ML103560434

St. Lucie, Unit 1 - Extended Power Uprate Licensing Report, Attachment 5, Appendix C, ANP-2903(NP), Revision 0, EPU Cycle Realistic Large Break LOCA Summary Report with Zr-4 Fuel Cladding

ML103560434
Person / Time
Site:	Saint Lucie
Issue date:	02/28/2010
From:	AREVA NP
To:	Office of Nuclear Reactor Regulation
References
L-2010-259 ANP-2903(NP), Rev 000
Download: ML103560434 (99)

Text

St. Lucie Unit 1 L-2010- Docket No. 50-335 Attachment 5 St. Lucie Unit 1 App C-1 Realistic Large Break LOCA Summary Report St. Lucie Unit 1 Extended Power Uprate Licensing Report Attachment 5 Appendix C St. Lucie Nuclear Plant Unit 1 EPU Cycle Realistic Large Break LOCA Summary Report With Zr-4 Fuel Cladding ANP-2903(NP) Revision 000 Areva NP Inc.

AREVA NP Inc.

Copyright © 2010 AREVA NP Inc. All Rights Reserved Nature of Changes Contents Tables Figures

Nomenclature Q Introduction

Summary Table 2-1 Summary of Major Parameters for Limiting Transient Analysis (1)(2)(3)(4)(5)3.1 Description of the LBLOCA Event

3.2 Description

of Analytical Models Realistic Large Break LOCA Methodology

3.3 Plant

Description and Summary of Analysis Parameters

[ ] 3.4 SER Compliance

3.5 Realistic

Large Break LOCA Results Table 3-1 Sampled LBLOCA Parameters Phenomenological Plant1 Table 3-2 Plant Operating Range Supported by the LOCA Analysis Event Operating Range 1.0 Plant Physical Description [ ]2.0 Plant Initial Operating Conditions Table 3-2 Plant Operating Range Supported by the LOCA Analysis (Continued)Event Operating Range 3.0 Accident Boundary Conditions

Table 3-3 Statistical Distributions Used for Process Parameters1 Parameter Operational Uncertainty Distribution Parameter Range Table 3-4 SER Conditions and Limitations SER Conditions and Limitations Response Table 3-4 SER Conditions and Limitations (Continued)SER Conditions and LimitationsResponse Table 3-5 Summary of Results for the Limiting PCT Case Case # 3 1 Table 3-6 Calculated Event Times for the Limiting PCT Case Event Time (s)

Table 3-7 Heat Transfer Parameters for the Limiting Case Table 3-8 Containment Initial and Boundary Conditions Containment Net Free Volume (ft3)Initial ConditionsContainment SprayContainment Fan Coolers Table 3-9 Passive Heat Sinks in Containment1 Heat SinkArea (ft2) Thickness (ft) Material Material Properties Thermal Conductivity (BTU/hr-ft-oF) Volumetric Heat Capacity (BTU/ft3-oF)

Figure 3-1 Primary System Noding Figure 3-2 Secondary System Noding Figure 3-3 Reactor Vessel Noding Figure 3-4 Core Noding Detail Figure 3-5 Upper Plenum Noding Detail StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-26One-SidedBreakArea(ft'/side)1.02.03.04.05.0BurnTime(hours)I*****************************0.05000.010000.015000.0CorePower(MW)LHGR(KW/ft)ASIPressurizerPressure(psia)PressurizerliquidLevel(Ok)RCS(Tcold)Temperature('F)t.:.:'.,.:::._3028.03028.53029.03029.53030.03030.53031.0:..:.....:,.....J14.414.614.815.015.215.4.....,.--0.10.00.00.10.1t:....,..__12200.02220.02240.02260.02280.02300.0.-:.......62.063.064.065.066.0....-:-_.-...::548.0550.0552.0554.0Figure3-6ScatterPlotofOperationalParametersAREVANPInc.

