ML11256A023

NUREG-1953 Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models - Surry and Peach Bottom.

ML11256A023
Person / Time
Site:	Peach Bottom, Surry
Issue date:	09/30/2011
From:	Buell R F, Hossein Esmaili, Helton D M, Koonce T, Marksberry D G, Schroeder J, Sherry R Office of Nuclear Regulatory Research, Idaho National Lab, Office of New Reactors
To:
Beltz, G
References
NUREG-1953
Download: ML11256A023 (451)

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Confirmatory Thermal-Hydraulic Analysis to Support Specific

Success Criteria in the Standardized Plant Analysis Risk Models-Surry and Peach Bottom Office of Nuclear Regulatory ResearchNUREG-1953 AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material As of November 1999, you may electronically access NUREG-series publications and other NRC records at

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UNITEDSTATESNUCLEARREGULATORYCOMMISSIONWASHINGTON,D.C.20555-0001 ErrataNUREG-1953,"ConfirmatoryThermal-HydraulictoSupportSpecificSuccessCriteriaintheStandardizedPlantAnalysisRiskModels-SurryandPeachBottom,"publishedSeptember2011Page10incorrectlyliststhemodeledpowerlevelforPeachBottomas3,458MWt.Themodeledpowerwasinfactthecorrectpowerlevel(asof2011)of3,514MWt.Thefollowingequationswereinadvertentlyomittedfrompage13:CumulativeProbabilityofFailure

=1-(1-PD)n, orCumulativeProbabilityofFailure

=1-(1-Pbnitial)(l_PDSUbSeQUent)n-lFootnote20(pg.39)shouldbereadwiththeknowledgethattheMELCORmodelwouldnotaccountfortheeffectofambientpumproomtemperatureonthepumpbearingtemperature.ThePeachBottomanalysesthatcreditreactorcoreisolationcooling(RCIC)inautomaticmodeuseasetpointforhigh-vessel-levelshutoffwhichis10inches(0.25m)belowtheactualplantvalue.Thisisnotexpectedtohaveasignificanteffectonthepresentedresults,inthatitwillaffectthespecificsofthefrequencyanddurationofRCICduties,butnotwhetherornotRCICisabletomaintaincorecooling.Someplotsinthereporthaveaccidentparametersignatures(Le.,timehistories)thatarepartiallyorcompletelyobscuredbyoverlyingsignatures(e.g.,thefigureatthetopofpageA-7).Inviewingthesefigures,thereadermustusecuestorecognizethisoverlap.Forexample,inthefigureatthebottomofpageA-7,itcanbediscernedthatthe3steamgenerators'pressuresoverlieoneanotherbecause:(i)thereisnodepartureattimezero,wherethethreepressureswouldberoughlythesameand(ii)thecurvesdodivergeattheendofthesimulation.Similarly,someplotscontainsignatureswherethevaluesarezerofortheentiresimulation.Theseareincludedbecausethezerovalueinandofitselfprovidesinformation(e.g.,confirmsthatasystemwasappropriatelydisabledforasimulationwhereitisassumedtobeunavailable,showsthatalow-pressurepumpwasdead-headedfortheentiresimulation).Suchsituations cangenerallybediscernedbydiscoloredx-axes.Finally,thefollowingabbreviationwasinadvertentlyomitted:*SPR-Containmentspray

NUREG-1953Confirmatory Thermal-HydraulicAnalysis to Support Specific Success Criteria in theStandardized Plant AnalysisRisk Models-Surry and Peach Bottom Prepared by:H. Esmaili, 1D. Helton, 1D. Marksberry, 1R. Sherry (NRC retired), 1P. Appignani, 1D. Dube, 2M. Tobin 1R. Buell, 3T. Koonce, 3J. Schroeder 3Manuscript Completed: May 2011Date Published: September 2011

_________________

1Officeof Nuclear Regulatory ResearchU.S. NuclearRegulatory CommissionWashington, DC 20555 2Officeof New ReactorsU.S. Nuclear Regulatory CommissionWashington, DC 20555 3Idaho National LaboratoryP.O. Box 1625 Idaho Falls, ID 83415

iiiABSTRACTIn a limited number of cases, thermal-hydraulic success criteria from the suite of standardized plant analysis risk (SPAR) models have apparent inconsistencies when compared to counterpart licensee probabilistic risk assessments (PRAs), other relevant SPAR models (i.e., models for similar plants), or relevant engineering studies. These inconsistencies are a natural outcome of the SPAR development process, and often reflect the apparent inconsistencies seen across licensee PRAs for similar plants. Even so, the U.S. Nuclear Regulatory Commission (NRC) staffwants to strengthen the technical basis for the SPAR models by performing targeted additional engineering analysis. The identified success criteria are for both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). This report describes MELCOR analyses performed to augment the technical basis for supporting or modifying these success criteria.The success criteria contained herein are intended to be confirmatory in nature, and while suitable for their intended use in supporting the SPAR models they are not intended to be used by licensees for risk-informed licensing submittals.This report first provides a basis for using a core damage surrogate of 2,200degreesFahrenheit (1,204degreesCelsius)peak cladding temperature.Following thisdiscussion are descriptions ofthe major plant characteristics for the two plants used for this analysis (SurryPower Stationand Peach BottomAtomic Power Station)andthe MELCOR models used to represent these plants. Finally, the report presents the results of many MELCOR calculations andcomparesthese results to the corresponding sequences and success criteria in theSPARmodelsfor Surry and Peach Bottom.The results provide additional timing information for many sequences, confirm many of the existing SPAR model modeling assumptions, and support a few specific changes. Specific changes that have been made to the SPAR models as a result of these analyses are:For six SPAR models corresponding to three-loop"high-head" Westinghouse PWRs:-Reduction of the adequate venting capability for feed and bleed from two power-operated relief valves (PORVs) to one PORV.-Adjustment of the sufficient injection flow during the early stages of a large-break loss-of-coolant accident from two accumulators to one accumulator or one high-head safety injection pump. For SPAR models corresponding to BWR Mark Is and Mark IIs:-Credit for two control rod drive (CRD) pumps providing adequate core cooling flow following the initial successful operation of the high-pressure coolant injection system or reactor core isolation cooling system.-Credit for one CRD pump providing adequatecore cooling for injection late in the accident sequence (if not already included).Some additional changes supported by the MELCOR analysis are not implemented becausethey are limited by other SPAR modeling assumptions (e.g., timing of core damage relative to battery depletion for station blackout sequences).

v FOREWORDThe U.S.Nuclear Regulatory Commission's standardized plant analysis risk(SPAR)modelsare used to support a number of risk-informed initiatives. The fidelity and realism of these models is ensured through a number of processes,including cross-comparison with industry models, review and use by a wide range of technical experts, and confirmatory analysis. The following report,prepared by staff inthe Office of Nuclear Regulatory Research in consultation with staff from the Office of Nuclear Reactor Regulation, experts from Idaho National Laboratory,and the agency's senior reactor analysts,represents a majorconfirmatory analysisactivity.One of the key strengths and challenges of probabilistic risk assessment (PRA) models is the integration of modeling capability from different disciplines, including human performance, thermal-hydraulics, severe accident progression, nuclear analysis, fuelsbehavior, structural analysis, and materials analysis. Thisreportinvestigatesthe thermal-hydraulic aspects of the SPAR models, with the goal of further strengthening the technical basis for decisionmaking that relies on the SPAR models.This analysisemploysthe MELCOR computer code,using plant models developed as part of the State-of-the-Art Reactor Consequence Analyses project. This report uses these models for a number of scenarioswithdifferent assumptions. In many cases,the operator responseis not modeled in order to establish minimal equipment needs or bounding operator action timings. The report clearly articulates all assumptions and limitations.The analyses summarized in this report providethe basis for confirming or changing successcriteria in the SPAR models for the Surry Power Station and Peach BottomAtomic Power Station.Further evaluationof these results was also performed toextend the resultsto similarplants. In addition, future work is plannedto perform similar analysis for other design classes.In addition, work is plannedto scope other aspects of this topical area, including the degree of variationtypical in common PRA sequencesand the quantification of conservatisms associated with core damage surrogates. The confirmation of success criteria and other aspects of PRA modeling using the agency's state-of-the-art tools (e.g.,the MELCOR computer code) is expected to receive continued focus as the agency movesforwardin this area.

vii CONTENTS Section PageABSTRACT................................................................................................................................iii FOREWORD.......................................................................................................................

