

lI 22XN-NF-86-55

5.0 MAXIMUMOVERPRESSURIZATION

Maximumsystempressurehasbeencalculated forthecontainment isolation event(rapidclosureofallmainsteamisolation valves)withanadversescenarioasspecified bytheASHEPressureVesselCode.ThisanalysisshowedthatthesafetyvalvesofSusquehanna Unit2havesufficient capacityandperformance topreventpressurefromreachingtheestablished transient pressuresafetylimitof110%ofthedesignpressure.

Themaximumsystempressures predicted duringtheeventareshowninTable2.1.Thisanalysisalsoassumedsixsafetyreliefvalvesoutofservice.5.1DesinBasisThereactorconditions usedintheevaluation ofthemaximumpressurization eventarethoseshowninTable3.1.Themostcriticalactivecomponent (scramonHSIVclosure)wasassumedtofailduringthetransient.

Thecalculation wasperformed withENC'sadvancedplantsimulator codeCOTRANSA,whichincludesanaxialone-dimensional neutronics model.5.2Pressurization Transients ENChasevaluated severalpressurization eventsandhasdetermined thatclosureofallHainSteamIsolation Valves(MSIVs)withoutdirectscramisthemostlimiting.

ThoughtheclosurerateoftheHSIVsissubstantially slowerthantheturbinestopvalvesorturbinecontrolvalves,thecompressibility oftheadditional fluidinthesteamlinescausestheseverityofthesefasterclosurestobeless.Essentially, therateofsteamvelocityreduction isconcentrated towardtheendofthevalvestroke,generating asubstantial compression wave.Oncethecontainment isisolatedthesubsequent corepowerproduction mustbeabsorbedinasmallervolumethanifaturbineisolation hadoccurred.

Calculations havedetermined thattheoverallresultistocauseisolation (MSIVclosures) tobemorelimitingforsystempressurethanturbineisolations.

23XN-NF-86-55 5.3ClosureofAllMainSteamIsolation ValvesThiscalculation assumedthatsixreliefvalveswereoutofserviceandthatallfoursteamisolation valveswereisolatedatthecontainment boundarywithin3seconds.Atabout5.5seconds,thereactorscramisinitiated byreaching'the highfluxtripsetpoints.

Sincescramperformance wasdegradedtoitsTechnical Specification limit,effective powershutdownisdelayeduntilafter7.1seconds.Substantial thermalpowerproduction enhancespressurization.

Pressures reachtherecirculation pumptripsetpoint(1170psig)beforethepressurization hasbeenreversed.

Lossofcoolantflowleadstoenhancedsteamproduction aslesssubcooled waterisavailable to'absorbcorethermalpower.Themaximumpressurecalculated inthesteamlineswas1305psigoccurring nearthevesselatabout10.1seconds.Themaximumvesselpressurewas1315psigoccurring inthelowerplenumatabout10.0seconds.TheanalysiswasrepeatedforICFandFFTRconditions andtheresultsaresummarized inTable4.1.Compaison oftheresultsinTable2.1andTable4.1showthatthedesignbasisconditions aremorelimitingthanICForFFTRconditions.

Atabout5.5seconds,thereactorscramisinitiated byreachingthehighfluxtripsetpoints.

Sincescramperformance wasdegradedtoitsTechnical Specification limit,effective powershutdownisdelayeduntilafter6.5seconds.Substantial thermalpowerproduction enhancespressuriza-tion.Pressures reachtherecirculation pumptripsetpoint(1170psig)beforethepressurization hasbeenreversed.

Lossofcoolantflowleadstoenhancedsteamproduction aslesssubcooled waterisavailable toabsorbcorethermalpower.Themaximumpressurecalculated inthesteamlineswas1296psigoccurring nearthevesselatabout10.2seconds.Themaximumvesselpressurewas1307psigoccurring inthelowerplenumatabout9.8seconds.

XN-NF-86-55

6.0 RECIRCULATION

PUMPRUN-UPAnalysisofpumprun-upeventsforoperation atlessthanratedrecirculation pumpcapacitydemonstrates theneedforanaugmentation ofthefullflowHCPRoperating limitforlowerflowconditions.

Thisisduetothepotential forlargereactorpowerincreases shouldanuncontrolled pumpflowincreaseoccur.Thissectiondiscusses pumpexcursions whentheplantisinmanualflowcontroloperation mode.Basedontheresultsobtainedfrompreviousanalyseswhichshowedtwopumpexcursions werethelimitingpumprun-upevent,onlytwopumpexcursions areevaluated forSusquehanna Unit2Cycle2.TheseresultsindicatethatMCPRwoulddecreasebelowthesafetylimitifthefullflowreference MCPRwasobservedatinitialconditions.

