ML17146B094
| ML17146B094 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/30/1987 |
| From: | WHITE J A ADVANCED MEDICAL SYSTEMS, INC. |
| To: | |
| Shared Package | |
| ML17146B090 | List: |
| References | |
| ANF-87-125, ANF-87-125-R01, ANF-87-125-R1, NUDOCS 8712310156 | |
| Download: ML17146B094 (60) | |
Text
ANF-87-125REVISIONAQMHCSDoHAIL,EARPUDDLESCORPORATIONSUSQUEHANNAUNIT2CYCLE3PLANTTRANSIENTANALYSISNOVEMBER198787i23iOi5687i'223PDRADOCK05000389',',P.....,,...,,~f'PgANAFFILIATEOFKRAFTIVERKUPIIONQ~KMfU ADVANCEDNUCLEARFUELSCORPORATIONANF-87-125Revision1IssueOate:11iIOi87SUSQUEHANNAUNIT2CYCLE3PLANTTRANSIENTANALYSISPreparedBy:J.A.WhiteBWRSafetyAnalysisLicensingandSafetyEngineeringFuelEngineeringandTechnicalServicesAHAFFIUATEOFKRAFTWFRKVHIOHQxwu CUSTOMERDISCLAIMERIMPORTANTNOTICEREGARDINGCONTENTSANDUSEOFTHISDOCUMENTPLEASEREADCAREFULLYAdvancedNuclearFuelsCorporation'swarrantiesandrepresentationscon-cemingthesubjectmatterofthisdocumentarethosesetforthIntheAgreementbetweenAdvancedNuclearFuelsCorporationandtheCustomerpursuanttowhichthisdocumentlsIssued.Accordingly,exceptasotherwiseexpresslypro-videdInsuchAgreement,neitherAdvancedNuclearFuelsCorporationnoranypersonactingonitsbehalfmakesanywarrantyorrepresentation,expressedorimplied.withrespecttotheaccuracy,completeness,orusefulnessoftheinfor-mationcontainedinthisdocument,orthattheuseofanyinformation,apparatus,methodorprocessdisclosedinthisdocumentwillnotinfringeprivatelyownedrights;orassumesanyliabilitieswithrespecttotheuseofanyinformation,ap-paratus,methodorprocessdisclosedinthisdocument.TheinformationcontainedhereinIsforthesoleuseofCustomer.Inordertoavoidimpairmentofr/ghtsofAdvancedNuclearFuelsCorporationinpatentsorinventionswhichmaybeincludedintheinformationcontainedInthisdocument.therecipient,byitsacceptanceofthisdocument,agreesnottopublishormakepublicusegnthepatentuseoftheterm)ofsuchinformationuntilsoauthorizedinwritingbyAdvancedNuclearFuelsCorporationoruntilaftersix(6)monthsfollowingterminationorexpirationoftheaforesaidAgreementandanyextensionthereof,unlessotherwiseexpresslyprovidedintheAgreement.Norightsorlicensesinortoanypatentsareimpliedbythefurnishingofthisdocu-ment.XN.NF.F00-765(1/Bi ANF-87-125Revision1TABLEOFCONTENTSSection~Pae
1.0INTRODUCTION
~t~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~~~~t~~~~~~~~~~~~~~~~~12.0SUMMARYo~~~~~~~~~~~~~~o~~~~~~~~~~~~~~~~~t~~~~~~~~~~~~~~~~~~~~~~~~~23.0TRANSIENTANALYSISFORTHERMALMARGIN..............................53.13.23.2.13.2.23.2'3.3DesignBasss.......................AnticipatedTransients.............LoadRejectionWithoutBypass......FeedwaterControllerFailure.......LossOfFeedwaterHeating..........Calculational,Model................1~~~~~~~~~~~~~I~~~t~~~~~~~~t~~~~5~~~~~~~~~t~\~~~~~~~~~~~~~~~~~~~6~~~~t~~~~~~~~~~~~~~~~~~~~~~~~~~6~~~~~~~~~~~~~~~~~~~t~~~7~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~t~~~~~~~~~~83.4SafetyLimit.........~~~~~~~~~~~~~I~~~4~~~~~~~~~~~~~~~~~~~9MAXIMUMOVERPRESSURIZATION...................................22D0esignBases...................................................224.2PressurizationTransients.~...............~~~~~224.2.1ClosureOfAllHainSteamIsolationValves.......,.235.0RECIRCULATIONPUMPRUN-UP.........................................2
46.0REFERENCES
.............~~~~~~~~t~~~~~~~~t~~~~~~~~~~~~~~~~~26APPENDICESA.SINGLELOOPOPERATION..........,.......'..........................A-1B.HCPRSAFETYLIMIT........................................B-l gt
~~11ANF-87-125Revision1LISTOFTABLESTable2.13.13.23.33.4A.lTransientAnalysisResultsAtDesignBasisConditions...ReactorDesignAndPlantCondi.tionsSusquehannaUnit2..otSignificantParameterValuesUsedInTheAnalysisForSusquehannaUnit2......................................ResultsOfSystemPlantTransientAnalyses..............FeedwaterControllerFailureAnalysisResultsAt100%FlSLOReactorAndPlantConditions...:....................~Pae~~~~~~~~~~4~~o~~~~~~~10o~~~~~~~~~llo~~~oo~~~~14owo~~~~~~o15~~~~~~~~~~LISTOFFIGURESFiciur'e3.13.23.33.4LoadRejectionWithoutBypass.............................LoadRejectionWithoutBypass..........................FeedwaterControllerFailure.........FeedwaterControllerFailure......................................~Pae161718193.53.65.1A.lA.2A.3B.3-1B.3-2B.3-3LossOfFeedwaterHeating..~...........................LossOfFeedwaterHeating..................SusquehannaUnit2Cycle3ReducedFlowMCPROperatingSingleLoopOperation-PumpSeizure.......SingleLoopOperation-PumpSeizure....CorePowerVersusCoreFlow..........................SusquehannaUnit2Cycle3DesignBasisRadialPowerHiDesignBasisLocalPowerDistributionAdvancedNuclearFuelsXN-29x9Fuel...............,.DesignBasisLocalPowerDistributionAdvancedNuclearFuelsXN-19x9Fuel..202125~~~~~~~~~~o~~~~~~~~~~~A9~~~~~~~~~~~A10stogram....B-4B-5~~~~~~~~~
0~~illANF-87-1PRevisionLISTOFFIGURES(Continued)~FiereParcaeB.3-4DesignBasisLocalPowerDistributionGeneralElectric(Central)8xSRFuel..............................B-7B.3-5DesignBasisLocalPowerDistributionGeneralElectric(Peripheral)SxSRFuel...........................B-8 ANF-87-125Revision1.