.......v CONTENTS..............................................................................................................................vii LIST OF FIGURES....................................................................................................................ix LIST OF TABLES.................................................................................................................

.......x ABBREVIATIONS AND ACRONYMS.......................................................................................xii

1. INTRODUCTION AND BA CKGROUND..............................................................................1

2. DEFINITION OF CORE DAMAGE......................................................................................3

3. RELATIONSHIP TO THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS/AMERICAN NUCLEAR SOCIETY PROBABILISTIC RISK ASSESSMENT STANDARD.......................................................................................................................

.7 4. MAJOR PLANT CHARACTERISTICS................................................................................9

4.1 Surry

Power Station.....................................................................................................9

4.2 Peach

Bottom Atomic Power Station..........................................................................10

5. MELCOR MODEL..............................................................................................................11

5.1 Plant

Representation..................................................................................................11

5.2 MELCOR

Validation....................................................................................................15

6. MELCOR RESULTS..........................................................................................................17 6.1 Small-Break Loss-of-Coolant Accident Dependency on Sump Recirculation (Surry)...17 6.2 Feed-and-Bleed Power-Operated Relief Valve Success Criteria (Surry).....................20

6.3 Steam

Generator Tube Rupture Event Tree Timing (Surry)........................................22 6.4 Pressurized-Water Reactor Station Blackout (Surry)...................................................24 6.5 Pressurized-Water Reactor Medium-and Large-Break Loss-of-Coolant Accident Initial Response (Surry).......................................................................................................30

6.6 Inadvertent

Open Relief Valve Success Criteria (Peach Bottom)................................36 6.7 Boiling-Water Reactor Station Blackout (Peach Bottom).............................................38

7. APPLICATION OF MELCOR RESULTS TO SURRY AND PEACH BOTTOM SPARMODELS...........................................................................................................................43 8. CONCLUSION...................................................................................................................49

9. REFERENCES..................................................................................................................51

viiiCONTENTS(Continued)

Section PageAPPENDIX ASURRY MELCOR ANALYSESA.1Summary of Surry Model Changes----------------.-----.A-1A.2Small-Break Loss-of-Coolant Accident Dependency on Sump Recirculation---..A-5A.3Feed and Bleed PORV Success Criteria------------------..A-59A.4Steam Generator Tube Rupture Event Tree Timing-------------..A-75A.5Pressurized-Water Reactor Station Blackout----------------..A-91A.6Pressurized-Water Reactor Medium-and Large-Break LOCA Initial Response.....A-135A.7References------------------------------..A-254APPENDIX BPEACH BOTTOM MELCOR ANALYSESB.1 Inadvertent Open Relief Valve Success Criteria---------------...B-1B.2Boiling-Water ReactorStation Blackout-------------------B-37APPENDIX C EVENTTREE MODELSFORSURRYAND PEACHBOTTOMC.1Surry Event Trees----------------------------..C-1C.2Peach Bottom Event Trees------------------------.C-11APPENDIX DRESPONSE TO PUBLICCOMMENTSD.1Introduction-----------------------------.-....D-1D.2Comments from Exelon Nuclear Dated December 15, 2010--------.-....D-3D.3Comments from the Nuclear Energy Institute Dated February 23, 2011---.-.-D-7D.4References-----------------------------...-.D-9 ixLIST OF FIGURESFigure PageFigure1 Summary of Core Damage Surrogate Calculations.....................................................5 Figure2 Plan View of the Surry MELCOR RCS Model.............................................................12 Figure 3 Schematic of the Peach Bottom RCS Nodalization....................................................13 Figure4 PCT Signatures for all Surry Station Blackout Cases.................................................28 Figure5 Surry Injection Recovery Sensitivity Cases................................................................30

xLIST OF TABLESTable PageMain ReportTable1 Comparison of this Project to the ASME/ANS PRA Standard.......................................7 Table 2 Major Plant Characteristics for Surry...........................................................................10 Table 3 Major Plant Characteristics for Peach Bottom.............................................................10 Table 4 Comparison of Failure per Demand Probabilities for Surry Stuck-Open Valves...........