SITPressure(psia)StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladding...:-_.11400150.0160.0170.0-_.11080.01100.01120.01140.01160.01180.0._---200.0220.0240.0260.02800300.0--_.-,***12.45e+062.50e+062.55e+062.60e+062.65e+06L:,:::,1110.0112.0114.0116.0118.0120.0122.0124.01260128.0130.0Figure3-6ScatterPlotofOperationalParameters(Continued)AREVANPInc.ANP-2903(NP)Revision000Page3-27 StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingPCTvsTimeofPCT20001800D1600DD.,*1400*DCL0,-",1200I-0D...010000800I*SplitBreakcloGuillotineBreak600ANP-2903(NP)Revision000Page3-28400a100200300TimeofpeT(5)400500Figure3-7PCTversusPCTTimeScatterPlotfrom59CalculationsAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingpeTvsOne-sidedBreakArea200018000***01600o.00**I*..*0*.80nO*0*,0*1400o0**0EcPO0I-12000D.-o10000800ANP-2903(NP)Revision000Page3-29600I*SplitBreakIoGuillotineBreakl5.0400L--_-----'--__---'I__-----'---__---'-I____...LI______'____________'1.02.03.04.0BreakArea(fe/side)Figure3-8PCTversusBreakSizeScatterPlotfrom59CalculationsAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingMaximumOxidationvspeT2.0ANP-2903(NP)Revision000Page3-301.81.61.41.2cg1.0ro"0'xo0.80.60.40.2I*SplitBreakIDGuillotineBreakl*D0.0400600'--'80010001200peT(oF)1400160018002000Figure3-9MaximumOxidationversusPCTScatterPlotfrom59CalculationsAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingTotalOxidationvspeTANP-2903(NP)Revision000Page3-310.08I*SplitBreakIDGuillotineBreakl0.06'{ic:0:;::;*co-0D'x0D0.04D*0.02II**D0.00IrnrnLJL-'120040060080010001400160018002000peTCF)Figure3-10TotalOxidationversusPCTScatterPlotfrom59CalculationsAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingPCTTraceforCase#3peT=1671.8of,atTime=26.63s,on6%GadRodANP-2903(NP)Revision000Page3-3220001500Q)'-:::::l+-'co'-Q)0..E1000Q)l-+-'c*0Q....c(f)Q)2:500oo200Time(s)400600Figure3-11PeakCladdingTemperature(IndependentofElevation)fortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingBreakFlow80--VesselSide----PumpSide---Total60ANP-2903(NP)Revision000Page3-33"'0""(j)40--E..0Q)....co0::50LL20o-20o200Time(5)40060010:22950160ec200915:31:36R50MXFigure3-12BreakFlowfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingCoreInletMassFlux1000--HotAssembly----SurroundAssembly---AverageCore---OuterCore500if)IE..0.........X::::lL!-V)V)co0dANP-2903(NP)Revision000Page3-34-500o200Time(s)40060010:22950160ec200915:31:36R50MXFigure3-13CoreInletMassFluxfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-35CoreOutletMassFlux900--HotAssembly----SurroundAssembly---AverageCore---OuterCore700500UJIE300.0><:::JLL(/)(/)100coI-100-300600400200-500L--__--'----__---'-___'___'__________'oTime(s)10:22950160ec200915:31:36R50MXFigure3-14CoreOutletMassFluxfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingPumpVoidFractionANP-2903(NP)Revision000Page3-361.00.80.6c.Q"0co"-l.L"'0*0>0.40.20.0o200Time(5)BrokenLoop1IntactLoop2IntactLoop31ntactLoop440060010:22950160ec200915:31:36R50MXFigure3-15VoidFractionatRCSPumpsfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-37ECCSFlows3000Loop1(broken)Loop2Loop3Loop4600400---r--200I'1\I,:1I'IIIiIiI,:1:1I'III,III,:1\'II(-...11,--------------------Ir_""'-oo\II1,1IIII2000f-1III(j)--E..0Q)..-co0:::50l.L1000Time(5)10:22950160ec200915:31:36R50MXFigure3-16ECCSFlows(IncludesSIT,LPSIandHPSI)fortheLimitingCaseAREVANPInc_