1.0INTRODUCTION
ThisreportpresentstheresultsofAdvancedNuclearFuelsCorporation's*evaluationofsystemtransienteventsforSusquehannaUnit2Cycle'operation.Theevaluationtogetherwithcoretransienteventsdeterminesthenecessarythermalmargin(HCPRlimits)toprotectagainsttheoccurrenceofboilingtransitionduringthemostlimitinganticipatedtransient.Thermalmarginsarecalculatedforoperationwithintheallowedregionsofthepower/flowoperatingmapuptothefullpower/fullflowoperatingcondition.Theevaluationalsodemonstratesthevesselintegrityforthemostlimitingpressurizationevent.ThebasesfortheseanalyseshavebeenprovidedinReferencel.'FormerlyExxonNuclearCompany(ENC).
f~'
ANF-87-125Revision12.0SUMMARYTheSusquehannaUnit2Cycle3corecanbe.describedasfollows:~FeelTeANFXN-2XN-2ANFXN-1GE8x8R8x8RNo.of~Ass140963241968BundleAverageEnrichment3.33/9Gd4*3.33/10Gd53.312.19e~UsingANF'smethodologyandconsideringtheCycle3core,themostlimitinganticipatedplantsystemtransientwithregardtothermalmarginatratedpowerandflowconditionswasconfirmedtobethegeneratorloadrejectionwithoutbypass(LRWB)transientwithrecirculationpumptrip(RPT)operable.TheMinimumCriticalPowerRatio(MCPR)limitsforpotentiallylimitinganticipatedplantsystemtransienteventsareshowninTable2.1forcomparison.ThevaluesinTable2.1weredeterminedassumingboundingconditionsintheanalyses.ResultswithRPToutofservicearereportedinSection3.2.1.Thesetransientswereevaluatedwithallco-residentfueltypesmodeledandthemostlimitingconditionwasusedtodeterminethereportedMCPRs.TheControlRodWithdrawalError(CRWE)analysisandresultingdeltaCPRresultsarereportedinReference2.Maximumsystempressurehasbeencalculatedforthecontainmentisolationevent,whichisarapidclosureofallmainsteamisolationvalves,usingthescenarioasspecifiedbytheASMEPressureVesselCode.ThisanalysisshowsthatthesafetyvalvesofSusquehannaUnit2havesufficientcapacityand*Thefirstnumberstatesthenumberofgadoliniarodsperbundleandthesecondnumberstatestheweightpercentgadoliniaperrod.Thegadoliniaconcentrationsandnumberofrodsperbundlearestatedforfreshfuelonly.Theothersarenotsignificant.
ANF-87-1Revisionperformancetopreventthepressurefromreachingtheestablishedtransientpressuresafetylimitof110%ofdesignpressure(l.1x1250=1375psig).Theanalysisalsoassumedsixsafetyreliefvalvesoutofservice.ThemaximumsystempressurespredictedduringtheeventareshowninTable2.1.Resultsofthesingleloopoperation(SLO)analysisareshowninAppendixA.ThesafetylimitanalysisforsingleloopoperationsupportsanincreaseintheHCPRSafetyLimitof.01.
ANF-87-125Revision1TABLE2.1TRANSIENTANALYSISRESULTSATDESIGNBASISCONDITIONS*hCPRMCPR**TransientLoadRejectionWithoutBypasswithRecirculationPumpTripFeedwaterControllerFailurewithBypassLossofFeedwaterHeatingANF9x90.24/1.300.23/1.290.16/1.22GE8x8R0.21/1.270.20/1.260.15/1.21MaximumPressuresiTransientMSIVClosureVesselDome1281VesselLowerPlenum1297SteamLine1284*104%power/100%flow.*BasedontheMCPRSafetyLimitof1.06confirmedherein.
pW ANF-87-125RevisionI3.0TRANSIENTANALYSISFORTHERMALMARGIN3.1DesinBasisConsistentwiththeFSARplanttransientanalysis,thermalmarginoperatingMCPRlimitsaredeterminedbasedonthe104%power/100%flowoperatingpoint.ThisthermalmarginoperatingMCPRlimitisthenmodifiedasafunctionofpowerandflowasrequiredtoprotectagainstboilingtransitionresultingfromanticipatedtransientsoccurringfromallowedconditionsonthepower/flowoperatingmap.Theplantconditionsforthe.104%power/100%flowpointareasshowninTable3.1.ThemostlimitingpointinCycle3hasbeendeterminedtobeatthemaximumCycle3licensingexposurelimitwhencontrolrodsarefullywithdrawnfromthecore.Thethermalmarginlimitestablishedforthisexposureconditionisconservativeforcaseswherecontrolrodsare~~artiallyinserted.FollowingrequirementsestablishedinthePlantOperatingLicenseandassociated,Techn'icalSpecifications,observanceofaMCPRoperatinglimitof1.30forANF9x9fueland1.27forGE8x8RfuelorgreaterconservativelyprotectsagainstboilingtransitionduringanticipatedplantsystemstransientsfromdesignbasisconditionsforSusquehannaUnit2Cycle3.ThecalculationalmodelsusedtodeterminethermalmarginincludeANF'splanttransientandcorethermal-hydrauliccodesasdescribedinpreviousdocumentation(I~).Fuelpellet-to-cladgapconductancesusedintheanalyseswerebasedoncalculationswithRODEX2().Table3.2summarizesthevaluesusedforimportantparametersthatprovidedaboundinganalysis.RecirculationPumpTrip(RPT)flowcoastdownwasinputbasedonmeasuredSusquehannaUnit2startuptestdata.ToconfirmtheneutronicsasrequestedbytheSERissuedforthesupplementsofReferencel(8),theSusquehannasystemtransientmodelwasbenchmarkedtoappropriateSusquehannaUnit2startuptestdata.XCOBRA-T()wasusedtocalculatethechangeincriticalowerratio(deltaCPR)forpressurization'ventanalyses.i