ML17146B094

Rev 1 to Susquehanna Unit 2 Cycle 3 Plant Transient Analysis.

ML17146B094
Person / Time
Site:	Susquehanna
Issue date:	11/30/1987
From:	WHITE J A ADVANCED MEDICAL SYSTEMS, INC.
To:
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ML17146B090	List: ML17146B089 ML17146B092 ML17146B094 ML17146B095 ML17146B096 ML17146B102 ML18040B194
References
ANF-87-125, ANF-87-125-R01, ANF-87-125-R1, NUDOCS 8712310156
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ANF-87-125 REVISIONAQMHCSDoHAIL,EARPUDDLESCORPORATION SUSQUEHANNA UNIT2CYCLE3PLANTTRANSIENT ANALYSISNOVEMBER198787i23iOi56 87i'223PDRADOCK05000389',

',P.....,,

...,,~f'PgANAFFILIATE OFKRAFTIVERKUPIIONQ~KMfU ADVANCEDNUCLEARFUELSCORPORATION ANF-87-125 Revision1IssueOate:11iIOi87SUSQUEHANNA UNIT2CYCLE3PLANTTRANSIENT ANALYSISPreparedBy:J.A.WhiteBWRSafetyAnalysisLicensing andSafetyEngineering FuelEngineering andTechnical ServicesAHAFFIUATEOFKRAFTWFRKVHIOHQxwu CUSTOMERDISCLAIMER IMPORTANT NOTICEREGARDING CONTENTSANDUSEOFTHISDOCUMENTPLEASEREADCAREFULLY AdvancedNuclearFuelsCorporation's warranties andrepresentations con-cemingthesubjectmatterofthisdocumentarethosesetforthIntheAgreement betweenAdvancedNuclearFuelsCorporation andtheCustomerpursuanttowhichthisdocumentlsIssued.Accordingly, exceptasotherwise expressly pro-videdInsuchAgreement, neitherAdvancedNuclearFuelsCorporation noranypersonactingonitsbehalfmakesanywarrantyorrepresentation, expressed orimplied.withrespecttotheaccuracy, completeness, orusefulness oftheinfor-mationcontained inthisdocument, orthattheuseofanyinformation, apparatus, methodorprocessdisclosed inthisdocumentwillnotinfringeprivately ownedrights;orassumesanyliabilities withrespecttotheuseofanyinformation, ap-paratus,methodorprocessdisclosed inthisdocument.

Theinformation contained hereinIsforthesoleuseofCustomer.

Inordertoavoidimpairment ofr/ghtsofAdvancedNuclearFuelsCorporation inpatentsorinventions whichmaybeincludedintheinformation contained Inthisdocument.

therecipient, byitsacceptance ofthisdocument, agreesnottopublishormakepublicusegnthepatentuseoftheterm)ofsuchinformation untilsoauthorized inwritingbyAdvancedNuclearFuelsCorporation oruntilaftersix(6)monthsfollowing termination orexpiration oftheaforesaid Agreement andanyextension thereof,unlessotherwise expressly providedintheAgreement.

Norightsorlicensesinortoanypatentsareimpliedbythefurnishing ofthisdocu-ment.XN.NF.F00-765 (1/Bi ANF-87-125 Revision1TABLEOFCONTENTSSection~Pae

1.0INTRODUCTION

~t~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~~~~t~~~~~~~~~~~~~~~~~12.0SUMMARYo~~~~~~~~~~~~~~o~~~~~~~~~~~~~~~~~t~~~~~~~~~~~~~~~~~~~~~~~~~23.0TRANSIENT ANALYSISFORTHERMALMARGIN..............................

53.13.23.2.13.2.23.2'3.3DesignBasss.......................

Anticipated Transients.............

LoadRejection WithoutBypass......

Feedwater Controller Failure.......

LossOfFeedwater Heating..........

Calculational, Model................

1~~~~~~~~~~~~~I~~~t~~~~~~~~t~~~~5~~~~~~~~~t~\~~~~~~~~~~~~~~~~~~~6~~~~t~~~~~~~~~~~~~~~~~~~~~~~~~~6~~~~~~~~~~~~~~~~~~~t~~~7~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~t~~~~~~~~~~83.4SafetyLimit.........

~~~~~~~~~~~~~I~~~4~~~~~~~~~~~~~~~~~~~9MAXIMUMOVERPRESSURIZAT ION...................................

22D0esignBases...................................................

224.2Pressurization Transients.

~...............

~~~~~224.2.1ClosureOfAllHainSteamIsolation Valves.......,.

235.0RECIRCULATION PUMPRUN-UP.........................................