14 Table 5 Comparison of Number-of-Lifts Values for Surry Stuck-Open Valves..........................14 Table6 Surry SBLOCA Sump Recirculation Results................................................................19 Table7 Surry SBLOCA Sump Recirculation Key Timings (Cases1-4)....................................19 Table8 Surry SBLOCA Sump Recirculation Key Timings (Cases5-8)....................................20 Table9 Surry Feed-and-Bleed PORV Success Criteria Results...............................................22 Table10 Surry Feed-and-Bleed PORV Success Criteria Key Timings.....................................22 Table11 Surry SGTR Results..................................................................................................23 Table12 Surry SGTR Key Timings..........................................................................................24 Table13 Reactor Coolant Pump Seal Leakage Details............................................................25 Table14 Surry Station Blackout Results..................................................................................27 Table15 Surry Station Blackout Key Timings (Cases 1-2)......................................................28 Table16 Surry Station Blackout Key Timings (Cases 3-6)......................................................29 Table17 Surry Station Blackout Key Timings (Cases 7-10)....................................................29 Table18 PCT Ranges for Accumulator Success Cases...........................................................31 Table19 Surry MBLOCA and LBLOCA Results.......................................................................32 Table20 Surry MBLOCA and LBLOCA Key Timings (2-in. Breaks).........................................33 Table21 Surry MBLOCA and LBLOCA Key Timings (4-in. Breaks Group1)...........................33 Table 22 Surry MBLOCA and LBLOCA Key Timings (4-in. Breaks Gr oup 2)...........................34 Table 23 Surry MBLOCA and LBLOCA Key Timings (6-in. Breaks Gr oup 1)...........................34 Table 24 Surry MBLOCA and LBLOCA Key Timings (6-in. Breaks Gr oup 2)...........................35 Table25 Surry MBLOCA and LBLOCA Key Timings (8-in. Breaks).........................................35 Table26 Surry MBLOCA and LBLOCA Key Timings (-in. Breaks).....................................35 Table27 Peach Bottom Inadvertent Open SRV Results..........................................................38 Table28 Peach Bottom Inadvertent Open SRV Key Timings (Cases 1-5)...............................38 Table29 Peach Bottom Station Blackout Results....................................................................40 Table30 Peach Bottom Station Blackout Key Timings (Cases1, 1a, and 2)............................40 Table31 Peach Bottom Station Blackout Key Timings (Cases3-6).........................................41 Table32 Peach Bottom Station Blackout Key Timings (Cases7-10).......................................41 Table33 Mapping of MELCOR Analyses to the Surry1 & 2 SPAR(v3.52) Model...................44 Table34 Mapping of MELCOR Analyses to the Peach Bottom2 SPAR(v3.50) Model............45 Table35 Comparison of Surry Station Blackout Results to the SPAR Model...........................45 Table36 Potential Success Criteria Updates Based on Surry Results.....................................46 Table 37 Potential Success Criteria Updates Based on Peach Bottom Results........................48 Appendix ATable 1 Conditionsfor Reactor Trip-------------------------A-1Table 2 High-Head Safety Injection Flow per Pump------------------A-2 Table 3 Low-HeadSafety Injection Flow per Pump------------------.A-3Table 4 Opening and Closing Pressures for PORV and SRV--------------A-4 xiLIST OF TABLES (continued)Table PageAppendix A(continued)Table 5 Surry SBLOCA Sump Recirculation Results-----------------..A-5Table 6 Surry SBLOCA Sump Recirculation Key Timings (Cases 1-4)---------...A-5Table 7 SurrySBLOCA Sump Recirculation Key Timings (Cases 5-8)---------...A-6Table 8 Surry Feed-and-Bleed PORV Success Criteria Results------------.A-59Table 9 Surry Feed-and-Bleed PORV Success Criteria Key Timings----------A-59Table 10 Surry SGTR Results--------------------------...A-75Table 11 Surry SGTR Key Timings------------------------..A-75Table 12 Surry Station Blackout Results----------------------.A-91Table 13 Surry Station Blackout Key Timings (Cases 1-2)--------------..A-92Table 14 Surry Station Blackout Key Timings (Cases 3-6)--------------..A-92Table 15 Surry Station Blackout Key Timings (Cases 7-10)--------------A-92Table 16 Surry MBLOCA and LBLOCA Results------------------..A-135Table 17 Surry MBLOCA and LBLOCA Key Timings (2 Inch Breaks)---------.A-136Table 18 Surry MBLOCA and LBLOCA Key Timings (4 Inch Breaks Group 1)-----..A-136Table 19 Surry MBLOCA and LBLOCA Key Timings (4-Inch Breaks Group 2)-----..A-136Table 20 Surry MBLOCA and LBLOCA Key Timings (6-Inch Breaks Group 1)-----..A-137Table 21 Surry MBLOCA and LBLOCA Key Timings (6-Inch Breaks Group 2)-----..A-137Table 22 Surry MBLOCA and LBLOCA Key Timings (8 Inch Breaks)---------.A-137Table 23 Surry MBLOCA and LBLOCA Key Timings (A-137AppendixBTable 1 Peach Bottom Inadvertent Open SRV Results----------------...B-1Table 2 Peach Bottom Inadvertent Open SRV Key Timings (Cases 1-5)--------...B-1Table 3 Peach Bottom Station Blackout Results-------------------B-37Table 4 PeachBottom Station Blackout Key Timings (Cases 1, 1a, and 2)-------..B-37Table 5 Peach Bottom Station Blackout Key Timings (Cases 3-6)-----------B-38Table 6 Peach Bottom Station Blackout Key Timings (Cases 7-10)----------..B-38Appendix D:Table 1 SBLOCA Case 2 with RCP Trip at 30 °F (16.7 °C) Subcooling---------..D-4Table 2 SBLOCA Case 8 with RCP Trip at 30 °F (16.7 °C) Subcooling---------..D-5Table 3 LOMFW Case 2 with RCP Trip at 10 Minutes-----------------D-5 xiiABBREVIATIONS AND ACRONYMSacalternating currentACCaccumulatorADAMSAgencywide Documents Access and Management SystemADSautomatic depressurization systemAFWauxiliary feedwaterANSAmerican Nuclear SocietyASMEAmerican Society of Mechanical EngineersBAFbottom of active fuelBRKbreakBTU/hrBritish thermal units per hour BWRboiling-water reactorCCelsiusCDFcore damage frequencyCETcore exit thermocouple CFRCode of Federal RegulationsCL-Acold leg of loop Acmcentimeter CORMELCOR core package CRDcontrol rod drive injectionCSTcondensate storage tankCVHcontrol volume hydrodynamics (MELCOR package)dcdirect currentDCdowncomerDtdelta temperatureDWdrywellECAemergency contingency actionECCSemergency core cooling systemeCSTemergency condensate storage tankEOPemergency operating procedureEPRIElectric Power Research InstitueFFahrenheitF/hrFahrenheit per hour ftfeet ft 3cubic feetFWfeedwater galgallon(s)gpmgallon(s)per minuteHCTLheat capacity temperature limitHHSIhigh-head safety injection HL-xhot leg of loop xHPCIhigh-pressure coolantinjectionhror hhour(s)in.inch(es)inj or InjinjectionIORVinadvertently openrelief valve JPjet pumpKKelvinkg/skilogram(s) per secondlb/spound(s) per secondLBLOCAorlarge-break loss-of-coolantLLOCAaccidentLHSIlow-head safety injectionLOCAloss-of-coolant accidentLOFTloss-of-fluid testLOMFWloss of main feedwater LOOPloss of offsite powerLPCIlow-pressure coolantinjectionLPCSlow-pressure core spraymmeter(s)m 3cubic meter(s) m 3/mincubic meter(s)per minute m 3/scubic meter(s) per secondMAAPModular Accident Analysis ProgramMBLOCAmedium-break loss-of-coolantor MLOCAaccident MCPmain coolant pump MD-AFWmotor-drivenauxiliary feedwaterMELCORNotan acronym MFWmain feedwater minminute(s) mmmillimeter(s)

MPamegapascal(s)MSIVmain steam isolation valveMWmegawatt(s)MWtmegawatt(s)thermalNPSHnet positive suction head NRCU.S. Nuclear Regulatory CommissionPCTpeak claddingtemperaturePORVpower-(or pilot-)operated relief valvePRAprobabilistic risk assessmentPRTpressurizer relief tank PRZperssurizerpsipound(s)per square inchpsiapound(s)per square inch absolutepsidpound(s)per square inch differential xiiipsigpound(s)per square inch gagePWRpressurized-water reactorRCICreactor core isolation coolingRCPreactor coolant pumpRCSreactor coolant systemrec or RecirrecirculationRHRresidual heat removal RPVreactor pressure vesselRWSTrefuelingwater storage tankSBLOCAsmall-break loss-of-coolantor SLOCAaccidentSCsuccess criteriascfmstandard cubic foot/feet per minutesec or ssecond(s)SGsteam generator SG-xsteam generatorin loop x SGTRsteam generator tube rupture SIsafety injectionSOARCAState-of-the-Art Reactor Consequence AnalysesSPsuppression poolSPCsuppression pool coolingSPARstandardized plant analysis riskSRVsafetyrelief valveSVsafety valveTAFtop of active fuel Tavgloop average temperatureTCthermocoupleTCLcladding termperatureTCVturbine control valve TD-AFWturbine-drivenauxiliary feedwaterTFUfuel temperatureTLIQliquid temperatureTRACETRAC/RELAP5Advanced Computational EngineTSATsaturation temperatureTVAPvapor temperatureWOGWestinghouse Owners GroupWWwetwell

11.INTRODUCTION ANDBACKGROUNDThe success criteria in the U.S.Nuclear Regulatory Commission's (NRC's)standardized plant analysis risk (SPAR) models are largely based on the success criteria used in the associated licenseeprobabilistic risk assessment(PRA)model.