ANF-87-1Revision3.2AnticiatedTransientsANFconsiderseightcategoriesofpotentialsystemtransientoccurrencesforJetPumpBWRsinXN-NF-79-71(1~).ThethreemostlimitingtransientsaredescribedhereindetailtoshowthethermalmarginforCycle3ofSusquehannaUnit2.Thesetransientsare:LoadRejectionWithoutBypass(LRWB)FeedwaterControllerFailure(FWCF)LossofFeedwater-Heating(LFWH)AsummaryofthetransientanalysesisshowninTable3.3.Otherplanttransienteventsareinherentlynonlimitingorclearlyboundedbyoneoftheaboveevents.3.2.1LoadRe'ectionWithoutBassThiseventisthemostlimitingoftheclassoftransientscharacterizedbyrapidvesselpressurization.Thegeneratorloadrejectioncausesaturbinecontrolvalvetrip,whichinitiatesareactorscramandarecirculatingpump~trip(RPT).Thecompressionwaveproducedbythefastcontrolvalveclosuretravelsthroughthesteamlinesintothevesselandcreatesthevesselpressurization.Turbinebypassflow,whichcouldmitigatethepressurizationeffect,isnotallowed.TheexcursionofcorepowerduetovoidcollapseisprimarilyterminatedbyreactorscramandvoidgrowthduetoRPT.Figures3.1and3.2depictthetimevarianceofcriticalreactorandplantparametersduringtheloadrejectiontransientcalculationwithboundingassumptions.TheboundingassumptionsareconsistentwithANF'sCOTRANSAcodeuncertaintiesanalysismethodologyasreportedinReference8andapprovedbytheNRC.Theboundingassumptionsinclude:
ANF-87-125Revision1TechnicalSpecificationminimumcontrolrodspeedTechnicalSpecificationmaximumscramdelaytimeIntegralpowerincreasedby10%Atdesignbasisconditions(104%power/100%flow)thisresultsinadeltaCPRof0.24fortheloadrejectionwithoutbypasswhenRPTis'perableforANF9x9fuels.ThecorrespondingdeltaCPRforGE8xSRfuelis0.21.TheloadrejectionwithoutbypasseventwasalsoanalyzedatthedesignbasisconditionswhenRPTisnotoperable.TheresultingdeltaCPR'sare0.37and0.32forANF9x9andGESxSRfuels,respectively.3.2.2FeedwaterControllerFailure~~Failureofthefeedwatercontrolsystemispostulatedtoleadtoamaximumincreaseinfeedwaterflowintothevessel.Astheexcessivefeedwaterflowsubcoolstherecirculatingwaterreturningtothereactorcore,thecorepowerwillriseandattainanewequilibriumifnootheractionistaken.Eventually,theinventoryofwaterinthedowncomerwillriseuntilthehighlevelvesseltripsettingisexceeded,Toprotectagainstspilloverofsubcooledwatertotheturbine,theturbinetrips,closingtheturbinestopvalvesandinitiatingareactorscram.Thecompressionwavethatiscreated,thoughmitigatedbybypassflow,pressurizesthecoreandcausesapowerexcursion.Thepowerincreaseisterminatedbyreactorscram,RPT,andpressurerelieffromthebypassvalvesopening.TheevaluationofthefloweventatdesignbasisconditionswasperformedwithboundingvaluesandresultedinadeltaCPRof0.23forANF9x9fuelsand0.20forGESx8Rfuel.Figures3.3and3.4presentkeyvariablesforthisfeedwatercontrollerfailureevent.Thiseventwasalsoexaminedforreducedpowerconditionsatfullflow.TheresultsfortheFWCFtransientsfromreducedpowerconditionsareshowninTable3.4forall9x9andSx8fuels.hecalculatedresultsshowthatFWCFdeltaCPR'svarywithdecreasingpower 8ANF-87-1Revisionatfullflowconditions.ThehighestdeltaCPRwascalculatedat40%powerand100%flowconditions.Thistransienteventatfullpowerandfullflowconditionswasalsoanalyzedassumingboundingconditionsandfailureofthebypassvalvestoopen.ThisresultsinadeltaCPRof0.28forANF9x9fuelsand0.25forGE8x8Rfuel.3.2.3ossOfFeedwaterHeatinThelossoffeedwaterheatingleadstoagradual.increaseinthesubcoolingofthewaterinthereactorlowerplenum.Reactorpowerslowlyrisestothethermalpowermonitorsystemtripsetpoint.Thegradualpowerchangeallowsfuelthermalresponsetomaintainpacewiththeincreaseinneutronflux.UsingthemethodologyofReference1thedeltaCPRfortheeventinCycle3is0.16forANF9x9fueland0.15forGE8x8Rfuel.Figures3.5and3.6depikeyvariablesforthelossoffeedwaterheatingevent.Thebypassvalvesdonotsignificantlyaffectthelossoffeedwaterheatingresults.Thus,thedeltaCPRlimitisapplicablewhetherthebypassvalvesareoperableornot.3.3CalculationalModelTheplanttransientcodeusedtoevaluatethegeneratorloadrejectionandfeedwaterflowincreasewasANF'scodeCOTRANSA().Theaxialone-dimensionalneutronicsmodelpredictedreactorpowershiftstowardthecoremiddleandtopaspressurizationoccurred.Thiswasaccountedforexplicitlyindeterminingthermalmarginchangesinthetransient.ThelossoffeedwaterheatingeventwasevaluatedwithPTSBWR3andXCOBRA(Reference1).AppendixA(1)oftheSusquehannaUnit1Cycle2analysi'sdelineatesthechangesmadetoCOTRANSA(1)tomergethePTSBWR3codewiththeCOTRANSAcode,torefinenumericaltechniquesandtoimproveinput.Reference9describestheXCOBRA-TcodeustocalculatethedeltaCPR'sforthepressurizationtransients.AppendixBo 9ANF-87-125Revision1Reference10delineatestheplantrelatedchangesmadetothesecodesfortheSusquehannaUnits1and2analyses.3.4SafetLimitThesafetylimitistheminimumvalueofthecriticalpowerratio(CPR)atwhichthefuelcouldbeoperatedwheretheexpectednumberofrodsinboilingtransitionwouldnotexceed0.1%ofthefuelrodsinthecore.ThesafetylimitistheHCPRwhichwouldbepermittedtooccurduringthelimitinganticipatedoperationaloccurrence.AHCPRsafetylimitof1.06forallfueltypesinSusquehannaUnit2Cycle3wassupportedbythemethodologypresentedinReference3.TheinputparametersanduncertaintiesusedtosupportthesafetylimitarepresentedinAppendixBofthisreport.