2

46.0REFERENCES

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

~~~~~~~~t~~~~~~~~t~~~~~~~~~~~~~~~~~26APPENDICES A.SINGLELOOPOPERATION..........,.......'........................

..A-1B.HCPRSAFETYLIMIT..............................

..........B-l gt

~~11ANF-87-125 Revision1LISTOFTABLESTable2.13.13.23.33.4A.lTransient AnalysisResultsAtDesignBasisConditions...

ReactorDesignAndPlantCondi.tions Susquehanna Unit2..otSignificant Parameter ValuesUsedInTheAnalysisForSusquehanna Unit2......................................

ResultsOfSystemPlantTransient Analyses..............

Feedwater Controller FailureAnalysisResultsAt100%FlSLOReactorAndPlantConditions...:....................

~Pae~~~~~~~~~~4~~o~~~~~~~10o~~~~~~~~~llo~~~oo~~~~14owo~~~~~~o15~~~~~~~~~~LISTOFFIGURESFiciur'e3.13.23.33.4LoadRejection WithoutBypass.............................

LoadRejection WithoutBypass..........................

Feedwater Controller Failure......

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

~Pae161718193.53.65.1A.lA.2A.3B.3-1B.3-2B.3-3LossOfFeedwater Heating..

~...........................

LossOfFeedwater Heating..................

Susquehanna Unit2Cycle3ReducedFlowMCPROperating SingleLoopOperation

-PumpSeizure.......

SingleLoopOperation

-PumpSeizure..

..CorePowerVersusCoreFlow...........

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

Susquehanna Unit2Cycle3DesignBasisRadialPowerHiDesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-29x9Fuel...............,.

DesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-19x9Fuel..202125~~~~~~~~~~o~~~~~~~~~~~A9~~~~~~~~~~~A10stogram....

B-4B-5~~~~~~~~~

0~~illANF-87-1P RevisionLISTOFFIGURES(Continued)

~FiereParcaeB.3-4DesignBasisLocalPowerDistribution GeneralElectric(Central) 8xSRFuel..............................

B-7B.3-5DesignBasisLocalPowerDistribution GeneralElectric(Peripheral)

SxSRFuel...........................

B-8 ANF-87-125 Revision1.

1.0INTRODUCTION

ThisreportpresentstheresultsofAdvancedNuclearFuelsCorporation's*

evaluation ofsystemtransient eventsforSusquehanna Unit2Cycle'operation.

Theevaluation togetherwithcoretransient eventsdetermines thenecessary thermalmargin(HCPRlimits)toprotectagainsttheoccurrence ofboilingtransition duringthemostlimitinganticipated transient.

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

Theevaluation alsodemonstrates thevesselintegrity forthemostlimitingpressurization event.ThebasesfortheseanalyseshavebeenprovidedinReference l.'Formerly ExxonNuclearCompany(ENC).

f~'

ANF-87-125 Revision12.0SUMMARYTheSusquehanna Unit2Cycle3corecanbe.described asfollows:~FeelTeANFXN-2XN-2ANFXN-1GE8x8R8x8RNo.of~Ass140963241968BundleAverageEnrichment 3.33/9Gd4*

3.33/10Gd5 3.312.19e~UsingANF'smethodology andconsidering theCycle3core,themostlimitinganticipated plantsystemtransient withregardtothermalmarginatratedpowerandflowconditions wasconfirmed tobethegenerator loadrejection withoutbypass(LRWB)transient withrecirculation pumptrip(RPT)operable.

TheMinimumCriticalPowerRatio(MCPR)limitsforpotentially limitinganticipated plantsystemtransient eventsareshowninTable2.1forcomparison.

ThevaluesinTable2.1weredetermined assumingboundingconditions intheanalyses.

ResultswithRPToutofservicearereportedinSection3.2.1.Thesetransients wereevaluated withallco-resident fueltypesmodeledandthemostlimitingcondition wasusedtodetermine thereportedMCPRs.TheControlRodWithdrawal Error(CRWE)analysisandresulting deltaCPRresultsarereportedinReference 2.Maximumsystempressurehasbeencalculated forthecontainment isolation event,whichisarapidclosureofallmainsteamisolation valves,usingthescenarioasspecified bytheASMEPressureVesselCode.ThisanalysisshowsthatthesafetyvalvesofSusquehanna Unit2havesufficient capacityand*Thefirstnumberstatesthenumberofgadolinia rodsperbundleandthesecondnumberstatestheweightpercentgadolinia perrod.Thegadolinia concentrations andnumberofrodsperbundlearestatedforfreshfuelonly.Theothersarenotsignificant.

ANF-87-1Revisionperformance topreventthepressurefromreachingtheestablished transient pressuresafetylimitof110%ofdesignpressure(l.1x1250=1375psig).Theanalysisalsoassumedsixsafetyreliefvalvesoutofservice.Themaximumsystempressures predicted duringtheeventareshowninTable2.1.Resultsofthesingleloopoperation (SLO)analysisareshowninAppendixA.Thesafetylimitanalysisforsingleloopoperation supportsanincreaseintheHCPRSafetyLimitof.01.

ANF-87-125 Revision1TABLE2.1TRANSIENT ANALYSISRESULTSATDESIGNBASISCONDITIONS*