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingUpperPlenumPressure30002000ro.iii8Q).....::JVlVl0..1000ANP-2903(NP)Revision000Page3-38oo\200Time(5)40060010:22950160ec200915:31:36R50MXFigure3-17UpperPlenumPressurefortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingDowncomerLiquidLevel30--Sector1(broken)...........Sector2----Sector3---Sector4--Average1\20I!gIIIIQ)I>IIIQ)...JIIIII"'0*SII\.g----l10*!I;Iiiji1II.I,I:i00200400600Time(s)10:22950160ec200915:31:36R50MXFigure3-18CollapsedLiquidLevelintheDowncomerfortheLimitingCaseAREVANPInc.ANP-2903(NP)Revision000Page3-39 StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-40LowerVesselLiquidLevel141210(tv§:8(I)>(I)--.J642600400200oTime(5)10:22950160ec200915:31:36R50MXFigure3-19CollapsedLiquidLevelintheLowerPlenumfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-41CoreLiquidLevel15,----------,----,------------r------.---------,,--------,--HotAssembly----CenterCore---AverageCore---OuterCore600400200"Iii,'III,,!\Tlo10gQ)>Q)....JII::l0-il:.::iI,I5Time(s)10:22950160ec200915:31:36R50MXFigure3-20CollapsedLiquidLevelintheCorefortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-42ContainmentandLoopPressures1009080--Containment----SGOutlet(primaryside)---UpperPlenum---OowncomerInlet7060ro*w8a.>50......::l(/)(/)a.>......0...40302010600400200OL---------'--------'------------'-------'------------"L---------'oTime(5)10:22950160ec200915:31:36R50MXFigure3-21ContainmentandLoopPressuresfortheLimitingCaseAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page3-432200---------------2200I.Zr4LOOP[JZr4No-LOOPI200020001800180016001400C-----******o*cCfico.CC0*[J*C.co*DO.*------------0-----cccO[JCc.c0**C.....*****c.C*Cc.CoC[JCo*1400---tE-i1600UC1200CC***o**[Jo[J1000I0I1000o1020304050601200CaseNumberFigure3-22GOC35LOOPversusNo-LOOPCasesAREVANPInc.

Generic Support for Transition Package 4.1 Reactor Power Question:It is indicated in the RLBLOCA analyses that the assumed reactor core power includes uncertainties. The use of a reactor power assumption other than 102 percent, regardless of BE or Appendix K methodology, is permitted by Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K.I.A, "Required and Acceptable Features of The Evaluation Models, 'Sources of Heat During a LOCA. However, Appendix K.I.A also states: ... An assumed power level lower than the level specified in this paragraph [1.02 times the licensed power level], (but not less than the licensed power level) may be used provided . . . Please explain. Response:4.2 Rod Quench Question: Does the version of S-RELAP5 used to perform the computer runs assure that the void fraction is less than 95 percent and the fuel cladding temperature is less than 900 °F before it allows rod quench?Response:

4.3 Rod-to-Rod Thermal Radiation Question:Provide justification that the S-RELAP5 rod-to-rod thermal radiation model applies to the St Lucie Unit 1 core.Response:

Table 4-1 Typical Measurement Uncertainties and Local Peaking Factors Plant F Measurement Uncertainty (percent) Local Pin Peaking Factor (-)

Table 4-2 FLECHT-SEASET & 17x17 FA Geometry Parameters Design Parameter FLECHT-SEASET 17x17 Fuel AssemblyFigure 4-1 R2RRAD 5 x 5 Rod Segment Table 4-3 FLECHT-SEASET Test Parameters Test Rod 7J PCT at 6-ft (°F) PCT Time (s)htc at PCTtime (Btu/hr-ft2-°F)Steam Temperature -at 7I (6-ft) (°F) Thimble Temperature at 6-ft (°F) set to steam temp Figure 4-2 Rod Thermal Radiation in FLECHT-SEASET Bundle and in a 17x17 FA 00.511.522.533.544.514001500160017001800190020002100220023002400PCT (°F)Radiation HTC (BTU-hr/ft^2-°F)FLECHT_SEASET set-1FLECHT_SEASET Set-2Fuel Assembly 4.4 Film Boiling Heat Transfer Limit Question:In the St Lucie Unit 1 Cycle 24 calculations, is the Forslund-Rohsenow model contribution to the heat transfer coefficient limited to less than or equal to 15 percent when the void fraction is greater than or equal to 0.9? Response:4.5 Downcomer Boiling Question:If the PCT is greater than 1800°F or the containment pressure is less than 30 psia, has the St Lucie Unit 1 downcomer model been rebenchmarked by performing sensitivity studies, assuming adequate downcomer noding in the water volume, vessel wall and other heat structures?Response:

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-9pressuresandhighcladdingtemperatures.Boiling,whereveritoccurs,isaphenomenonthatcodeslikeS-RELAP5havebeendevelopedtopredict.Downcomerboilingistheresultofthereleaseofenergystoredinvesselmetalmass.WithinS-RELAP5,downcomerboilingissimulatedinthenucleateboilingregimewiththeChencorrelation.ThismodelinghasbeenvalidatedthroughthepredictionofseveralassessmentsonboilingphenomenonprovidedintheS-RELAP5CodeVerificationandValidationdocument(Reference12).(1\.0.0\/'tr--CHY{*-------..mFigure4-3ReactorVesselDowncomerBoilingDiagramHotdowncomerwallspenalizePCTbytwomechanisms:byreducingsubcoolingofcoolantenteringthecoreandthroughthereductionindowncomerhydraulicheadwhichisthedrivingforceforcorereflood.Althoughboilinginthedowncomeroccursduringblowdown,thebiggestpotentialforimpactoncladtemperaturesisduringlaterefloodfollowingtheendofaccumulatorinjection.Atthistime,thereisalargestepreductionincoolantflowfromtheECCsystems.AsAREVANPInc.

Figure 4-4 S-RELAP5 versus Closed Form Solution Distance from Inner Wall, feetMetal Temperature, F StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladding4.5.1.2PlantModelSensitivityStudyANP-2903(NP)Revision000Page4-13Asadditionalverification,atypical4-100pplantcasewasusedtoevaluatetheadequacyofthemeshspacingwithinthedowncomerwallheatstructure.Eachmeshintervalinthebasecasedowncomervesselwallwasdividedintotwoequalintervals.Thus,anewinputmodelwascreatedbyincreasingthenumberofmeshintervalsfrom9to18.Thefollowingfourfiguresshowthetotaldowncomermetalheatreleaserate,peTindependentofelevation,downcomerliquidlevel,andthecoreliquidlevel,respectively,forthebasecaseandthemodifiedcase.Theseresultsconfirmtheconclusionfromtheexactsolutionstudythatthemeshspacingusedintheplantmodelforthedowncomervesselwallisadequate.30000.00,----,---_-----,--;,"-'24000.00f---+-------j'---0'Q)C/)18000.00::::l......co'-"Q)C/)co12000.00Q)a;0:::>.01'-Q)cW6000.00coSTime(sec)Figure4-5DowncomerWallHeatRelease-WallMeshPointSensitivityAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-14400.0320,0240.0160.080,0I=::::Wall(9-meSh1AI..v."VSLWall.rJ-V-"'..,.,,",".....'......".-0.000.0600.002400.001800,00.---..u..0())I-:::::l......COI-1200.00())0..E())f-Time(sec)Figure4-6peTIndependentofElevation-WallMeshPointSensitivityAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-1530.0020.00rt-------+Ii:--IIIfll-+*------fM7nh-----,-------,--+-h...,.......,.,-10.000.000.080.0160.0240.0320.0400.0Time(sec)Figure4-7DowncomerLiquidLevel-WallMeshPointSensitivityAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-1612.00,--__-,--__--,10.00:;::;-)1---------+------+-------1---------+-------1Q)4.00m------+--j-----+-------t-------+-------I2.001tt-I:-----,F,--+------+-------l--------+--------10.000.080.0160.0240.0320.0400.0Time(sec)Figure4-8CoreLiquidLevel-WallMeshPointSensitivityAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingBasemodel@(lH9@(Hy@(HYRevised9RegionModel(IHY(Hy(HLFigure4-9AzimuthalNoding4.5.2.2AxialNodalizationANP-2903(NP)Revision000Page4-18TheRLBLOCAmethodologydividesthedowncomerintosixnodesaxially.Inboth3-loopand4-loopmodels,thedowncomersegmentattheactivecoreelevationisrepresentedbytwoequallengthnodes.Formostoperatingplants,theactivecorelengthis12feetandthedowncomersegmentsattheactivecoreelevationareeach6-feethigh.(Fora14footcore,thesenodeswouldbe7-feethigh.)Themodelforthesensitivitystudypresentedherecomprisesa3-loopplantwithanicecondensercontainmentanda12footcore.Forthestudy,thetwonodesspanningtheactivecoreheightaredividedinhalf,revisingthemodeltoincludeeightaxialAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-19nodes.Further,therefinednodingislocatedwithinthepotentialboilingregionofthedowncomerwhere,ifthereisanaxialresolutioninfluence,thesensitivitytothatimpactwouldbegreatest.Theresultsshowthattheaxialnodingusedinthebasemethodologyissufficientforplantsexperiencingtheverylowsystempressurescharacteristicoficecondensercontainments.Figure4-10providesthecontainmentbackpressureforthebasemodeling.Figure4-11throughFigure4-14showthetotaldowncomermetalheatreleaserate,peTindependentofelevation,downcomerliquidlevel,andthecoreliquidlevel,respectively,forthebasecaseandthemodifiedcase.Theresultsdemonstratethattheaxialresolutionprovidedinthebasecase,6axialdowncomernodedivisionswith2divisionsspanningthecoreactiveregion,aresufficienttoaccuratelyresolvevoiddistributionswithinthedowncomer.Thus,thismodelingissufficientforthepredictionofdowncomerdrivingheadandtheresolutionofdowncomerboilingeffects.400.0320.0240.0160.080.01--Base6x6'"o.0.0040.008.0016.0024.0032.00Time(sec)Figure4-10LowerCompartmentPressureversusTimeAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-2018000.00r-i-------j---f-+'12000.0030000.00,-----,----,----r.r;Q)(/)coQ)Q)0:::01.....Q)cWcoS0-24000.00I---j-----+-----.:6000.00H-----+------t------j-------+---------1Time(sec)Figure4-11DowncomerWallHeatRelease-AxialNodingSensitivityStudyAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-212400.001800.00----l.LQ).....::Jco00.....Q)Cl..EQ)I-600.00.':.IL"-",'.\.""'.,......,..h**,.....,******...*."...Time(sec)Figure4-12peTIndependentofElevation-AxialNodingSensitivityStudyAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-2230.0010.00Time(sec)Figure4-13DowncomerLiquidLevel-AxialNodingSensitivityStudyAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-2312.0010.00.........8.00Q3Q5>Q).....J6.00:Q::::l0-:.:::i4.00::'2.001tt-t------;f.It--+------+------+------+----------10.000'::-.0-----=::.eTime(sec)Figure4-14CoreLiquidLevel-AxialNodingSensitivityStudy4.5.3DowncomerBoilingConclusionsTofurtherjustifytheabilityoftheRLBLOCAmethodologytopredictthepotentialforandimpactofdowncomerboiling,studieswereperformedonthedowncomerwallheatreleasemodelingwithinthemethodologyandontheabilityofS-RELAP5topredictthemigrationofsteamthroughthedowncomer.Bothazimuthalandaxialnodingsensitivitystudieswereperformed.Theaxialnodingstudywasbasedonanicecondenserplantthatisnearatmosphericpressureduringreflood.ThesestudiesdemonstratethatS-RELAP5deliversenergytothedowncomerliquidvolumesatanappropriaterateandthatthedowncomernodingdetailissufficienttotrackthedistributionofanysteamformed.Thus,therequiredmethodologyforthepredictionofdowncomerboilingatsystempressuresapproximatingthoseachievedinplantswithpressuresaslowasicecondensercontainmentshasbeendemonstrated.AREVANPInc.