10ANF-87-l~Revision~TABLE3.1REACTORDESIGNANDPLANTCONDITIONSSUSQUEHANNAUNIT2ReactorThermalPower(104%)'otalCoreFlow(1005)CoreIn-ChannelFlowCoreBypassFlowCoreInletEnthalpyVesselPressuresSteamDomeUpperPlenumCoreLowerPlenumTurbinePressureFeedwater/SteamFlowFeedwaterEnthalpyRecirculationPumpFlow(perpump)3439Mwt100.0Mlb/hr89.9Mlb/hr10.1Mlb/hr518.0Btu/ibm1035psia1045psia1052psia1066psia975psia14.15Mlb/hr360.8Btu/ibm15.75Mlb/hr ANF-87-125Revision1TABLE3.2SIGNIFICANTPARAMETERVALUESUSEDINTHEANALYSISFORSUSQUEHANNAUNIT2HighNeutronFluxTripControlRodInserti'onTimeControlRodWorthVoidReactivityFeedbackTimetoDeenergizedPilotScramSolenoidValvesTimetoSenseFastTurbineControlValveClosureTimefromHighNeutronFluxTimetoControlRodMotionTurbineStopValveStrokeTimeTurbineStopValvePositionTripTurbineControlValveStrokeTime(Total)Fuel/CladdingGapConductanceCoreAverage(Constant)Safety/ReliefValvePerformanceSettingsReliefValveCapacityPilotOperatedValveDelay/Stroke125.3%3.49sec/90%insertednominalnominal200msec(maximum)30msec290msec100msec90%open70msec758.0Btu/hr-ft2-FTechnicalSpecifications225.4ibm/sec(1110psig)400/150msec 12ANF-87-1RevisionTABLE3.2SIGNIFICANTPARAMETERVALUESUSEDINTHEANALYSISFORSUSQUEHANNAUNIT2(CONTINUED)MSIVStrokeTimeMSIVPositionTripSetpointTurbineBypassValvePerformanceTotalCapacityDelaytoOpening(80%open)FractionofEnergyGeneratedinFuelVesselMaterLevel(abovei'nstrumentzero)HighLevelTripNormalLowLevelTripMaximumFeedwaterRunoutFlowThreePumpsRecirculationPumpTripSetpointVesselPressure3.0sec90%open936.11ibm/sec300msec0.96558.7in35in*8in5049ibm/sec1170psig*COTRANSAplotsaregivingwaterlevelaboveseparatorskirtandthevalhereisaboveinstrumentzero.
13ANF-87-125.Revision1TABLE3.2SIGNIFICANTPARAMETERVALUESUSEDINTHEANALYSISFORSUSQUEHANNAUNIT2(CONTINUED)ControlCharacteristicsSensorTimeConstantsPressureOthersFeedwaterControlModeFeedwaterMasterControllerProportionalGainResetRateFeedwater100%MismatchWaterLevelErrorSteamFlowEquivalentFlowControlModePressureRegulatorSettingsLeadLagGain500msec250msecThree-Element500(%/%)(%/ft)1.70(%/sec/ft)4.0ft4034ft/100%Manual3.0sec7.0sec3'3%/psid ANF-87-1RevisionTABLE3.3RESULTSOFSYSTEMPLANTTRANSIENTANALYSESEventLoadRejectionWithoutBypassFeedwaterControllerFailureLossofFeedwaterHeatingMSIVClosurewithFluxScramMaximumNeutronFlux%Rated267233123342Maximum-CoreAverageHeatFlux%Rated116.2116.8121.3133.2MaximumSystemPressure~sia1194117910781312hCPRFor9x9Fuels0.240.230.16NOTE:Alleventsareboundingcaseat104%power/100%flow.
15ANF-87-125Revision1'ABLE3.4FEEDWATERCONTROLLERFAILUREANALYSISRESULTSAT100%FLOW-%PowerDelt'aCPR1048040ANF9x90.230.250.280.31GE8x8R0.200.230.260.28 l,.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATIONFLOW4.VESSELSTEAMFLOW5.FEEDWATERFLOWCIOlCl~O~nLKOZOLUCJCELUCL45123233425CIICI000.20.50.71.01.2TIME,SEC1.51.72.02.22.5Figure3.1LoadionWithoutBypass i.VESSELPRESSURECHANGE(PSI)2.VESSELHATERLEVEL(IN)ClLO1W.O0.20.50.71.01.2TIME,SEC1.51.72.02.22.5Figure3.2LoadRejectionWithout8ypass i.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATIONFLOW4.VESSELSTEAMFLOW5.FEEDWATERFLOW55<24i412i620TIME,SEC2428323640figure3.3Feedw.ontrol1erFailure IDCtCUi.VESSELPRESSURECHANGE(PSI)2.VESSELWATERLEVEL(IN)CIOJoP121620TIME,SEC2428323640Figure3.4FeedwaterControllerFailure ClCl124545123i.NEUTRON.FLUXLEVEL2.HEATFLUX3.RECIRCULATIONFLOW4.VESSELSTEAMFLOW5.FEEDWATERFLOW33I-OtOC3IXUJ0C)I21t010-2030405060TIME,SEC708090100Figure3.5LosseedwaterHeating i.VESSELPRESSURECHANGE(PSI)2.VESSELWATERLEVEL(IN)22LAILLICILAILAIIQ1020304050TIME,SEC60708090100Figure3.6LossOfFeedwaterHeating lKV!-.$1r~0 22ANF-87-125"Revision14.0MAXIMUMOVERPRESSURIZATIONMaximumsystempressurehasbeencalculatedforthecontainmentisolationevent(rapidclosureofallmainsteamisolationvalves)withanadversescenarioasspecifiedbytheASHEPressureVesselCode.ThisanalysisshowedthatthesafetyvalvesofSusquehannaUnit2havesufficientcapacityandperformancetopreventpressurefromreachingtheestablishedtransientpressuresafetylimitof110%ofthedesignpressure(1375psig).ThemaximumsystempressurespredictedduringtheeventareshowninTable2.1.Thisanalysisassumedsixsafetyreliefvalvesoutofservice.4.1Desin'asisThereactorconditionsusedintheevaluationofthemaximumpressurizationventarethoseshowninTable3.1.Themostcriticalactivecomponent(scramonHSIVclosure)wasassumedtofailduringthetransient.ThecalculationwasperformedwithANF'sadvancedplantsimulatorcodeCOTRANSA(),whichincludesanaxialone-dimensionalneutronicsmodel.4.2PressurizationTransientsANFhasevaluatedseveralpressurizationeventsandhasdeterminedthatclosureofallHainSteamIsolationValves(HSIVs)withoutdirectscramisthemostlimiting.-AlthoughtheclosurerateoftheHSIVsissubstantiallyslowerthantheturbinestopvalvesorturbinecontrolvalves,thecompressibilityoftheadditionalfluidinthesteamlinesresultsinalessseveretransientforthefasterturbinestop/controlvalveclosuretransients.Essentially,therateofsteamvelocityreductionisconcentratedtowardtheendofthevalvestroke,generatingasubstantialcompressionwave.Oncethecontainmentisisolatedthesubsequentcorepowerproductionmustbeabsorbedinasmallervolumethanifaturbinetriphadoccurred.Calculationshavedeterminedthattheoverallresultistocauseisolation(HSIV)closurestobemorelimitingorsystempressurethanturbinetrips.