hCPRMCPR**Transient LoadRejection WithoutBypasswithRecirculation PumpTripFeedwater Controller FailurewithBypassLossofFeedwater HeatingANF9x90.24/1.30 0.23/1.29 0.16/1.22 GE8x8R0.21/1.27 0.20/1.26 0.15/1.21 MaximumPressuresiTransient MSIVClosureVesselDome1281VesselLowerPlenum1297SteamLine1284*104%power/100%

flow.*BasedontheMCPRSafetyLimitof1.06confirmed herein.

pW ANF-87-125 RevisionI3.0TRANSIENT ANALYSISFORTHERMALMARGIN3.1DesinBasisConsistent withtheFSARplanttransient

analysis, thermalmarginoperating MCPRlimitsaredetermined basedonthe104%power/100%

flowoperating point.Thisthermalmarginoperating MCPRlimitisthenmodifiedasafunctionofpowerandflowasrequiredtoprotectagainstboilingtransition resulting fromanticipated transients occurring fromallowedconditions onthepower/flow operating map.Theplantconditions forthe.104%power/100%

flowpointareasshowninTable3.1.ThemostlimitingpointinCycle3hasbeendetermined tobeatthemaximumCycle3licensing exposurelimitwhencontrolrodsarefullywithdrawn fromthecore.Thethermalmarginlimitestablished forthisexposurecondition isconservative forcaseswherecontrolrodsare~~artiallyinserted.

Following requirements established inthePlantOperating Licenseandassociated, Techn'ical Specifications, observance ofaMCPRoperating limitof1.30forANF9x9fueland1.27forGE8x8Rfuelorgreaterconservatively protectsagainstboilingtransition duringanticipated plantsystemstransients fromdesignbasisconditions forSusquehanna Unit2Cycle3.Thecalculational modelsusedtodetermine thermalmarginincludeANF'splanttransient andcorethermal-hydraulic codesasdescribed inpreviousdocumentation(I~

).Fuelpellet-to-clad gapconductances usedintheanalyseswerebasedoncalculations withRODEX2().Table3.2summarizes thevaluesusedforimportant parameters thatprovidedaboundinganalysis.

Recirculation PumpTrip(RPT)flowcoastdown wasinputbasedonmeasuredSusquehanna Unit2startuptestdata.Toconfirmtheneutronics asrequested bytheSERissuedforthesupplements ofReference l(8),theSusquehanna systemtransient modelwasbenchmarked toappropriate Susquehanna Unit2startuptestdata.XCOBRA-T(

)wasusedtocalculate thechangeincriticalowerratio(deltaCPR)forpressurization'vent analyses.

i ANF-87-1Revision3.2AnticiatedTransients ANFconsiders eightcategories ofpotential systemtransient occurrences forJetPumpBWRsinXN-NF-79-71(1~

).Thethreemostlimitingtransients aredescribed hereindetailtoshowthethermalmarginforCycle3ofSusquehanna Unit2.Thesetransients are:LoadRejection WithoutBypass(LRWB)Feedwater Controller Failure(FWCF)LossofFeedwater-Heating (LFWH)Asummaryofthetransient analysesisshowninTable3.3.Otherplanttransient eventsareinherently nonlimiting orclearlyboundedbyoneoftheaboveevents.3.2.1LoadRe'ectionWithoutBassThiseventisthemostlimitingoftheclassoftransients characterized byrapidvesselpressurization.

Thegenerator loadrejection causesaturbinecontrolvalvetrip,whichinitiates areactorscramandarecirculating pump~trip(RPT).Thecompression waveproducedbythefastcontrolvalveclosuretravelsthroughthesteamlinesintothevesselandcreatesthevesselpressurization.

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

Theboundingassumptions areconsistent withANF'sCOTRANSAcodeuncertainties analysismethodology asreportedinReference 8andapprovedbytheNRC.Theboundingassumptions include:

ANF-87-125 Revision1Technical Specification minimumcontrolrodspeedTechnical Specification maximumscramdelaytimeIntegralpowerincreased by10%Atdesignbasisconditions (104%power/100%

flow)thisresultsinadeltaCPRof0.24fortheloadrejection withoutbypasswhenRPTis'perable forANF9x9fuels.Thecorresponding deltaCPRforGE8xSRfuelis0.21.Theloadrejection withoutbypasseventwasalsoanalyzedatthedesignbasisconditions whenRPTisnotoperable.

Theresulting deltaCPR'sare0.37and0.32forANF9x9andGESxSRfuels,respectively.

3.2.2Feedwater Controller Failure~~Failureofthefeedwater 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 withboundingvaluesandresultedinadeltaCPRof0.23forANF9x9fuelsand0.20forGESx8Rfuel.Figures3.3and3.4presentkeyvariables forthisfeedwater controller failureevent.Thiseventwasalsoexaminedforreducedpowerconditions atfullflow.TheresultsfortheFWCFtransients fromreducedpowerconditions areshowninTable3.4forall9x9andSx8fuels.hecalculated resultsshowthatFWCFdeltaCPR'svarywithdecreasing power 8ANF-87-1Revisionatfullflowconditions.

ThehighestdeltaCPRwascalculated at40%powerand100%flowconditions.

Thistransient eventatfullpowerandfullflowconditions wasalsoanalyzedassumingboundingconditions andfailureofthebypassvalvestoopen.ThisresultsinadeltaCPRof0.28forANF9x9fuelsand0.25forGE8x8Rfuel.3.2.3ossOfFeedwater HeatinThelossoffeedwater heatingleadstoagradual.increaseinthesubcooling ofthewaterinthereactorlowerplenum.Reactorpowerslowlyrisestothethermalpowermonitorsystemtripsetpoint.