Thus,anaugmented HCPRisneededforpartialflowoperation toprotectthetwopumpexcursion event.Theevaluation ofthetworecirculation pumpflowexcursion forSusquehanna Unit2showedthatestablishment ofHCPRlimitsforthiseventwhichpreventsboilingtransition willalsoboundsinglepumprunups.Theanalysisofthetwopumpflowexcursion indicates thatthelimitingeventscenarioisagradualquasi-steady run-upduetotheinletenthalpylagassociated withamorerapidrun-up.TheSusquehanna Unit2Cycle2analysisconservatively assumedtherun-upeventinitiated at57%power/40%

flowandreached111%ratedpowerat110%ratedflow.110%flowisconsistent withincreased coreflowanalysis; Thispowertoflowrelationship boundsthatcalculated byXTGBWRfortheconstantXenonassumption.

Theresultsofthetwopumprun-upanalysesformanual,flowcontrolarepresented inFigure6.1.ThecyclespecificHCPRlimitforSusquehanna Unit2Cycle2shallbethemaximumofthereducedflowMCPRoperating limitandthefullflowHCPRoperating limit.

1.4R1.3KCl1.21.11.04Figure6.1TotalCoreRecirculating Flow(IRated)ReducedFlowMdPROperating Limitl,OCITlICX)ChICJlCJl 26XN-NF-86-55

7.0REFERENCES

2.3.5.6.7.8.9.10.R.H.Kelley,"ExxonNuclearPlantTransient Methodology forBoilingWRt,"X~,R1*12(ppid),ENuclearCo.,Inc.,Richland, WA99352,November1981.T.H.Keheley,"Susquehanna Unit2Cycle2ReloadAnalysis, DesignandSafetyAnalyses,"

XN-NF-86-60, ExxonNuclearCo.,Inc.,Richland, WA99352,April1986.T.H.Keheley,"Susquehanna Unit1Cycle2PlantTransient Analyses,"

XN-NF-84-118 including Supplement 1,ExxonNuclearCompany,Richland, WA99352,December1984.T.L.Krysinski andJ.C.Chandler, "ExxonNuclearMethodology forBoilingWaterReactors; THERMEXThermalLimitsMethodology; SummaryPIi,"~,EE,R 11(,ENIC.,Inc.,Richland, WA99352,April1981.T.W.Patten,"ExxonNuclearCriticalPowerMethodology forBoilingWR,"X~,11,ENI2Richland, WA99352,November1979.R.H.Kelley,"DresdenUnit3Cycle8PlantTransient Analysis2,"('--,III,ENI.,I.,Rihld,IIA99352,December1981.R.H.KelleyandN.F.Fausz,"PlantTransient AnalysisforDresden2,1,"~X---,(21.,1.,(tihi d,llA99352,October1982.K.R.Merckx,"RODEX2FuelRodMechanical ResponseEvaluation Model,"~X---,RI1,EIII.,I.,IWhld,IIA99352,March1984.T.H.Keheley,"Susquehanna Unit1Cycle3PlantTransient Analysis,"

XN-NF-85-130, ExxonNuclearCompany,Richland, WA99352,November1985.R.G.Grummer,"AGenericLossofFeedwater HeatingTransient ForW<<,"~X-->>,dI,,Ihid,WA99352,February1986.

E A-IXN-NF-86-55 APPENDIXAHCPRSAFETYLIHITA.lINTRODUCTION TheHCPRfuelcladdingintegrity safetylimitwascalculated usingthemethodology anduncertainties described inReference A.l.Inthismethodology, aHonteCarloprocedure isusedtoevaluateplantmeasurement andpowerpredictions uncertainties suchthatduringsustained operation attheHCPRCladdingIntegrity SafetyLimit,atleast99.9%ofthefuelrodsinthecorewouldbeexpectedtoavoidboilingtransition.

Thisappendixdescribes thecalculation andpresentstheanalytical results A-2XN-NF-86-55 A.2CONCLUSIONS Duringsustained operation ataHCPRof1.06withthedesignbasispowerdistribution described below,atleast99.9%ofthefuelrodsinthecoreareexpectedtoavoidboilingtransition ataconfidence levelof95%.

A-3XN-NF-86-55

'.3DESIGNBASISPOWERDISTRIBUTION Predicted powerdistributions wereextracted fromthefuelmanagement analysisforSusquehanna Unit2Cycle2.Theseradialpowerdistributions wereevaluated forperformance asthedesignbasisradialpowermap,andthedistribution at10,500MWD/HTcycleexposurewasselectedasthemostsevereexpecteddistribution forthecycle.Thedistribution wasskewedtowardhigherpowerfactorsbytheadditionofbundleswitharadialpeakingfactorapproximating anoperating HCPRlevelof1.26atfullpower.Theresulting designbasisradialpowerdistribution isshowninFigureA.3-1.Thefuelmanagement analysisindicated thatthemaximumpowerENCbundleinthecoreatthisstatepoint waspredicted tobeoperating atanexposurelevelof12,600HWD/HT,soalocalpowerdistribution typicalofanodalexposureofl5,000MWD/MT'as selectedasthedesignbasislocalpowerdistribution.