23ANF-87-1PRevision'.2.1ClosureOfAllHainSteamIsolationValvesThiscalculationassumedthatsixreliefvalveswereoutofserviceandthatallfoursteamisolationvalveswereisolatedatthecontainmentboundarywithin3seconds.Atabout3.0seconds,thereactorscramisinitiatedbyreachingthehighfluxtripsetpoints.SincescramperformancewasdegradedtoitsTechnicalSpecificationlimit,effectivepowershutdownisdelayeduntilafter4.4seconds.Substantialthermalpowerproductionenhancespressurization.Pressuresreachtherecirculationpumptripsetpoint(1170psig)beforethepressurizationisreversed.Lossofcoolantflowleadstoenhancedsteamproductionaslesssubcooledwaterisavailabletoabsorbcorethermalpower.Themaximumpressurecalculatedinthesteamlineswas1284psigoccurringnearthevesselatabout6.5seconds.Themaximumvesselpressurewas1297psigoccurringinthelowerplenumatabout6.3seconds.
ANF-87-125Revision15.0RECIRCULATIONPUMPRUN-UPAnalysisofpumprun-upeventsforoperationatlessthanratedrecirculationpumpcapacitydemonstratestheneedforanaugmentationofthefullflowMCPRoperatinglimitforlowerflowconditions.Thisisduetothepotentialforlargereactorpowerincreasesshouldanuncontrolledpumpflowincreaseoccur.Thissectiondiscussespumpexcursionswhentheplantisinmanualflowcontroloperationmode.Resultsobtainedfrompreviousanalysesshowedthetwopumprun-upboundsthesinglepumprun-up.Onlythetwopumprun-upisevaluatedforSusquehannaUnit2Cycle3.TheseresultsindicatethatHCPRwoulddecreasebelowthesafetylimitifthefullflowreferenceMCPRisobservedatinitialconditions.Thus,anaugmentedHCPRisneededforpartialflowoperationtopreventviolationoftheHCPRSafetyLimitforthetwopump~~xcursionevent.Theanalysisofthetwopumpflowexcursionindicatesthatthelimitingeventisagradualpowerincreaseinwhichtheheatfluxtrackspower.TheSusquehannaUnit2Cycle3analysisconservativelyassumedtherun-upeventinitiatedat57%power/40%flowandreached111%ratedpowerat100%ratedflow.Theeventterminatedat105%ofratedflowwithaminimumCPRof1.06.Theresultsofthetwopumprun-upanalysesformanualflowcontrolarepresentedinFigure5.1.ThecyclespecificHCPRlimitforSusquehannaUnit2Cycle3shallbethemaximumofthereducedflowMCPRoperatinglimit,thefullflowHCPRoperatinglimit,orthepowerdependentHCPRoperatinglimit.
1.50ANF9X9FUELS----GE8X8RFUEL1.401.301.20~Luo4060BO708090100TOTALCORERECIRCULATIONFLOW(%RATED)Figure5.1SusquehannaUnit2.3ReducedFlowHCPROperatingLimit 26ANF-87-125Revision16.0.REFERENCESR.H.Kelley,"ExxonNuclearPlantTransientMethodologyforBoilingIltRt,"~XN-NF.I-II,RII2,AddNIICorporation*,Richland,WA99352,November1981.2.3.4.5.,7.8.9.10.J.A.White,"SusquehannaUnit2Cycle3ReloadAnalysis,DesignandSafetyAnalyses,"8NF-87-126,.AdvancedNuclearFuelsCorporation,Richland,WA99352,October1987.J.A.White,"ExxonNuclearMethodologyforBoilingWaterReactors,THERMEX:ThermalLimitsHethodology,SummaryDescription,"XN-NF-80-~19PA,Volume,3,Revision2,AdvancedNuclearFuelsCorporation,Richland,WA99352,January1987.T.W.Patten,"ExxonNuclearCriticalPowerMethodologyforBoilingWaterR,"~52-525A,R11I,AddNIFIC5Richland,WA99352,November1983.T.H.Keheley,"SusquehannaUnit2Cycle2PlantTransientAnalysis,"XN-NF-86-55,Revision1,AdvancedNuclearFuels'orporation,Richland,WA99352,Hay1986.T.H.Keheley,"SusquehannaUnit1Cycle4PlantTransientAnalysis,"XN-NF-87-22,AdvancedNuclearFuelsCorporation,Richland,WA99352,April1987.K.R.Herckx,"RODEX2FuelRodMechanicalResponseEvaluationModel,"XN-~NF.I-N,RII2,AddllIFIC5tl,Illlld.,IIA99352,April1984.S.E.Jensen,"ExxonNuclearPlantTransientMethodologyforBoilingIltRt,"~XN-II-I-I,RIII,AdvancedNuclearFuelsCorporation,Richland,WA99352,March1986.H.J.Ades,"XCOBRA-T:AComputerCodeforBWRTransientThermal-HydraulicCoreAnalysis,"XN-NF-84-105PA,Volume1&Volume1,Supp.18Supp.2,AdvancedNuclearFuelsCorporation,Richland,WA99352,February1987.T.H.Keheley,"SusquehannaUnit1Cycle2PlantTransientAnalyses,"XN-NF-84-118,includingSupplement1,AdvancedNuclearFuelsCorporation,Richland,WA99352,December1984.FormerlyExxonNuclearCompany(ENC).