Thegradualpowerchangeallowsfuelthermalresponsetomaintainpacewiththeincreaseinneutronflux.Usingthemethodology ofReference 1thedeltaCPRfortheeventinCycle3is0.16forANF9x9fueland0.15forGE8x8Rfuel.Figures3.5and3.6depikeyvariables forthelossoffeedwater heatingevent.Thebypassvalvesdonotsignificantly affectthelossoffeedwater heatingresults.Thus,thedeltaCPRlimitisapplicable whetherthebypassvalvesareoperableornot.3.3Calculational ModelTheplanttransient codeusedtoevaluatethegenerator loadrejection andfeedwater flowincreasewasANF'scodeCOTRANSA(

).Theaxialone-dimensional neutronics modelpredicted reactorpowershiftstowardthecoremiddleandtopaspressurization occurred.

Thiswasaccounted forexplicitly indetermining thermalmarginchangesinthetransient.

Thelossoffeedwater heatingeventwasevaluated withPTSBWR3andXCOBRA(Reference 1).AppendixA(1)oftheSusquehanna Unit1Cycle2analysi's delineates thechangesmadetoCOTRANSA(1) tomergethePTSBWR3codewiththeCOTRANSAcode,torefinenumerical techniques andtoimproveinput.Reference 9describes theXCOBRA-Tcodeustocalculate thedeltaCPR'sforthepressurization transients.

AppendixBo 9ANF-87-125 Revision1Reference 10delineates theplantrelatedchangesmadetothesecodesfortheSusquehanna Units1and2analyses.

3.4SafetLimitThesafetylimitistheminimumvalueofthecriticalpowerratio(CPR)atwhichthefuelcouldbeoperatedwheretheexpectednumberofrodsinboilingtransition wouldnotexceed0.1%ofthefuelrodsinthecore.ThesafetylimitistheHCPRwhichwouldbepermitted tooccurduringthelimitinganticipated operational occurrence.

AHCPRsafetylimitof1.06forallfueltypesinSusquehanna Unit2Cycle3wassupported bythemethodology presented inReference 3.Theinputparameters anduncertainties usedtosupportthesafetylimitarepresented inAppendixBofthisreport.

10ANF-87-l~

Revision~

TABLE3.1REACTORDESIGNANDPLANTCONDITIONS SUSQUEHANNA UNIT2ReactorThermalPower(104%)'otal CoreFlow(1005)CoreIn-Channel FlowCoreBypassFlowCoreInletEnthalpyVesselPressures SteamDomeUpperPlenumCoreLowerPlenumTurbinePressureFeedwater/Steam FlowFeedwater EnthalpyRecirculation PumpFlow(perpump)3439Mwt100.0Mlb/hr89.9Mlb/hr10.1Mlb/hr518.0Btu/ibm1035psia1045psia1052psia1066psia975psia14.15Mlb/hr360.8Btu/ibm15.75Mlb/hr ANF-87-125 Revision1TABLE3.2SIGNIFICANT PARAMETER VALUESUSEDINTHEANALYSISFORSUSQUEHANNA UNIT2HighNeutronFluxTripControlRodInserti'on TimeControlRodWorthVoidReactivity FeedbackTimetoDeenergized PilotScramSolenoidValvesTimetoSenseFastTurbineControlValveClosureTimefromHighNeutronFluxTimetoControlRodMotionTurbineStopValveStrokeTimeTurbineStopValvePositionTripTurbineControlValveStrokeTime(Total)Fuel/Cladding GapConductance CoreAverage(Constant)

Safety/Relief ValvePerformance SettingsReliefValveCapacityPilotOperatedValveDelay/Stroke 125.3%3.49sec/90%insertednominalnominal200msec(maximum) 30msec290msec100msec90%open70msec758.0Btu/hr-ft2-F Technical Specifications 225.4ibm/sec(1110psig)400/150msec 12ANF-87-1RevisionTABLE3.2SIGNIFICANT PARAMETER VALUESUSEDINTHEANALYSISFORSUSQUEHANNA UNIT2(CONTINUED)

MSIVStrokeTimeMSIVPositionTripSetpointTurbineBypassValvePerformance TotalCapacityDelaytoOpening(80%open)FractionofEnergyGenerated inFuelVesselMaterLevel(abovei'nstrument zero)HighLevelTripNormalLowLevelTripMaximumFeedwater RunoutFlowThreePumpsRecirculation PumpTripSetpointVesselPressure3.0sec90%open936.11ibm/sec300msec0.96558.7in35in*8in5049ibm/sec1170psig*COTRANSA plotsaregivingwaterlevelaboveseparator skirtandthevalhereisaboveinstrument zero.

13ANF-87-125

.Revision1TABLE3.2SIGNIFICANT PARAMETER VALUESUSEDINTHEANALYSISFORSUSQUEHANNA UNIT2(CONTINUED)

ControlCharacteristics SensorTimeConstants PressureOthersFeedwater ControlModeFeedwater MasterController Proportional GainResetRateFeedwater 100%MismatchWaterLevelErrorSteamFlowEquivalent FlowControlModePressureRegulator SettingsLeadLagGain500msec250msecThree-Element 500(%/%)(%/ft)1.70(%/sec/ft) 4.0ft4034ft/100%Manual3.0sec7.0sec3'3%/psid ANF-87-1RevisionTABLE3.3RESULTSOFSYSTEMPLANTTRANSIENT ANALYSESEventLoadRejection WithoutBypassFeedwater Controller FailureLossofFeedwater HeatingMSIVClosurewithFluxScramMaximumNeutronFlux%Rated267233123342Maximum-CoreAverageHeatFlux%Rated116.2116.8121.3133.2MaximumSystemPressure~sia1194117910781312hCPRFor9x9Fuels0.240.230.16NOTE:Alleventsareboundingcaseat104%power/100%

flow.