1Licensees have used a variety of methods to determine success criteria, including conservative design-basis analyses and more realistic best-estimate methods. Consequently, in some situations plants that should behave similarly from an accident sequence standpoint have different success criteria for specific scenarios. This issue has been recognized for some time, but until recently the infrastructure was not in place atthe NRC to support refinement of these success criteria.To facilitate improvements in this area, the NRC staff ran MELCOR calculations for specific sequences to provide the basis for confirming or changingthe corresponding SPAR models. This analysis used theSurry Power Station (Surry) andthe Peach Bottom Atomic Power Station(Peach Bottom).Thestaff chose these plants because ofthe availability of mature and well-exercised MELCOR input models arising from the State-of-the-Art Reactor Consequence Analyses (SOARCA) project.The sequences analyzed are not necessarily the most probable sequences because ofthe assumed unavailability of systems or the assumed lack of operator action. This situation is an appropriate effect of the nature of this work (i.e., the informing of particular pieces of the PRA model). In all cases, this report gives these assumptions in the results description.This report summarizes the analyses that havebeen performed, including the following topics:the basis for the core damage definition employedmajor plant characteristics for Surry and Peach Bottomadescription of the two MELCOR models usedresults of various MELCOR calculationsapplication of the MELCOR results to the Surry and Peach Bottom SPAR models, as well asto the SPAR models for other similar plantsThe success criteria contained herein are intended to be confirmatory in nature, and while suitable for their intended use in supporting the SPAR models they are not intended to be used by licensees for risk-informed licensing submittals.

1In some cases,success criteria are based on other sources, such as NRC studies (e.g.,NUREG/CR-5072, "Decay Heat Removal Using Feed and Bleed for U.S. Pressurized Water Reactors," issued June1988 (NRC,1988)).

32.DEFINITION OF CORE DAMAGETo perform supporting analysis of success criteria, it is necessary to define what is meant by core damage (i.e.,sequence success versus failure)because no universal quantitative definition of core damage exists.The American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standardRA-Sa-2009, "Standard for Level1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," issued March2009(ASME/ANS,2009)defines core damage as "uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage are anticipated and involving enough of the core,if released, to result in offsite public health effects." The standard later requires the analysis to specify the plant parametersused to determine core damage in Section2-2.3, "Supporting RequirementSC-A2"(ASME/ANS,2009).The core damage surrogateprovides the linkage between the qualitative definition above and the quantitative, measurable computer code outputs. The surrogate is necessary since a full Level3 PRA is not being performed.For this analysis,the staff ran a number of MELCOR calculations to identify a realistically conservative core damage surrogate. This report does not thoroughly describe the MELCOR modelsused for this part of the projectfor the following reasons:All results are relative, meaning that a change in the model would generally not be expected to affect the delta-timebetween the surrogate core damage definition and the onset of rapid cladding oxidation(which is in fact another surrogate, as described further below).The model is based onthe general-purpose models used in the SOARCA project, whichwill be documented thoroughly as part of that project.The analysis used MELCOR version1.8.6(NRC,2005)to assess several possible surrogate definitions for a variety of pressurized-water reactor (PWR)andboiling-water reactor (BWR)accident sequences. For the PWR(SurryPower Station), the following sequences were analyzed:station blackout with a 182gallons per minute (gpm)(0.689cubic meters per minute (m 3/min)) per reactor coolant pump (RCP) seal leak rate 2station blackout with a 500gpm (1.89m 3/min) per RCPseal leak ratehot-leg loss-of-coolant accident (LOCA)for 2-inch (in.) (5.1-centimeter (cm)), 4-in.(10.2-cm), and 10-in.(25.4-cm)equivalent diameter break sizesFor the BWR (Peach BottomAtomic Power Station), the following sequences were analyzed:station blackout

2Note that the seal leakage assumptions used in this analysisdifferfromthose used in the SOARCA project(see additional discussion in Section6.4). Also note that the leakage rate provided here is the leakage rate at full system pressure. As the system depressurizes, the leak rate will decrease.

4recirculation line LOCA for 2-in.(5.1-cm), 6-in.(15.2-cm), and 10-in.(25.4-cm)equivalent diameter breaksizesBecause no universal definition of core damage exists, the definition used here for comparison with the surrogates will be the temperature at which the transition occurs in the Urbanic-Heidrick zirconium/water reaction correlation (i.e.,a peak cladding temperature(PCT)of approximately 1,580degreesCelsius (C)(2,876 degrees Fahrenheit(F))to 1,600degreesC(2,912degreesF)).This is the point at which the reaction becomes more energetic,and significant oxidation of the cladding is more likely.A number ofpotential surrogates that have traditionally been used in PRAs, severalof which are called out in the PRA standard(Section2-2.3)(ASME/ANS,2009)were considered. These includedvarious parameters associated with collapsed reactor vessel water level, peak core-exit thermocouple temperature, and PCT.Figure1shows the results of the MELCOR calculations to investigate these surrogates. The ordinateaxis is the time that the proposed surrogate (e.g.,1,204degreesC(2,200degreesF)) is reached, relative to the time that the zirconium/water transition temperature range (1,580degreesCto 1,600degreesC) is reached. In all cases butone (the surrogate representing acore exit thermocouple temperature greater than 1,200 degrees F plus a 30-minute offset),the proposed surrogate is reached beforethe oxidation transition temperature (see "Time Rapid Core Damage"in Figure1). A PCTof 1,204degreesC(2,200degreesF)achievesallof the following characteristics:It always precedes oxidation transition.It is not overly conservative.It is equally applicable for both PWRs and BWRs.The timing between 1,204degreesC(2,200degreesF) and oxidation transition is relatively similar among the different sequences analyzed.It is consistent with the criteria contained in Title10 of the Code of Federal Regulations(10CFR)50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors"(10CFR,2007).With regard to the latter bullet, the conservatism (i.e.,safety margin) in 10CFR50.46 is due to uncertainty inlarge-break loss-of-coolant accident(LBLOCA)thermal-hydraulic analysis. For PRA usage, the margin has, in part,a different reason:the desire to have a specific criterion that can be used for all sequencescombined with overall analysis uncertainty. For the reasonsstated above,a PCT of 1,204degreesC(2,200degreesF)is the surrogate used to define core damage for the MELCOR analysesin this report.

5Figure1Summary of Core Damage Surrogate Calculations(1,200°F=649°C; 2,200°F=1,204°C; 2,500°F=1,371°C)

73.RELATIONSHIPTOTHEAMERICAN SOCIETY OF MECHANICAL ENGINEERS/AMERICAN NUCLEAR SOCIETY PROBABILISTIC RISK ASSESSMENTSTANDARDCore damagespecification is one of several aspects of success criteria analysis covered by the ASME/ANS PRA standard(ASME/ANS,2009).Although the present projectisconfirmatory innature, it is still prudent to cross-check the effort against the PRA standard requirements(seeTable1).Capability Category IIis used for comparison, since this is the category identified in Regulatory Guide1.200, Revision2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," issued March 2009, as currentindustrygood practice(NRC,2009).Because the current report focuses primarily on the actual thermal-hydraulic andaccident progression analysisand defers the actual PRA model changes for a subsequent report, there are some cases where the comparison to the standard has limited applicability. Table1notes these instances as appropriate.Table1Comparison of this Project to the ASME/ANS PRA StandardPRA Standard Supporting Requirement for Capability CategoryIIThis ProjectSC-A1: Use provided core damage definition or justify the definition used.The standard provides a qualitative core damage definition. The definition used here is believed to be consistent with the definition, but is necessarily quantitative. The basis for the definition (in terms of quantitative accident analysis and comparison of alternatives) is provided. Sensitivity calculations of direct current (dc)power recovery during station blackout have demonstrated that there is not excessive margin in the definition used.SC-A2: Specify the quantitative surrogate used for core damage and provide basis.SC-A3: Specify success criteria for each safety function for each accident sequence.Theexisting SPAR model essentially satisfies the requirement.Any changes proposedto the success criteria should not inappropriately remove criteria for important safety functions;this is believed to be the case.SC-A4: Identify systems shared by units and how they perform during initiating events affecting both units.In the context of this project, this requirement only applies to changes in whichthe success criteria is modified to include systems that are shared by multiple units that were not previously in the success criteria.This is not believed to be the case for any of the changes proposed.SC-A5:Specify the mission times being used (and use appropriate mission times).These calculationsusean overall mission time of 24hours, when appropriate. For most calculations, either a stable condition has been reached before24hours or core damage has been predicted before24hours.SC-A6: Confirm that the bases for the success criteria are consistent with the operating philosophy of the plant.Many of the specific sequences that are being quantified assume fewoperator actions. By design, these sequences presume a lack of operator action and donot agree with the operating philosophy of the plant(e.g., emergency operating procedures (EOPs)).Incases in whichoperator action is being modeled, and in all cases involving system operation, significant effort has been made to ensurethat the analyses appropriately mimic the operation of the plant.Cases withambiguity or limitationsare noted.Additional effort has been takento look at the EOPs, havesenior staff review the analyses, have lead SOARCA analysts review the analyses, and so forth.SC-B1: Use realistic generic analyses evaluations.Forthis project, the use of realistic plant-specific analyses means that Capability CategoryIIIis being met, though the last clause in CategoryIIIaboutusing no assumptions that could yield conservative criteria is debatable.