I A-IANF-87-125RevisionIAPPENDIXASINGLELOOPOPERATIONTheNSSSsupplierhasprovidedanalyseswhichdemonstratethesafetyofplantoperationwithasinglerecirculationloopoutofserviceforanextendedperiodoftime.Theseanalysesrestricttheoveralloperationoftheplanttolowerbundlepowerlevelsandlowernodalpowerlevelsthanareallowedwhenbothrecirculationsystemsareinoperation.=Thephysicalinterdependencebetweencorepowerandrecirculationflowrateinherentlylimitsthecoretolessthanratedpower.ANFfuelwasdesignedtobecompatiblewiththeco-residentfuelinthermalhydraulic,nuclear,andmechanicaldesignperformance.TheANFmethodologyhasgivenresultswhichareconsistentwithhoseofthepreviousanalysesfornormaltwo-loopoperation.ManyanalysesperformedbytheNSSSsupplierforsingleloopoperationarealsoapplicabletosingleloopoperationwithfuelandanalysesprovidedbyANF.Adiscussionoftherelevanteventsandlimits.forsingleloopoperationfollow.AlsoincludedareresultsofANFanalyseswhichconfirmtheNSSSvendorconclusions.A.lABNORMALOPERATINGTRANSIENTSMCPRlimitsestablishedforfullflowtwoloopoperationareconservativeforsinglelooptransientsbecauseofthephysicalphenomenarelatedtopart-powerpart-flowoperation,notbecauseoffeaturesinreactoranalysismodelsorcompatiblefueldesigns.AreviewofthemostlimitingdeltaCPRtransientsforsingleloopoperationwasconducted.Undersingleloopconditions,steadystateoperationcannotexceedapproximately76%powerand61%coreflowbecauseofthecapabilityoftherecirculationlooppump.Thus,theMCPR~~limitatmaximumpowerishigherthanthetwopumpoperatingMCPRlimitduetotheflowdependentMCPRfunction.Thisflowdependenceisbasedonaflow A-2ANF-87-1~Revision~increasetransientfromrunup'ftwopumps.Flowrunupsfromasinglerecirculationpumpwouldbemuchlesssevere,thoughtheconservativetwopumplimitisretained.LoadRe'ectionWithoutBassThelimitinganticipatedsystemtransientfortheSusquehannaUnitsistheLoadRejectionWithoutBypass(LRWB)pressurizationtransient.Inthistransient,theprimaryphenomenaisthepressurizationcausedbyabruptly'toppingthesteamflowthroughrapidclosureqftheturbinecontrolvalve.Whentherapidpressurizationreachesthecoreitcausesapowerexcursionduetovoidcollapse.ThereducedpowerandflowanalysesfortheSusquehannaUnitsdescribedinReferenceA-1undertwoloopoperationshowsthattheresultingpowexcursionandassociateddeltaCPRarereducedbelowthoseofthefulpower/fullflowcase.ThusfortheSusquehannaUnitstheHCPRlimitsbasedonLRWBanalysesatfullpowerareconservativelyapplicabletothelowerpowers/flowsassociatedwithsingleloopconditions.Furthermore,LRWBanalysesbyANFatreducedpowerandflowconditionsinotherBWR'swithsingleloopoperationconfirmthistrend.A.1.2FeedwaterControllerFailureThesecondworstlimitingtransientatfullpowerandflowistheFeedwater'ontrollerFailure(FWCF)tomaximumdemand.Thistransientisalsolesssevereatthepowerandflowconditionsassociatedwithsingleloopoperation.Thistransientassumesthefeedwatercontrollerfailstomaximumdemandandresultsinthemaximumamountofsubcooledfeedwaterinthedowncomer.Whenthiscoolerwaterreachesthecorethepowerrises.Thecorepowerriseisterminatedbyareactorscraminitiatedbyaturbinetrip.Theturbinetr' A-3ANF-87-125RevisionIistheresultofthehighwaterleveltripcausedbytheadditionalamountoffeedwaterbeinginjected.IAtthereducedrecirculationflows,thesubcoolinginthedowncomerduetothehighfeedwaterflowtakeslongertotransversethecoresothatahighwaterleveltripoccurs.beforecorepowercanriseashighasitdoesinthefullflowcase.AswiththeLRWB,thepressurizationeventresultingfromtheturbinetripislesssevereforthereducedpowerinSLO..Thus,becauseoftheslowerenthalpytransportphenomenacausedbythelowerrecirculationflowandbecauseofthelowersteamlineflowinthepressurizationportionofthetransient,theFWCFhaslargermargintotheoperatinglimitinsingleloopoperationthanintwoloopoperation..1.3PumSeizureAccidentPumpseizureisapostulatedaccidentwheretheoperatingrecirculationpumpsuddenlystopsrotating.Thiscausesarapiddecreaseincoreflow,adecreaseintherateatwhichheatcanbetransferredfromthefuelrodsandadecreaseinthecriticalpowerratio.AnalyseswithCOTRANSAandXCOBRA-TshowthatforCycle3theCPRforANFfuelwould.decreaseby0.30duringapumpseizureforsingleloopoperation.TheCOTRANSAcodewasusedtosimulatesystemresponsetoapumpseizureinsingleloopoperationfromtheconditionsspecifiedinTableA.l.Theoperatingrecirculationpumprotorwasstoppedin0.1secondscausingasuddendecreaseinactivejetpumpdriveflow.Atabout6.7secondstheinactivejetpumpdiffuserflowwentfromnegativeflowtopositiveflow.In7.3secondsthedomepressuredecreasedtoaminimumvalueof970.3psiaandthenstartedtoincreaseagain.FiguresA.IandA.2presentagraphicalrepresentationofimportantsystemparametersduringthetransient.
A-4ANF-87-1,RevisionThedeltaCPRforthiseventwascalculatedusingXCOBRA-T.TheANF9x9fuelreachedamaximumdeltaCPRof0.30at2.2secondsintothetransient.TheGE8x8RfuelreachedamaximumdeltaCPRof0.29at2.15secondsintothetransient.A.1.4MCPRSafetLimitForsingleloopoperation,theNSSSvendorfoundthatanincreaseof0.01intheMCPRsafetylimitwasneededtoaccountfortheincreasedflowmeasurementuncertaintiesandincreasedTIPuncertaintiesassociatedwithsinglepumpoperation.ANFhasevaluatedtheeffectsoftheincreasedflowmeasurementuncertaintiesonthesafetylimitMCPRandfoundthattheNSSSvendordeterminedincreaseintheallowedsafetylimitMCPRisalsoapplicabletoANFfuelduringsingleloopoperation.Thus,increasingthesafetylimitMCPRby0.01forsingleloopoperation(1.07)withANFfuelissufficientconservativetoalsoboundtheincreasedflowmeasurementuncertaintiesfosingleloopoperation.A.1.5~SummarThelimitingMCPRoperatinglimitForsingleloopoperationisconservativelysetusingthelimitingpumpseizureaccidentdeltaCPRplusthesingleloopoperationMCPRsafetylimit.ThislimittogetherwiththeMCPRfcurvefortwoloopoperationplus.01andtheMCPRpcurvefortwoloopoperationplus.Olconservativelyboundalltransients.