4.6 Break

Size Question:Were all break sizes assumed greater than or equal to 1.0 ft2?Response:

Table 4-4 Minimum Break Area for Large Break LOCA Spectrum Plant Description System Pressure (psia) Cold Leg Enthalpy (Btu/lbm)Subcooled Gbreak (lbm/ft2-s) Saturated Gbreak (HEM) (lbm/ft2-s) RCP flow (lbm/s) Spectrum Minimum Break Area (ft2) Spectrum Minimum Break Area (DEGB)

Table 4-5 Minimum PCT Temperature Difference - True Large and Intermediate Breaks Plant Description Generic Plant Label (Table 4-4) Maximum PCT (°F) Intermediate Size Break Maximum PCT (°F) Large Size Break Delta PCT (°F) Average Delta PCT (°F)

Figure 4-15 Plant A - Westinghouse 3-Loop Design6008001000120014001600180020000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area Figure 4-16 Plant B - Westinghouse 3-Loop Design6008001000120014001600180020000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area 6008001000120014001600180020000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area Figure 4-17 Plant C - Westinghouse 3-Loop Design 6008001000120014001600180020000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area Figure 4-18 Plant D - Combustion Engineering 2x4 Design 6008001000120014001600180020000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area Figure 4-19 Plant E - Combustion Engineering 2x4 Design 600.0000800.00001000.00001200.00001400.00001600.00001800.00002000.00002200.00000.00000.10000.20000.30000.40000.50000.60000.70000.80000.90001.0000Break Area Normalized to Double Ended GuillotinePCT (°F)Upper End of SBLOCA Break Size SpectrumLarge Break Spectrum Minimum Break Area Figure 4-20 Plant F - Westinghouse 3-loop Design