8PRA Standard Supporting Requirement for Capability CategoryIIThis ProjectSC-B2:Do not use expert judgment except when sufficient information / analytical methods are unavailable

.Other than cases in whichMELCOR models are based on expert judgment, or judgment is used for selecting operator timings, these analyses do not use expert judgment. Some judgment will be inevitable when the analyses are translated to specific changes in the success criteriafor other, similar plants.SC-B3:Use analysis that is appropriate to the scenario and contains the necessary level of detail.This requirementis clearly met by the use of MELCOR on a sequence-by-sequence basis for the sequences being studied.SC-B4:Use appropriate modelsandcodes, and use them within their limits of applicability.MELCOR is not formally assessed in the same manner as a design-basis analysiscode, but it does undergo some of the same steps(e.g.,comparison of results against relevant experimental results). The documentation for this project provides some high-level information about this assessment but does not attempt to make a comprehensive argument for MELCOR's applicability.In general, MELCOR is consideredan appropriate tool for this application.In the case in which its applicability is most ambiguous (i.e., LBLOCA), the extent of calculation margin is addressed.SC-B5: Confirm that the analyses results are reasonable and acceptable.All analyses have been reviewed by multiple experienced engineers to confirm that the results are reasonable and acceptable.In addition, theresults for many analyses have been compared to similar analyses performed by the SOARCA project. The SOARCA lead PWR analyst reviewed all results in the interim report.Results for station blackout were compared to similar Westinghouse calculations.Results for Surry feed and bleed were compared to similar TRACE calculations.SC-C1: Document the analyses to support PRA applications, upgrades, and peer review.The analyses are being comprehensively documented.The judgment used in applying the analyses as the basis for making specific SPAR model changes will be documented separately.SC-C2: Document the overall analysis comprehensively, including consideration of a provided list of documentation areas.In general, the level of documentation being provided with these analyses is consistent with this Supporting Requirement.The one area that is currently weak is thediscussion of limitations of MELCOR.Specific MELCOR applicability assessments for each initiator are beyond the scope of this confirmatory analysis.SC-C3:Document the sources of model uncertainty and related assumptions.This has not been formally done,except that a general sense of modeling uncertainty prompted some of the additional analyses (e.g.,RCP seal LOCA model). Another aspect that has received consideration is the relationship between uncertainty and the margin in a given calculation. For example,MELCOR may have higher uncertaintyin the modeling of LBLOCAs.Of the 15Surry LOCA cases with a break size 15cm (6in.), the highest PCTfor a case that was deemed to be successful is 812degreesC (1,494degreesF), about 400degreesCbelow the core damage definition. This suggests that,for these cases, a higher degree of uncertainty is acceptable because there is significant margin.

94.MAJOR PLANT CHARACTERISTICSThefollowing subsections describe the aspects of the analyzed plants that are germane to the analysis performed in this report.4.1SurryPower StationTothe level of detail needed for this analysis, Surry Units1 and2 were considered to be identical. Each unit is a three-loop Westinghouse with a subatmospheric containment. Each hasthree high-head safety injection (HHSI) pumpsandtwolow-head safety injection (LHSI) pumps. The latter are also required for high-pressure recirculation (in order to provide sufficient net positive suction head (NPSH)to the high-head pumps when using the containment sump as a water source). The minimum technical specification refueling water storage tank (RWST)volume is 387,100gallons (gal) (1,470cubic meters (m 3)). The water source forthe emergency core cooling system (ECCS)automatically transfers from the RWST to the containment sump when the RWST water level drops below 13.5percent.

3This transfer operation takes 2.5minutes because ofthe timeit takesfor the sump isolation valves to fully open.

4The containment spray system in injection mode relies on twopumps ratedat3,200gpm (12.1m 3/min) per pump (which includes approximately 300gpm (1.14m 3/min)per pump of bleed-off flow

5) and draws from the RWST. Containment spray automatically actuates at 25pounds per square inch absolute (psia)(0.17megapascal (MPa))containment pressure, andthe operators are directed by theEOPs to secure(and reset)containment sprays once containment pressure drops back below 12psia (0.083MPa). The containment spray system in recirculation mode uses four pumps (two in containment and twooutside of containment) thatare eachrated at3,500gpm (13.2m 3/min) and take suctionfrom the containment sump.

6Table 2summarizes major plant characteristics.

3Note that the relationship between RWST volume and percentinventory is not intuitive because zero percentcorresponds to about 14,000gal (53m 3), 13.5percentcorresponds to 66,000gallons (250m 3), about 97percentcorresponds to the technical specification limit, and 100percentcorresponds to 399,000gal (1,510m 3).Also note that switchover requires two-out-of-four RWST low level signals coincident with Recirculation Mode Transfer switches selected in the proper position.

4The MELCOR input model does not model the effects of this delay in terms of RWST inventory reduction.

5This bleed-off flow goes to the suction of the outside containment recirculation spray pumps to ensure that adequate NPSH is available.

6Note that successful sump recirculation function requires containment heat removal through the recirculation spray system.

10Table 2Major Plant Characteristics for SurryCharacteristicValueDesign TypeThree-loop WestinghouseContainment TypeSubatmosphericPower Level2,546 MWt 1Number of HHSI PumpsThreeNumber of HHSI TrainsTwoShutoff Head for HHSI5,905 ft/ 2,560 psi(17.65 MPa)Lowest PORV Opening/Closing Setpoint2,350 psi (16.2 MPa) / 2,260 psi (15.6 MPa)Number of Cold-Leg AccumulatorsOne per loop (three total)Nominal Operating Pressure 2,250 psia (15.5 MPa)RWST VolumeTechnicalSpecification387,000 gal (1,470 m 3)1The power level used in this report is the power level beforethe October2010 measurement uncertainty recapture power uprateof 1.6%. In general, the results in this report are not expected to be sensitive to a power change of this amount, or the potential adjustments to protection system, control system, and operating procedure set-points associated with the change. However, this is a qualitative assertion, as only the calculations in Section 6.2consider a higher power level (and in that case it is a much higher power level).4.2Peach BottomAtomic Power StationAs with Surry, to the level of detail needed for this analysis, Peach Bottom Units2 and3 wereconsidered to be identical. Both are General Electric BWR/4swith Mark-I containment.Peach Bottom's reactor core isolation cooling (RCIC)system has a capacity of 600gpm (2.3m 3/min) at 150 to 1,150poundsper square inch gage (psig)(1.0 to 7.9MPa).Thehigh-pressure coolantinjection (HPCI)system capacity is 5,000gpm (18.9m 3/min).The condensate storage tank (CST)is the preferred source until alow level in the CST (less than5feet(1.5meters))causes an automatic switchover to the suppression pool.The RCIC and HPCI turbines will automatically trip with a highturbine exhaust pressure of 50psig and 150psig (0.34 and 1.03MPa),respectively.RCIC and HPCI systems will automatically isolate with a low steamline pressure of 75psig (0.51MPa).RCIC and HPCI pump bearings are rated for 210degreesF(99degreesC). The high-capacity low-pressure coolantinjection (LPCI)system has ashutoff head of 295psig (2.0MPa).The volume of the CST is 200,000gal (756m 3).The suppression pool has a technical specificationmaximum temperature limit of 95degreesF (35degreesC) and a volume of 127,300cubic feet (3,605m 3).Majorplant characteristicsare summarized in Table 3.Table 3Major Plant Characteristics for Peach BottomCharacteristicValueDesign TypeGeneral Electric Type 4Containment TypeMark 1Power Level3,458 MWtRCIC Capacity600 gpm (2.27 m 3/min)HPCI Capacity5,000 gpm (18.9 m 3/min)Lowest SRV Opening/Closing Setpoint in Relief Mode1,133.5 psid 1(7.81 MPa) /1,099.5 psid(7.58MPa)Nominal Operating Pressure1,050 psia (7.24 MPa)Suppression Pool Inventory952,000 gal (3,605 m 3)1Pounds per square inch differential (psid)is the differential pressure in psi between the main steamline and the wetwell.