A-5ANF-87-125Revision1A.2MAPLHGRLIMITSANFperformedLOCAanalysesforsingleloopconditionsanddeterminedthattheMAPLHGRlimitcurvefortwo-loopoperationisalsoapplicabletosingleloopoperationforANFfuels(ReferenceA-2).
A-6ANF-87-1RevisionA.3STABILITY.TheTechnicalSpecificationsrequireAPRM/LPRMsurveillancetotheleftofthe45%ConstantFlowlineandabovethe80%RodBlockline.BasedoncorehydrodynamicstabilityanalysesforCycle3,operationatpower/flowcombinationsaboveandtotheleftofthelineconnectingthe66%Power/45%Flowand69%Power/47%FlowpointsneedstobeaddedtotheAPRM/LPRMsurveillancerequirements.FigureA.3showsthecorepowerversuscoreflowestablishedforCycle3.
A-7ANF-87-]25Revision1TABLEA.1SLOREACTORANDPLANTCONDITIONSReactorThermalPowerTotalRecirculationFlowCoreBypassFlowCoreInletEnthalpyVesselPressuresSteamDomeLowerPlenumTurbinePressureSteamFlowFeedwaterEnthalpy2489HMt60.35Hlb/hr5.70Hlb/hr507.3Btu/lb994.5psia,1011.3psia965.4psia9.8Hlb/hr330.7Btu/lb l.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATIONFLOW4.VESSELSTEAMFLOW5.FEEOWATERFLOW5iioTIME,SEC14i6iB20FigureA.lSingleLeration-PumpSeizure CI1.VESSELPRESSURECHANGE(PSI)2.VESSELHATERLEVEL(IN)CUOIQ10TIME,SEC1214161820FigureA.2SingleLoopOperation-PumpSeizure A-10ANF-87-125Revisi'o'n0~~~~~0~~~~~'g~~~~~~~0~~~0~~~~Q~~~~'~80NCC70OBO403020rrCAPRM:SCRAMLINE;~)0~~~~0~~~0'0~~>0~~r~rv~r~~~0~~~~~~~ega~0~~~<~~~~~~q~~~~~0~~10)XXer..RODLINE\~00~0A$100~~~~~~~~~0)0~~~~~~~$~0~~~~~~ROD;BLOCK;MOMTOR~)10~~~~0~f~~~~~~~~~~~~~~~)~~~~~~~(66/45)0g~~~~~~0000~~ep~~~~~~~~)~APRM.'ODBLOCKIe~op000~~~~~0~00~00~~~$~~J'0~/0JI01~~1)0~~00~$~00~~001~~~8II.~~10~04~~~1~~~~45Kt,'OREFLdWRODLINE~i'.~~4~~000~0~~~~~~00~~~80K~~~~J~~00~00el'L~~~~000~~00el~.~~~~~J~JJ0~~~Joe0~~~~~~~~~~>~~~~~~~~~~~~LN$TCERC2'-,PUMPMINFLOW:10~~~~~~~~~~~00102030406060708090COREFLOW,%RATED100FigureA03CorePowerVersusCoreFlow A-llANF-87-125Revision1A.4A-1.REFERENCESJ.C.Chandler,"SusquehannaUnit1Cycle3ReloadAnalysis,"XN-NF-85-132,Revision1,AdvancedNuclearFuelsCorporation,Richland,WA99352,December1985.A-2.':R.Swope,"SusquehannaLOCAAnalysisforSingleLoopOperation,"XN-NF-86-125,AdvancedNuclearFuelsCorporation,Richland,WA99352,November1986.A-3.K.D.Hartley,etal.,"SusquehannaUnit2Cycle2StabilityTestResults,"XN-NF-86-90,Supplement1,AdvancedNuclearFuelsCorporation,Richland,WA99352,January1987.
0 B-1ANF-87-125Revision1APPENDIXBMCPRSAFETYLIMITB.1INTRODUCTIONTheHCPRfuelcladdingintegritysafetylimitwascalculatedusingthemethodologyanduncertaintiesdescribedinReferenceB.1.Inthismethodology,aHonteCarloprocedureisusedtoevaluateplantmeasurementandpowerpredictionsuncertaintiessuchthatduringsustainedoperationattheHCPRCladdingIntegritySafetyLimit,atleast99.9%ofthefuelrodsinthecorewouldbeexpectedtoavoidboilingtransition.Thisappendixdescribesthecalculationandpresentstheanalyticalresults.i B-2ANF-87-1RevisionB.2CONCLUSIONSDuringsustainedoperationataHCPRof1.06withthedesignbasispowerdistributiondescribedbelow,atleast99.9%ofthefuelrodsinthecoreareexpectedtoavoidboilingtransitionataconfidencelevelof95%.
B-3ANF-87-125Revision1B.3DESIGNBASISPOWERDISTRIBUTIONPredictedpowerdistributionswereextractedfromthefuelmanagementanalysisforSusquehannaUnit2Cycle3.Theradialpowerdistributionswereevaluatedforperformanceasthedesignbasisradialpowermap,andthedistributionat8,000HWd/HTUexposurewasselectedasthemostsevereexpecteddistributionforthecycle.ThedistributionwasskewedtowardhigherpowerfactorsbytheadditionofbundleswitharadialpeakingfactorapproximatinganoperatingHCPRlevelof1.32atfullpower.TheresultingdesignbasisradialpowerdistributionisshowninFigureB.3-1.ThefuelmanagementanalysisindicatedthatthemaximumpowerANFbundle(XN-2)inthecoreattheend-of-cycleexposure(10;829.6HWd/HTU)waspredictedtobeoperatingatanexposurelevelof14,319HWd/HTU,soalocalpower~~istributiontypicalofanodalexposureof15,000HWd/HTUwasselectedasthedesignbasislocalpowerdistribution.UncontrolledlocalpowerpeakingdistributionsforbothANF9x9XN-29GD4%fueland10GD5%fuelwere'eviewed.Thelimitinglocalswerefoundtooccurat15,000HWd/HTUfor9GD4%fuel.ThisdistributionisshowninFigureB.3-2.Localpowerdistributionsfortwo'ndthreecycleirradiatedfuelwerechosenconservativelyforANFXN-1andGE8x8Rfuel,andareshowninFiguresB.3-3andB.3-4.AboundinglyflatlocalpowerdistributionwasselectedfortheGeneralElectricfuelintheperipherallowpowerregion.ThisdistributionisshowninFigureB-3.5.Thelimitingaxialpowerprofileselectedforthe8,000HWd/HTUstatepointofCycle3wasconservativelyselectedbasedonestablishedcriteria.