4.7 Detail

information for Containment Model StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingPCTvsContainmentVolume1800-D**1600D*D*-*D...D***wibJ]*D**dI*D***1400cJD-**DEDeYDI-1200-00...D1000-D800-ANP-2903(NP)Revision000Page4-36600I*SplitBreak1-DGuillotineBreakl2.5500e+06ContainmentVolume(ft3)2.6500e+06AREVANPInc.Figure4-21PCTvs.ContainmentVolume StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingANP-2903(NP)Revision000Page4-37PCTvsContainmentTemperature200018000-0-1600cJIoIII-IioII.-..'d-o_00*0-.-1400**0E00Df-12000D...D1000D800600*SplitBreakoGuillotineBreak130120ContainmentTemperatureCF)400L--_____'___'___________'110Figure4-22PCTV5.InitialContainmentTemperatureAREVANPInc.

StLucieNuclearPlantUnit1EPUCycleRealisticLargeBreakLOCASummaryReportwithZr-4FuelCladdingContainmentPressures1-10nn.(B)ANP-2903(NP)Revision000Page4-38Figure4-23ContainmentPressureasfunctionoftimeforlimitingcaseAREVANPInc.

4.8 Cross-References to North Anna Question:In order to conduct its review of the St Lucie Unit 1 application of AREVA's realistic LBLOCA methods in an efficient manner, the NRC staff would like to make reference to the responses to NRC staff requests for additional information that were developed for the application of the AREVA methods to the North Anna Power Station, Units 1 and 2, and found acceptable during that review. The NRC Staff safety evaluation was issued on April 1, 2004 (Agency-wide Documentation and Management System (ADAMS) accession number ML040960040). The staff would like to make use of the information that was provided by the North Anna licensee that is not applicable only to North Anna or only to subatmospheric containments. This information is contained in letters to the NRC from the North Anna licensee dated September 26, 2003 (ADAMS accession number ML032790396) and November 10, 2003 (ADAMS accession number ML033240451). The specific responses that the staff would like to reference are: September 26, 2003 letter: NRC Question 1 NRC Question 2 NRC Question 4 NRC Question 6 November 10, 2003 letter: NRC Question 1 Please verify that the information in these letters is applicable to the AREVA model applied to St Lucie Unit 1 except for that information related specifically to North Anna and to sub-atmospheric containments. Response:

4.9 GDC 35 - LOOP and No-LOOP Case Sets Question:10CFR50, Appendix A, GDC [General Design Criterion] 35 [Emergency core cooling] states that, Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite electric power is not available) and for offsite electric power operation (assuming onsite power is not available) the system function can be accomplished, assuming a single failure.The Staff interpretation is that two cases (loss of offsite power with onsite power available, and loss of onsite power with offsite power available) must be run independently to satisfy GDC 35. Each of these cases is separate from the other in that each case is represented by a different statistical response spectrum. To accomplish the task of identifying the worst case would require more runs. However, for LBLOCA analyses (only), the high likelihood of loss of onsite power being the most limiting is so small that only loss of offsite power cases need be run. (This is unless a particular plant design, e.g., CE [Combustion Engineering] plant design, is also vulnerable to a loss of onsite power, in which situation the NRC may require that both cases be analyzed separately. This would require more case runs to satisfy the statistical requirement than for just loss of offsite power.) What is your basis for assuming a 50% probability of loss of offsite power? Your statistical runs need to assume that offsite power is lost (in an independent set of runs). If, as stated above, it has been determined that Palisades, being of CE design, is also vulnerable to a loss of onsite power, this also should be addressed (with an independent set of runs). Response:

4.10 Input Variables Statement Question:Provide a statement confirming that Florida Power & Light (FP&L)and its LBLOCA analyses vendor have ongoing processes that assure that the input variables and ranges of parameters for the LBLOCA analyses conservatively bound the values and ranges of those parameters for the operated St Lucie Nuclear Plant Unit 1(SLA). This statement addresses certain programmatic requirements of 10 CFR 50.46, Section (c).Response:FP&L Conclusions References Realistic Large Break LOCA MethodologyQuantifying Reactor Safety Margins.Mixing of Emergency Core Cooling Water with Steam: 1/3 - Scale Test and SummaryS-RELAP5: Code Verification and Validation