15ANF-87-125 Revision1'ABLE3.4FEEDWATER CONTROLLER FAILUREANALYSISRESULTSAT100%FLOW-%PowerDelt'aCPR1048040ANF9x90.230.250.280.31GE8x8R0.200.230.260.28 l,.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATION FLOW4.VESSELSTEAMFLOW5.FEEDWATER FLOWCIOlCl~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,SEC1.51.72.02.22.5Figure3.2LoadRejection Without8ypass i.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATION FLOW4.VESSELSTEAMFLOW5.FEEDWATER FLOW55<24i412i620TIME,SEC2428323640figure3.3Feedw.ontrol1erFailure IDCtCUi.VESSELPRESSURECHANGE(PSI)2.VESSELWATERLEVEL(IN)CIOJoP121620TIME,SEC2428323640Figure3.4Feedwater Controller Failure ClCl124545123i.NEUTRON.FLUXLEVEL2.HEATFLUX3.RECIRCULATION FLOW4.VESSELSTEAMFLOW5.FEEDWATER FLOW33I-OtOC3IXUJ0C)I21t010-2030405060TIME,SEC708090100Figure3.5LosseedwaterHeating i.VESSELPRESSURECHANGE(PSI)2.VESSELWATERLEVEL(IN)22LAILLICILAILAIIQ1020304050TIME,SEC60708090100Figure3.6LossOfFeedwater Heating lKV!-.$1r~0 22ANF-87-125" Revision14.0MAXIMUMOVERPRESSURIZATION Maximumsystempressurehasbeencalculated forthecontainment isolation event(rapidclosureofallmainsteamisolation valves)withanadversescenarioasspecified bytheASHEPressureVesselCode.ThisanalysisshowedthatthesafetyvalvesofSusquehanna Unit2havesufficient capacityandperformance topreventpressurefromreachingtheestablished transient pressuresafetylimitof110%ofthedesignpressure(1375psig).Themaximumsystempressures predicted duringtheeventareshowninTable2.1.Thisanalysisassumedsixsafetyreliefvalvesoutofservice.4.1Desin'asisThereactorconditions usedintheevaluation ofthemaximumpressurization ventarethoseshowninTable3.1.Themostcriticalactivecomponent (scramonHSIVclosure)wasassumedtofailduringthetransient.

Thecalculation wasperformed withANF'sadvancedplantsimulator codeCOTRANSA(

),whichincludesanaxialone-dimensional neutronics model.4.2Pressurization Transients ANFhasevaluated severalpressurization eventsandhasdetermined thatclosureofallHainSteamIsolation Valves(HSIVs)withoutdirectscramisthemostlimiting.

-Although theclosurerateoftheHSIVsissubstantially slowerthantheturbinestopvalvesorturbinecontrolvalves,thecompressibility oftheadditional fluidinthesteamlinesresultsinalessseveretransient forthefasterturbinestop/control valveclosuretransients.

Essentially, therateofsteamvelocityreduction isconcentrated towardtheendofthevalvestroke,generating asubstantial compression wave.Oncethecontainment isisolatedthesubsequent corepowerproduction mustbeabsorbedinasmallervolumethanifaturbinetriphadoccurred.

Calculations havedetermined thattheoverallresultistocauseisolation (HSIV)closurestobemorelimitingorsystempressurethanturbinetrips.

23ANF-87-1P Revision'.2.1ClosureOfAllHainSteamIsolation ValvesThiscalculation assumedthatsixreliefvalveswereoutofserviceandthatallfoursteamisolation valveswereisolatedatthecontainment boundarywithin3seconds.Atabout3.0seconds,thereactorscramisinitiated byreachingthehighfluxtripsetpoints.

Sincescramperformance wasdegradedtoitsTechnical Specification limit,effective powershutdownisdelayeduntilafter4.4seconds.Substantial thermalpowerproduction enhancespressurization.

Pressures reachtherecirculation pumptripsetpoint(1170psig)beforethepressurization isreversed.

Lossofcoolantflowleadstoenhancedsteamproduction aslesssubcooled waterisavailable toabsorbcorethermalpower.Themaximumpressurecalculated inthesteamlineswas1284psigoccurring nearthevesselatabout6.5seconds.Themaximumvesselpressurewas1297psigoccurring inthelowerplenumatabout6.3seconds.

ANF-87-125 Revision15.0RECIRCULATION PUMPRUN-UPAnalysisofpumprun-upeventsforoperation atlessthanratedrecirculation pumpcapacitydemonstrates theneedforanaugmentation ofthefullflowMCPRoperating limitforlowerflowconditions.

Thisisduetothepotential forlargereactorpowerincreases shouldanuncontrolled pumpflowincreaseoccur.Thissectiondiscusses pumpexcursions whentheplantisinmanualflowcontroloperation mode.Resultsobtainedfrompreviousanalysesshowedthetwopumprun-upboundsthesinglepumprun-up.Onlythetwopumprun-upisevaluated forSusquehanna Unit2Cycle3.TheseresultsindicatethatHCPRwoulddecreasebelowthesafetylimitifthefullflowreference MCPRisobservedatinitialconditions.

Thus,anaugmented HCPRisneededforpartialflowoperation topreventviolation oftheHCPRSafetyLimitforthetwopump~~xcursionevent.Theanalysisofthetwopumpflowexcursion indicates thatthelimitingeventisagradualpowerincreaseinwhichtheheatfluxtrackspower.TheSusquehanna Unit2Cycle3analysisconservatively assumedtherun-upeventinitiated at57%power/40%

flowandreached111%ratedpowerat100%ratedflow.Theeventterminated at105%ofratedflowwithaminimumCPRof1.06.Theresultsofthetwopumprun-upanalysesformanualflowcontrolarepresented inFigure5.1.ThecyclespecificHCPRlimitforSusquehanna Unit2Cycle3shallbethemaximumofthereducedflowMCPRoperating limit,thefullflowHCPRoperating limit,orthepowerdependent HCPRoperating limit.