8070605000C)So2010000.20.00.60.8'RRDIRLPOWERPERKING1.2SCAotAQ)O6FigureB.3-1SusquehannaUnitle3DesignBasisRadialPowerHistogram 8-5ANF-87-125Revision1*~0:0.88:0.91:0.96:1.04:1.02:1.04:0.96:1.00:0.96**~*:0.91:0.93:0.98:1.07:0.911.07:0.97:1.04:1.01*~:0.96:0.98:0.90:1.04:1.03:1.04:1.04:0.99:0.96:**~*:1.04:1.07:1.04:1.00:0.99:1.001.05:0.94:1.04:*~*:1.02:0.91~~*~*1.03:0.99:0.00:0.98:1.05:1.07:1.04*~**~*.04:1.071.04:1.00;0.98:0.00:1.03:0.94:1.05:0.96:0.97:1.04:1.05:1.05:1.03:1.06:1.00:0.971.00:1.04:0.99:0.94:1.07:0.94-:1.00:0.94:1.010.96:1.010.96:1.04:1.04:1.05:0'7:1.01:0.97FigureB.3-2DesignBasisLocalPowerDistributionAdvancedNuclearFuelsXN-29X9Fuel B-6ANF-87-17Revision'~*:0.91:0.92:0.95:1.01:1.01:1.01:0.96:0.98:0.95*~**~*".0.92:0.94:0.98:0.97:1.05:0.95:0.99:0.95:0.98*~**~*:0.95:0.98:0.93:1.06:1.05:1.06:1.05:0.97:0.96*~**~1.01:0.97:1.06:1.03:1.03:1.04:1.07:1.06:1.02**~*~*~*~*1.01:1.05:1.05:1.03:0.00:1.01:1.07:1.06:1.011.01:0.95:1.06:1,04:1.01:0.00:1.04:0.96:1.020:0.96:0.99:1.05:1.07:1.07:1.04:1.06:1.00:0.96~~:0.98:0.95:0.97.:1.06:1.06:0.96:1.00:0.95:0.980.95:0.98:0.96:1.02:1.01:1.02:0.96:0.98:0.96FigureB.3-3DesignBasisLocalPowerDistributionAdvancedNuclearFuelsXN-19X9Fuel B-'7ANF-87-125Revision1*~*~*~**~*~*~*1.03:1.00:1.00:1.00:1.00:1,00:1.01:1.031.00:0.98:1.00:1.02:1.02:1.03:1.00:1.01yH*~*~*~*1.00:1.001.01:;1.011.01:0.90:1.03:1.00:*~*~1.00:1.02:1.01:0.89:0.00:1.01:1.02:1.00:*~*~*1.00:1.021.01:0.000.89:1.01:0.99:1.00:*~*~*~1.00:1'3:0,90:1.011.01:0.981.00:1.00:1.01:1.001,03:1.020.99:1.00:0.98:1.00:1.03:1.011.00,:1.00:1.00:1.00:1.001.03FigureB.3-4DesignBasisLocalPow'erDistributionGeneralElectric(Central)8XSRFuel B-S,ANF.-'87-1Revision,*~*:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:*~**~*:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:*~*:1.00:.1.00**1.00:1.00:1.00:1.00:1.00:1.00*~*:1.00*~**~1.00:1.00:1.00:0.00:1.00:1.00*~1.00:1.00:1.00:0.00:1.00;1.00:1.00:1.00*~*~*1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.001.00:1.00FigureB.3-5DesignBasisLocalPowerDistributionGeneralElectric(Peripheral)SXSRFuel B-9ANF-87-125Revision1B.4CALCULATIONOFTHENUMBEROFRODSINBOILINGTRANSITIONThemethodologyofReference8-1wasusedtoanalyzethenumberoffuelrodsinboilingtransition.TheXN-3correlation(B)wasusedtopredictcriticalheatfluxphenomena.FivehundredMonteCarlotrialswereperformedtosupporttheMCPRsafetylimit.Non-parametrictolerancelimits(B)wereusedinlieuofPearsoncurvefitting.TheuncertaintiesusedintheanalysisfornormalconditionswerethoseidentifiedinReferenceB-l.Atleast99.9%ofthefuelrodsinthecoreareexpectedtoavoidboilingtransitionwithaconfidencelevelof95%.
B-10ANF-87-'1RevisionB.5B-1.B-2.B-3.REFERENCEST.W.Patten,"ExxonNuclearCriticalPowerMethodologyforBoilingIltRt,"~XN-NptA,R11I,AddNIFCorporation,Richland,WA99352,November1983.R.B.MacduffandT.W:Patten,"TheXN-3CriticalPower5Iti,"X~N-Np-51A,R11I,dRppltI,AddNuclearFuelsCorporation,Richland,WA99352,October1982.PaulN.Somerville,"TablesforObtainingNon-ParametricToleranceLimits,"AnnalsofMathematicalStatistics,Vol.29,No.2(June1958),pp.599-601.
ANF-87-125Revision1IssueDate:SUSQUEHANNAUNIT2CYCLE3PLANTTRANSIENTANALYSISDistribution:D.A.AdkissonD.J.BraunR.E.CollinghamL.J.FedericoS.F.GainesR.G.GrummerK.0.HartleyH.J.HibbardS.E.JensenT.H.KeheleyJ.N.MorganL.A.NielsenD.F.RicheyG.L.RitterC.J.VolmerJ.A.WhiteH.E.WilliamsonH.G.Shaw/PP8L(20)DocumentControl(5)