115.MELCOR MODEL5.1Plant RepresentationThe Surry and Peach Bottom models used for this analysis are based on the models utilized in the SOARCA study.Efforts to ensure that the models appropriately reflect the as-built,as-operated plant includeddiscussions with plant operation and engineering staff, site visits, and review of plant documentation and operating procedures. Detailed documentation of the models will be provided in the near future as part of that project and is therefore not duplicated in this report. In some cases, additional information (e.g.,additional containment spraytrip logic) was added to the SOARCA model to address systems and sequence characteristics needed for this study that were not needed for the SOARCAstudy. ForRCP seal leakage,the models used in this analysisdiffer from those used in the SOARCA analysis. The modeling of RCP seal leakage is described in the section onthe Surry station blackout analysislater in this report (Section6.4).Below is a briefoverview of the Surry and Peach Bottommodels, followed by some discussion of MELCOR's validation base.AppendixAofthis report outlines the basic features of the Surry model. Included are the reactor trip signals modeled, the ECCS injection setpoints, the HHSI and LHSI pump curves, details of the switchover of ECCS suction from the RWST to the containment sump, accumulator characteristics, containment spray system characteristics, containment fan cooler characteristics, and relief valve setpoints.Figure2showsa plan view ofthe MELCOR model for the Surry reactor coolant system (RCS).All three RCS loops are modeled individually.The detailed nodalization of the RCS loop piping as well as the reactor core and vessel upper plenum allows modeling of the in-vessel and hot-leg counter-current natural circulation during core heatup. This feature has been shown to be relevant even within the temperature ranges of interest in the analysis (i.e.,those preceding core damage).The RCPs are tripped on power failure or voiding(related to pump vibration)in the loop.7The core region is nodalized into 10axialthermal response nodes (the MELCOR core package (COR)) mapped to 5axial hydrodynamic volumes(the MELCOR control volume hydrodynamics package (CVH)), and is comprised of 5 radial rings.Safety systems are modeled using injection points, and the relevant portionsof the reactor protection system and control systems are modeled using MELCOR control functions. For the secondary side, both turbine-driven auxiliary feedwater (TD-AFW) and motor-driven auxiliary feedwater (MD-AFW)are modeled (including provisions for water level control). The core decay power is based on a number of ORIGEN calculations for each radial ring.The containment is divided into nine control volumes representing the major compartments. Containment sprays and fancoolers are also modeled.

7Since the present analyses do not credit operator actions to trip the RCPsearly in the transient (for cases in whichprocedures would direct this action), a global void fraction in the vicinity of the pumps of 10percentis selected to represent a conditionin whichpump cavitation would prompt shutdown of these pumps. Asystem-level code such as MELCOR does not have the capability to directly model actual pump performance under degraded conditions.

12Figure2Plan Viewof the Surry MELCOR RCS ModelFigure 3shows a schematic of the Peach Bottom MELCOR model,including the reactor pressure vessel(RPV), wetwell,and safety systems.The drywell (not shown) has four control volumes representing the pedestal, lower drywell, upper drywell, and upper head regions.The vessel (excluding the core region) is represented by seven control volumes with connections to various safety systems,including control rod drive injection (CRD), RCIC, HPCI, low-pressure core spray (LPCS), and residual heat removal (RHR)(vessel injection and containment cooling modes). The models for HPCI and RCIC include separate control volumes for the turbine exhausting into the suppression pool. All safety relief valves (SRVs),including dedicated automatic depressurization system (ADS)valves,are modeled with flowpaths on two steamlines (a single steamlineA and acombined steamline forB,C, andD). The core nodalization is similar to the Surry model,with 10axial levels (with a2:1COR:CVHratio) and fiveradial rings. Like the Surry model, the core decay power is based on a number ofORIGENcalculations for each radial ring.Because very fewchanges were made to the SOARCA model, AppendixBofthis report does not include the same introductory plantmodelinformation for Peach Bottom as AppendixAdoes for Surry.CV530-PRT Reactor Vessel (see detailed drawings)Steam Generator 1-RC-E-1B (see hot leg/steam generator Loop B detail)

Steam Generator 1-RC-E-1A (see hot leg/steam generator Loop A detail)

Steam Generator 1-RC-E-1C (see hot leg/steam generator Loop C detail)

CV220 CV235 CV101 CV153 CV201/CV222 CV202/CV 221 CV 401/CV422 CV 402/CV 421 CV437 CV440 CV237 CV240 CV435 CV420 CV500 CV520 Pressurizer FL241 FL242 FL243 FL243 FL244 FL442 FL443 FL444 FL441 FL201/FL240 FL202/FL292 FL231 FL431 CF206 Accumulator (Mass Source)Accumulator (Mass Source)CF206 HHSI LHSI CF222 CF218 RWST Suction (Mass Source)

Containment Sump Suction(Mass Source)

CF942 FL401/FL440 FL402/FL492 LHSI HHSI CF222 CF218 RWST Suction (Mass Source)

Containment Sump Suction(Mass Source)

CF942 Rupture Disk FL540 PRT Cubicle CV042 SRV-3 FL538 SRV-2 FL537PORV-1,FL530PORV-2,FL531 SRV-1 FL535 FL505 FL510 CV301/CV322 CV302/CV321 FL301/FL340 FL202/FL292 CV320 CV335 CV337 CV340 FL342 FL331 FL341 FL343 FL344 Accumulator (Mass Source)CF206 HHSI LHSI CF218 RWST Suction (Mass Source)

Containment Sump Suction(Mass Source)

CF942 CF222 RC Pump RC Pump RC Pump RCP Seal Leakage to SG A Cubicle - CV020FL901(21gpm)FL911(182gpm)FL921(500gpm)FL903(21gpm)FL913(182gpm)

FL923(500gpm)FL902(21gpm)FL912(182gpm)FL922(500gpm)Aux.Bldg.CV817FL997-Letdown Line Bypass RCP Seal Leakage to SG C Cubicle - CV040 RCP Seal Leakage to SG B Cubicle - CV030 Pressurizer Surge Line Loop B Loop C Loop A FL283 CV270 CV280 CV281 Turbine MSIV FL286 Steam Trap FL287SRV-2/3/4/5,FL264SRV-1,FL263PORV,FL262 CV998 Environment Flow Restrictor

FL285 FL288 CV295 TDAFW Pump FL483 CV470 CV480 CV281 Turbine MSIV FL486 Steam Trap FL487 Flow Restrictor