1.50ANF9X9FUELS----GE8X8RFUEL1.401.301.20~Luo4060BO708090100TOTALCORERECIRCULATION FLOW(%RATED)Figure5.1Susquehanna Unit2.3ReducedFlowHCPROperating Limit 26ANF-87-125 Revision16.0.REFERENCES R.H.Kelley,"ExxonNuclearPlantTransient Methodology forBoilingIltRt,"~XN-NF.I-II, RII2,AddNIICorporation*,

Richland, WA99352,November1981.2.3.4.5.,7.8.9.10.J.A.White,"Susquehanna Unit2Cycle3ReloadAnalysis, DesignandSafetyAnalyses, "8NF-87-126,

.AdvancedNuclearFuelsCorporation,

Richland, WA99352,October1987.J.A.White,"ExxonNuclearMethodology forBoilingWaterReactors, THERMEX:ThermalLimitsHethodology, SummaryDescription,"

XN-NF ~19PA,Volume,3,Revision2,AdvancedNuclearFuelsCorporation,

Richland, WA99352,January1987.T.W.Patten,"ExxonNuclearCriticalPowerMethodology forBoilingWaterR,"~52-525A, R11I,AddNIFIC5Richland, WA99352,November1983.T.H.Keheley,"Susquehanna Unit2Cycle2PlantTransient Analysis,"

XN-NF-86-55, Revision1,AdvancedNuclearFuels'orporation,

Richland, WA99352,Hay1986.T.H.Keheley,"Susquehanna Unit1Cycle4PlantTransient Analysis,"

XN-NF-87-22, AdvancedNuclearFuelsCorporation,

Richland, WA99352,April1987.K.R.Herckx,"RODEX2FuelRodMechanical ResponseEvaluation Model,"XN-~NF.I-N,R II2,AddllIFIC5tl,Illlld.,IIA99352,April1984.S.E.Jensen,"ExxonNuclearPlantTransient Methodology forBoilingIltRt,"~XN-II-I-I,RIII,AdvancedNuclearFuelsCorporation,

Richland, WA99352,March1986.H.J.Ades,"XCOBRA-T:

AComputerCodeforBWRTransient Thermal-Hydraulic CoreAnalysis,"

XN-NF-84-105 PA,Volume1&Volume1,Supp.18Supp.2,AdvancedNuclearFuelsCorporation,

Richland, WA99352,February1987.T.H.Keheley,"Susquehanna Unit1Cycle2PlantTransient Analyses,"

XN-NF-84-118, including Supplement 1,AdvancedNuclearFuelsCorporation,

Richland, WA99352,December1984.FormerlyExxonNuclearCompany(ENC).

I A-IANF-87-125 RevisionIAPPENDIXASINGLELOOPOPERATION TheNSSSsupplierhasprovidedanalyseswhichdemonstrate thesafetyofplantoperation withasinglerecirculation loopoutofserviceforanextendedperiodoftime.Theseanalysesrestricttheoveralloperation oftheplanttolowerbundlepowerlevelsandlowernodalpowerlevelsthanareallowedwhenbothrecirculation systemsareinoperation.

=Thephysicalinterdependence betweencorepowerandrecirculation flowrateinherently limitsthecoretolessthanratedpower.ANFfuelwasdesignedtobecompatible withtheco-residentfuelinthermalhydraulic, nuclear,andmechanical designperformance.

TheANFmethodology hasgivenresultswhichareconsistent withhoseofthepreviousanalysesfornormaltwo-loopoperation.

Manyanalysesperformed bytheNSSSsupplierforsingleloopoperation arealsoapplicable tosingleloopoperation withfuelandanalysesprovidedbyANF.Adiscussion oftherelevanteventsandlimits.forsingleloopoperation follow.AlsoincludedareresultsofANFanalyseswhichconfirmtheNSSSvendorconclusions.

A.lABNORMALOPERATING TRANSIENTS MCPRlimitsestablished forfullflowtwoloopoperation areconservative forsinglelooptransients becauseofthephysicalphenomena relatedtopart-power part-flow operation, notbecauseoffeaturesinreactoranalysismodelsorcompatible fueldesigns.AreviewofthemostlimitingdeltaCPRtransients forsingleloopoperation wasconducted.

Undersingleloopconditions, steadystateoperation cannotexceedapproximately 76%powerand61%coreflowbecauseofthecapability oftherecirculation looppump.Thus,theMCPR~~limitatmaximumpowerishigherthanthetwopumpoperating MCPRlimitduetotheflowdependent MCPRfunction.

Thisflowdependence isbasedonaflow A-2ANF-87-1~

Revision~

increasetransient fromrunup'ftwopumps.Flowrunupsfromasinglerecirculation pumpwouldbemuchlesssevere,thoughtheconservative twopumplimitisretained.

LoadRe'ectionWithoutBassThelimitinganticipated systemtransient fortheSusquehanna UnitsistheLoadRejection WithoutBypass(LRWB)pressurization transient.

Inthistransient, theprimaryphenomena isthepressurization causedbyabruptly'toppingthesteamflowthroughrapidclosureqftheturbinecontrolvalve.Whentherapidpressurization reachesthecoreitcausesapowerexcursion duetovoidcollapse.

ThereducedpowerandflowanalysesfortheSusquehanna Unitsdescribed inReference A-1undertwoloopoperation showsthattheresulting powexcursion andassociated deltaCPRarereducedbelowthoseofthefulpower/full flowcase.ThusfortheSusquehanna UnitstheHCPRlimitsbasedonLRWBanalysesatfullpowerareconservatively applicable tothelowerpowers/flows associated withsingleloopconditions.

Furthermore, LRWBanalysesbyANFatreducedpowerandflowconditions inotherBWR'swithsingleloopoperation confirmthistrend.A.1.2Feedwater Controller FailureThesecondworstlimitingtransient atfullpowerandflowistheFeedwater