Thisdistribution isshowninFigureA.3-2.Aboundingly flatlocalpowerdistribution wasselectedfortheco-resident G.E.Fuel.Thisdistribution isshowninFigureA.3-3.Becausethepredicted powerdistributions duringthecyclewerenotallcharacterized bybottompeakedaxialdistributions, representative safetylimitevaluations wereperformed atseveralrepresentative cycleburnupstatepoints throughout thecycle,including allpointsatwhichthepowerwasskewedtowardtheupperhalfofthecore.Theseanalysesconfirmed thethatmostseverepowerdistribution conditions werethosewhicharepredicted toexistattheendofCycle2.The1.06safetylimitwasconfirmed atallthepointsevaluated.

90807060V)SOC)403020100.20.40.60.81.2RRDIRLPERKINGF'RCTORFigureA.3-lDesignBasisRadialPowerHistogram A-5XN-NF-86-55

0.97:1.01:0.97:1.04:1.04:1.05:0.97:1.02:0.97:1.01:0.94:0.97:1.07:1.06:0.95:1.00:0.95:1.020.97:0.99:1.04:1.05:1.05:1.02:1.06:1.00:0.971,04:0.93:1.05:1.01:0.97:0.00:1.02:0'5:1.05:1.03:1.05:1.03:1.00:0:00:0.97:1.05:1.06:1.04:1.04:0.94:1.04:1.00:1.00:1.01:1.05:1.07:1.04::0'7:0.98:0.90:1.04:1.03:1.05:1.04:0.97:0.97::0.91:0.94:0.98:0.94:1.05:0.93:0.99:0.94:1.01:0.88:0.91:0.97:1.04:1.03:1.04:0'7:1.01:0.97:FiaureA.3-2DESIGNBASISLOCALPOWERDISTRIBUTION ENCXN-19X9FUEL*Rodadjacenttocontrolbladecornerlocation A-6XN-Nf-86-55 1.03:1.00:0.99:0,99:0.99:0.99:1.00:1.031.00:0.99:1.03:1.02:0.99:0.99:0,97:1.000.99:1.03:0.91:1'2:1,01:0.98:0.99:0.990.991.031.020.00:1.02:1.01:0.99:0.990.99:1.02:1.01:0.91:0.00:1.02:1.02:0.9900.99:0.99:1.02:1.01:1.02:0.91:1.03:0.990~~~41.00:0.970.99:1.02:1.03:1.03:0.99:1.0001.03:1.00:0.99:0.99:0.99:0.99:1.00:1.03~*FigureA.3-3DESIGNBASISLOCALPOWERDISTRIBUTION G.E.8X8FUEL*Rodadjacenttocontrolbladecornerlocation A-7XN-NF-86-55 A.4CALCULATION OFTHENUHBEROFRODSINBOILINGTRANSITION TheSAFTLIHcomputercodewasusedtoanalyzethenumberoffuelrodsinboilingtransition.

TheXN-3correlation wasusedtopredictcriticalheatfluxphenomena.

FivehundredHonteCarlotrialswereperformed tosupporttheHCPRsafetylimit.Non-parametric tolerance limitswereusedinlieuofPearsoncurvefitting.Theuncertainties usedintheanalysisfornormalconditions werethoseidentified inReference A-1.Atleast99.9%ofthefuelrodsinthecorewereexpectedtoavoidboilingtransition withaconfidence levelof95%.'

A-8XN-NF-86-55 A.5A-l.A-2.A-3.REFERENCES "ExxonNuclearCriticalPowerMethodology forBoilingWaterR",R111,~X--,XN1CPRichland, WA(November 1983)."TRXN-1111CExxonNuclearCompany,Richland, WA(March1981).PaulN.Somerville, "TablesforObtaining Non-Parametric Tolerance Limits",AnnalsofMathematical Statistics, Vol.29,No.2(June1958),pp.599-601.

XN-NF-86-55 IssueDate:5/15/86SUS(UEHANNA UNIT2CYCLE2PLANTTRANSIENT ANALYSISDistribution D.J.J.C.R.E.S.F.R.G.K.D.S.E.T.H.J.E.T.L.J.N.L.A.T.W.G.L.H.G.D.R.G.N.H.E.BraunChandlerCollingham GainesGrummerHartleyJensenKeheleyKrajicekKrysinski MorganNielsonPattenRitterShaw/PP&L (40)SwopeWardWilliamson DocumentControl(5)

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