FL485SRV-1,FL463PORV,FL462 CV998 Environment FL488 CV295 TDAFW PumpSRV-2/3/4/5,FL464 CV370 CV380 CV281 Turbine MSIV FL386 Steam Trap FL387 Flow Restrictor

FL385SRV-1,FL363PORV,FL362 CV998 Environment FL388 CV295 TDAFW PumpSRV-2/3/4/5,FL364 FL383 FL893 Surge Line Creep Rupture Basement CV005 13-Figure 3Schematic of the Peach Bottom RCS NodalizationTo model failure of pressurizerpower-operated relief valves (PORVs) or SRVs, one of three approaches is used,as designated in the boundary condition descriptions for eachcase:(1)the relief valve cannot stick open, (2)the relief valve sticksopen on the first lift, or (3)the relief valve sticks open after nlifts, where nis a user-prescribed number. The purpose of the third approach is to provide intermediate results (relative to the two extremes), for assessing the variation in plant response. Generally speaking, the SPAR models treat the situation in a binary fashion-the valve is either stuckopen or it isnot.For the purposes of this analysis, a simplified treatment of valve cycling and failure is adopted for this intermediate situation. Table4andTable 5provide a synopsis of the basis for the values used for Surry, including the specific value used for each type of valve. These tables also provide comparative values from the Surry Individual Plant Examination and a more recent relief valve reliability study (NRC, 2011). The values in Table 5are tabulated using the following formulas:, or 14where P Dequals the probability of failure per demand and nequals the number of lifts. This report used amedian value (cumulative probability equal to 0.5). Two key limitationsassociated with the way this report treats failures due to valve cycling are (1) theuse of a constant failure probability per demand and (2) theassumption that the failure probability is the same regardless of whether the valve is passing steam, water, or a two-phase mixture.Table 4Comparison of Failure per Demand Probabilities for Surry Stuck-Open ValvesValveProbability of Sticking Open per DemandSurry Individual Plant ExaminationNUREG/CR-7037 Automatic Demand (Initial/Subsequent)NUREG/CR-7037 Liquid Demand (Initial/Subsequent) 1Circa 2006 Surry PRA(used in the present analyses)Pressurizer PORV0.01230.00495 / 0.002750.0625 / 0.0007150.0028Main Steamline PORV0.01230.00295 / 0.0109N/A0.0058Pressurizer SRV0.01230.5 / N/A 2N/A0.0027Main Steamline

SRV0.01230.0270 / 0.00254N/A0.0027 1No liquid demands were witnessed for the main steamline valves or for the SRVs, so no estimates could be given.

2Note that only fourdemands were observed. No subsequent demands on the pressurizer SRV were witnessed, so no estimates could be given. Table 5Comparison of Number-of-Lifts Values for Surry Stuck-Open ValvesValve# of Lifts for Cumulative Probability of Sticking Open = 0.5Surry Individual Plant ExaminationNUREG/CR-7037 Automatic DemandNUREG/CR-7037 Liquid Demand Circa 2006 Surry PRA (used in the present analyses)Pressurizer PORV56251880247Main Steamline PORV5664N/A119Pressurizer SRV561 2N/A256Main Steamline SRV56263N/A256 1No liquid demands were witnessed for the main steamline valves or for the SRVs, so no estimates could be given.

2Note that only fourinitial demands and no subsequent demands on the pressurizer SRV were witnessed, so this value is highly sensitive to the availability of sparse data.The value used for Peach Bottom was187lifts, which corresponds to a cumulative failure probability of 0.5 for a probability of failure per demand of 0.0037, in comparison to the values from NUREG/CR-7037,which gave an observed behavior after scram value of 956lifts. (These values correspond to a failure per demand of 7.08x10

-4and 7.25x10

-4,for initial and subsequent demands.) For the liquid demand, the observed behavior in NUREG/CR-7037 after scram had a value of 79lifts,which corresponds to a failure per demand of 8.77x10

-3.For both Surry and Peach Bottom, the values used for failure of the relief valves due to cycling may differ from the values used in the SOARCA study. It is also useful to point out that the valve temperatures 15associated with the high-temperature seizure failure mechanism being considered under the SOARCA study correspond to fuel temperatures reached after significant heatup (generally at or beyond the time of initial core damage). Since the present study only considers the phase of the accident up and until the start of core damage, this valve failure mechanism is not believed to be relevant for this analysis.5.2MELCOR Validation The MELCOR code is designed to run best-estimate accident simulations(NRC,2005). The code has been assessed against a number of experiments and plant calculations. The current test suite for MELCOR contains over 170separate input decks. MELCOR has been used forfinal safety analysis reportaudit calculations(related to engineered safety feature design and performance, containment design and performance, design-basis accident analysis, and severe accident analysis);the post-September11,2001,security assessments;and the SOARCA project. It has also been used to assess significance determination process issues. For these reasons,it is an ideal tool to use in this project.Specific experiments and plant calculations relevant to this project for which MELCOR has been assessed includethe following:Quenchexperiment11,simulating a small-break loss-of-coolant accident (SBLOCA) with late vessel depressurization to investigate response of overheated rods under flooding conditions(Hering, 2007)theThree Mile Island Unit2 accident(NRC, 1980)loss-of-fluid test(LOFT)LP-FP-2,simulating anLBLOCA(Adams, 1985)Russian Academy of Sciences MEI experiments involving a spectrum of LOCA sizes to study critical flow and vessel response(e.g., Dementiev, 1977)NEPTUN experiments to test pool boiling models and void fraction treatment(NRC,1992)General Electriclevel swell and vessel blowdown experiments characterizing single-and two-phase blowdown, liquid carryover,and water level swell(e.g., Appendix A to NRC,1981)General ElectricMarkIII tests with steam blowdown into the suppression pool investigating vent clearing and heat transfer modelscontainment thermal-hydraulic phenomena studied in various experimental facilities,including Nuclear Power Engineering Corporation for mixing and stratification(e.g., NUPEC,1993),Heissdampfreaktor for blowdown into containment, and Carolinas-Virginia Tube Reactor for steam condensation in the presence of noncondensables(SNL, 2008)small-scale experiments to test condensation models,including Wisconsin flat plate experiments(e.g., Huhtinemi, 1993)and Dehbitests

176.MELCOR RESULTSThedetailed results for Surry and Peach Bottomare provided in AppendicesA and B, respectively.

8The following subsections summarize these resultsin a standard format:(1)abrief description of the scenario, (2)a list of key assumptions andoperator actions, (3)a table of results, and (4)a table ofthetiming to key events.The analysis evaluated anumber of different scenarios. The following scenarios were analyzed for Surry:SBLOCAs to investigate the time available until RWST depletion and core damagefeedand bleed (during loss of all feedwater) to investigate the number of pressurizer PORVs and HHSI pumps neededsteam generator tube rupture (SGTR) events to provide updated accident sequence timingsstation blackout eventsto provide updatedaccident sequence timingsmedium-and large-break LOCAs to look at the systems needed for successful inventory controlduring the injection phaseThe following scenarios were analyzed for Peach Bottom:inadvertent openrelief valve cases to investigate the effects of various sources of high-pressureinjectionstation blackout eventsto investigate the time for alternating current (ac)power recovery, the time for suppression pool heatup, and the times associated with the loss of turbine-driven high-pressure systemsIn many cases,the analyzed sequence progressions make assumptions about the unavailability of systems and about operator actions that are not taken. These assumptions often stem from the particular sequence in the event tree that is being studied, which may not be the most probable sequence. In other cases, these characteristicsare not included because ofresource constraints. In all cases, the relevantsubsectionsbelow note these assumptions.Section6of this report places these analyses in the context of the associated SPAR models.6.1Small-Break Loss-of-Coolant AccidentDependency on Sump Recirculation (Surry)This series of cases investigates the timing to RWST depletion (and thus switchover to recirculation) for SBLOCAs in which operators take very few actions. In reality, the operators would enter procedureE-0, "Reactor Trip or Safety Injection" (e.g.,verify reactor and turbine trip, verify mitigative system availabilities and alignments), transition to E-1, "Loss of Reactor or Secondary Cooling" (e.g.,reduce RCS injection flow, initiate evaluation of plant status),and

8Plots of reactor vessel water level in AppendicesA and B show the actualwater level(i.e., they include two-phase effects where appropriate).