'ontroller Failure(FWCF)tomaximumdemand.Thistransient isalsolesssevereatthepowerandflowconditions associated withsingleloopoperation.

Thistransient assumesthefeedwater controller failstomaximumdemandandresultsinthemaximumamountofsubcooled feedwater inthedowncomer.

Whenthiscoolerwaterreachesthecorethepowerrises.Thecorepowerriseisterminated byareactorscraminitiated byaturbinetrip.Theturbinetr' A-3ANF-87-125 RevisionIistheresultofthehighwaterleveltripcausedbytheadditional amountoffeedwater beinginjected.

IAtthereducedrecirculation flows,thesubcooling inthedowncomer duetothehighfeedwater flowtakeslongertotransverse thecoresothatahighwaterleveltripoccurs.beforecorepowercanriseashighasitdoesinthefullflowcase.AswiththeLRWB,thepressurization eventresulting fromtheturbinetripislesssevereforthereducedpowerinSLO..Thus,becauseoftheslowerenthalpytransport phenomena causedbythelowerrecirculation flowandbecauseofthelowersteamlineflowinthepressurization portionofthetransient, theFWCFhaslargermargintotheoperating limitinsingleloopoperation thanintwoloopoperation.

.1.3PumSeizureAccidentPumpseizureisapostulated accidentwheretheoperating recirculation pumpsuddenlystopsrotating.

Thiscausesarapiddecreaseincoreflow,adecreaseintherateatwhichheatcanbetransferred fromthefuelrodsandadecreaseinthecriticalpowerratio.AnalyseswithCOTRANSAandXCOBRA-TshowthatforCycle3theCPRforANFfuelwould.decreaseby0.30duringapumpseizureforsingleloopoperation.

TheCOTRANSAcodewasusedtosimulatesystemresponsetoapumpseizureinsingleloopoperation fromtheconditions specified inTableA.l.Theoperating recirculation pumprotorwasstoppedin0.1secondscausingasuddendecreaseinactivejetpumpdriveflow.Atabout6.7secondstheinactivejetpumpdiffuserflowwentfromnegativeflowtopositiveflow.In7.3secondsthedomepressuredecreased toaminimumvalueof970.3psiaandthenstartedtoincreaseagain.FiguresA.IandA.2presentagraphical representation ofimportant systemparameters duringthetransient.

A-4ANF-87-1, RevisionThedeltaCPRforthiseventwascalculated usingXCOBRA-T.

TheANF9x9fuelreachedamaximumdeltaCPRof0.30at2.2secondsintothetransient.

TheGE8x8RfuelreachedamaximumdeltaCPRof0.29at2.15secondsintothetransient.

A.1.4MCPRSafetLimitForsingleloopoperation, theNSSSvendorfoundthatanincreaseof0.01intheMCPRsafetylimitwasneededtoaccountfortheincreased flowmeasurement uncertainties andincreased TIPuncertainties associated withsinglepumpoperation.

ANFhasevaluated theeffectsoftheincreased flowmeasurement uncertainties onthesafetylimitMCPRandfoundthattheNSSSvendordetermined increaseintheallowedsafetylimitMCPRisalsoapplicable toANFfuelduringsingleloopoperation.

Thus,increasing thesafetylimitMCPRby0.01forsingleloopoperation (1.07)withANFfuelissufficient conservative toalsoboundtheincreased flowmeasurement uncertainties fosingleloopoperation.

A.1.5~SummarThelimitingMCPRoperating limitForsingleloopoperation isconservatively setusingthelimitingpumpseizureaccidentdeltaCPRplusthesingleloopoperation MCPRsafetylimit.ThislimittogetherwiththeMCPRfcurvefortwoloopoperation plus.01andtheMCPRpcurvefortwoloopoperation plus.Olconservatively boundalltransients.

A-5ANF-87-125 Revision1A.2MAPLHGRLIMITSANFperformed LOCAanalysesforsingleloopconditions anddetermined thattheMAPLHGRlimitcurvefortwo-loopoperation isalsoapplicable tosingleloopoperation forANFfuels(Reference A-2).

A-6ANF-87-1RevisionA.3STABILITY

.TheTechnical Specifications requireAPRM/LPRM surveillance totheleftofthe45%ConstantFlowlineandabovethe80%RodBlockline.Basedoncorehydrodynamic stability analysesforCycle3,operation atpower/flow combinations aboveandtotheleftofthelineconnecting the66%Power/45%

Flowand69%Power/47%

FlowpointsneedstobeaddedtotheAPRM/LPRM surveillance requirements.

FigureA.3showsthecorepowerversuscoreflowestablished forCycle3.

A-7ANF-87-]25 Revision1TABLEA.1SLOREACTORANDPLANTCONDITIONS ReactorThermalPowerTotalRecirculation FlowCoreBypassFlowCoreInletEnthalpyVesselPressures SteamDomeLowerPlenumTurbinePressureSteamFlowFeedwater Enthalpy2489HMt60.35Hlb/hr5.70Hlb/hr507.3Btu/lb994.5psia,1011.3psia965.4psia9.8Hlb/hr330.7Btu/lb l.NEUTRONFLUXLEVEL2.HEATFLUX3.RECIRCULATION FLOW4.VESSELSTEAMFLOW5.FEEOWATER FLOW5iioTIME,SEC14i6iB20FigureA.lSingleLeration-PumpSeizure CI1.VESSELPRESSURECHANGE(PSI)2.VESSELHATERLEVEL(IN)CUOIQ10TIME,SEC1214161820FigureA.2SingleLoopOperation

-PumpSeizure A-10ANF-87-125 Revisi'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-87-125 Revision1A.4A-1.REFERENCES J.C.Chandler, "Susquehanna Unit1Cycle3ReloadAnalysis,"

XN-NF-85-132,Revision1,AdvancedNuclearFuelsCorporation,

Richland, WA99352,December1985.A-2.':R.Swope,"Susquehanna LOCAAnalysisforSingleLoopOperation,"

XN-NF-86-125, AdvancedNuclearFuelsCorporation,

Richland, WA99352,November1986.A-3.K.D.Hartley,etal.,"Susquehanna Unit2Cycle2Stability TestResults,"

XN-NF-86-90, Supplement 1,AdvancedNuclearFuelsCorporation,

Richland, WA99352,January1987.

0 B-1ANF-87-125 Revision1APPENDIXBMCPRSAFETYLIMITB.1INTRODUCTION TheHCPRfuelcladdingintegrity safetylimitwascalculated usingthemethodology anduncertainties described inReference B.1.Inthismethodology, aHonteCarloprocedure isusedtoevaluateplantmeasurement andpowerpredictions uncertainties suchthatduringsustained operation attheHCPRCladdingIntegrity SafetyLimit,atleast99.9%ofthefuelrodsinthecorewouldbeexpectedtoavoidboilingtransition.

Thisappendixdescribes thecalculation andpresentstheanalytical results.i B-2ANF-87-1RevisionB.2CONCLUSIONS Duringsustained operation ataHCPRof1.06withthedesignbasispowerdistribution described below,atleast99.9%ofthefuelrodsinthecoreareexpectedtoavoidboilingtransition ataconfidence levelof95%.

B-3ANF-87-125 Revision1B.3DESIGNBASISPOWERDISTRIBUTION Predicted powerdistributions wereextracted fromthefuelmanagement analysisforSusquehanna Unit2Cycle3.Theradialpowerdistributions wereevaluated forperformance asthedesignbasisradialpowermap,andthedistribution at8,000HWd/HTUexposurewasselectedasthemostsevereexpecteddistribution forthecycle.Thedistribution wasskewedtowardhigherpowerfactorsbytheadditionofbundleswitharadialpeakingfactorapproximating anoperating HCPRlevelof1.32atfullpower.Theresulting designbasisradialpowerdistribution isshowninFigureB.3-1.Thefuelmanagement analysisindicated thatthemaximumpowerANFbundle(XN-2)inthecoreattheend-of-cycle exposure(10;829.6 HWd/HTU)waspredicted tobeoperating atanexposurelevelof14,319HWd/HTU,soalocalpower~~istribution typicalofanodalexposureof15,000HWd/HTUwasselectedasthedesignbasislocalpowerdistribution.

Uncontrolled localpowerpeakingdistributions forbothANF9x9XN-29GD4%fueland10GD5%fuelwere'eviewed.

Thelimitinglocalswerefoundtooccurat15,000HWd/HTUfor9GD4%fuel.Thisdistribution isshowninFigureB.3-2.Localpowerdistributions fortwo'ndthreecycleirradiated fuelwerechosenconservatively forANFXN-1andGE8x8Rfuel,andareshowninFiguresB.3-3andB.3-4.Aboundingly flatlocalpowerdistribution wasselectedfortheGeneralElectricfuelintheperipheral lowpowerregion.Thisdistribution isshowninFigureB-3.5.Thelimitingaxialpowerprofileselectedforthe8,000HWd/HTUstatepoint ofCycle3wasconservatively selectedbasedonestablished criteria.

8070605000C)So2010000.20.00.60.8'RRDIRLPOWERPERKING1.2SCAotAQ)O6FigureB.3-1Susquehanna Unitle3DesignBasisRadialPowerHistogram 8-5ANF-87-125 Revision1*~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.3-2DesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-29X9Fuel B-6ANF-87-17 Revision'~*: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.3-3DesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-19X9Fuel B-'7ANF-87-125 Revision1*~*~*~**~*~*~*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'erDistribution GeneralElectric(Central) 8XSRFuel B-S,ANF.-'87-1 Revision,*~*: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.3-5DesignBasisLocalPowerDistribution GeneralElectric(Peripheral)

SXSRFuel B-9ANF-87-125 Revision1B.4CALCULATION OFTHENUMBEROFRODSINBOILINGTRANSITION Themethodology ofReference 8-1wasusedtoanalyzethenumberoffuelrodsinboilingtransition.

TheXN-3correlation(B

)wasusedtopredictcriticalheatfluxphenomena.

FivehundredMonteCarlotrialswereperformed tosupporttheMCPRsafetylimit.Non-parametric tolerance limits(B)wereusedinlieuofPearsoncurvefitting.Theuncertainties usedintheanalysisfornormalconditions werethoseidentified inReference B-l.Atleast99.9%ofthefuelrodsinthecoreareexpectedtoavoidboilingtransition withaconfidence levelof95%.

B-10ANF-87-'1 RevisionB.5B-1.B-2.B-3.REFERENCES T.W.Patten,"ExxonNuclearCriticalPowerMethodology forBoilingIltRt,"~XN-NptA,R11I,AddNIFCorporation,

Richland, WA99352,November1983.R.B.MacduffandT.W:Patten,"TheXN-3CriticalPower5Iti,"X~N-Np-51 A,R11I,dRppltI,AddNuclearFuelsCorporation,

Richland, WA99352,October1982.PaulN.Somerville, "TablesforObtaining Non-Parametric Tolerance Limits,"AnnalsofMathematical Statistics, Vol.29,No.2(June1958),pp.599-601.

ANF-87-125 Revision1IssueDate:SUSQUEHANNA UNIT2CYCLE3PLANTTRANSIENT ANALYSISDistribution:

D.A.AdkissonD.J.BraunR.E.Collingham L.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.Williamson H.G.Shaw/PP8L (20)DocumentControl(5)