18later transition to ES-1.2, "Post LOCA Cooldown and Depressurization"(e.g.,dump steam to condenser, depressurize RCS to refill pressurizer).The varied parameters are break size (0.5in.(1.3 cm), 1in.(2.5 cm), and 2in.(5.1cm)), the assumption on relief valve sticking,and containment spray function (availableornot available). In all12casesinvestigated, the break location is the horizontal section of the cold leg. In addition, sensitivity cases were conducted to look at the effects of securing HHSI pumps (Cases2a and 6a) and performing secondary-side cooldown (Cases 2b and 6b). These sensitivity cases demonstrate the impact of HHSI and secondary-side cooldown on RCS pressure and RHR entry timing. Because of project resource considerations, the modeling uses a simplified scoping approach and does not necessarily represent the actual plant operating procedures.

9Table6For this reason, the results should be used with caution. Results are provided in ,Table7, and Table8.In addition to the key timing tables below, plots for various results of interestare provided in Appendix A, Section A.2.For the 2-in.(5.1-cm) breaksinvestigated, the RCSdepressurizesas a result ofthe break. The loss of high-head injection following RWST depletion (high-head recirculation was not modeled) furtherreducesthe primary side pressure to less thanthe maximumpressurefor LHSI recirculation;thus,HHSI recirculation is not necessary. The same is true for 0.5-in.(1.3-cm) breakswhen the PORV is assumed to stick open after 247lifts(seeTable 5)because this causes the0.5-in.(1.3-cm) breakto becomea 1.9-in.(4.8-cm)break.

10Note that operator action to reduce injection(in response to PORV cycling)and thus limit pressurizer PORV cycling was not modeled. Also note that some cases do include throttlingHHSI for the purpose of scoping operator actions to depressurize and cooldown. For the 0.5-in.(1.3-cm) cases in which the PORV does not stick open, the system does not depressurize. Finally, for the 1-in.(2.54-cm) cases, the break is not large enough to cause depressurization (because ofHHSI injection) and the PORV does not open. As a result, the system pressure is still high at the time of RWST depletion. Loss of HHSI at RWST depletion causes depressurization, but not enough to allow for LHSI recirculation.Key assumptionsand operator actionsin these calculations include the following:For the 0.5-in.(1.3-cm) breaks, the PORV sticks open after 247cyclesunless (1)it does not lift that many times (Case6b)or (2)noted otherwise (Cases7 and8).Operators do not throttle injection for the purpose of preventing valve chattering, which is relevant for 0.5-in.(1.3-cm) breaks.Operators do not take action to refill the RWST.Prior to RWST depletion, operators secure containment sprays (and reset to allow subsequent actuation) in accordance with the EOPsafter containment pressure drops below 12psia (0.083MPa).

9Specifically, for these cases the model assumes that at 30 minutes the third HHSI pump is securedand the steam generator(SG) PORV is opened on all three SGsto an opening fraction that will result in a cooldown ofapproximately 100 degrees F per hour(55.6 degrees C per hour)on the secondary side (corresponding to a similar cooldown on the primary side). This differs from the operating procedures that utilize more complex approaches depending on the exact situation (e.g., isolating HHSI, establishing and controlling normal charging flow, using pressurizer sprays).

10The equivalent diameter of the PORV is 1.39in.(3.53cm).

19RCPs trip at 10-percentvoiding(seeSection 5.1for more information on this modeling assumptionand Appendix D for more information on its effect).HHSI recirculation is not modeled. Operator actions for manual cooldown and depressurization are not modeled, except in a simplified manner forsensitivity Cases2b and 6b. MD-AFW and TD-AFW is available.

Accumulators are available.Note that this assumption is not typical for SBLOCA success criteria analysis. It is not expected to affect the end-state results, but could affect some intermediate timings.See Appendix D for additional information.Table6Surry SBLOCA Sump Recirculation Results CaseSize (inch)5 HHSIPumpsPORV TreatmentSpraysSecondary-SideCooldown Core Uncovery(hr)Core Damage(hr)1 1 3 N/A 0 No 9.2 1 11.9 1 2 2 7.3 1 9.9 1 2a 23/17.9 1 10.0 1 2b 33/1/0YesNo 4 No 4 3 2 3 0 NoNoNo42NoNo 5 0.5Sticks open after 247lifts0NoNo 6 2NoNo 6a 23/18.8 1 9.6 1 6b 33/1/0N/AYesNo 4 No 4 7 3Does not stick open 0 No 17.8 1 25.1 18214.4 1 21.4 1 1Core damage is an artifact ofthe assumed unavailability of HHSI recirculation.

2It is assumed thattwoHHSI pumpsare secured at 15minutes.

3It is assumed thattwoHHSI pumps are secured at 15minutes, and the third pump is secured at 30minutes,followed by secondary-side cooldown at 100degreesFper hour(55.6degreesCper

hour).4These cases reach RHR entry conditions (both temperature and pressure) beforeheatup.

51in.=2.54cm; 2in.=5.1cm; 0.5in.=1.3cm.Table7Surry SBLOCA Sump Recirculation Key Timings (Cases1-4)EventCase1(hr)Case2(hr)Case2a(hr)Case2b (hr)Case3(hr)Case4(hr)Reactor trip0.030.030.030.030.010.01HHSI injection0.030.030.030.030.010.01LHSI injection---2.02--First actuation of contain. sprays-2.653.29--1.76RWST depletion (<13.5%)5.834.305.80-3.122.63Spray recirculation-4.305.80--2.63LHSI recirculation----3.382.86Accumulator starts to inject6.004.525.830.820.230.23RCP trip (10% void)7.385.766.731.41--Core uncovery9.237.327.9---Core damage (max.temp.>2,200°F) 111.99.9310.0---

12,200°F=1,204 °C 20Table8Surry SBLOCA Sump Recirculation Key Timings (Cases5-8)EventCase5(hr)Case6(hr)Case6a(hr)Case6b(hr)Case7(hr)Case8(hr)Reactor trip0.010.010.010.010.010.01HHSI injection0.010.010.010.010.010.01LHSI injection---3.49--PORV stuck open0.830.834.65---First actuation of contain. sprays-2.205.30--3.23RWST depletion (<13.5%)4.143.437.45-8.175.52Spray recirculation-3.437.45--5.53LHSI recirculation4.723.97--26.6-Accumulator starts to inject4.143.437.141.118.285.65RCP trip (10% void)-4.685.0013.811.710.3Core uncovery--8.77-17.814.4Core damage (max.temp.>2,200°F) 1--9.61-25.121.4 12,200°F=1,204 °C6.2Feed-and-Bleed Power-Operated Relief ValveSuccess Criteria (Surry)The initiating event of interest for thesecalculationsis loss of main feedwater(LOMFW).Additionally, auxiliary feedwater is assumed unavailable. The parameter of interest is how many pressurizer PORVs need to be available for the feed-and-bleed procedure to be effectiveat removing decay heat. The injection source is HHSI (initially from RWST) and the bleed path is the PORVs. Repeated actuation of the PORV leads to an increase in the pressure in the pressurizer relief tank (PRT). Following failure of the PRT rupture disk, primary side coolant exiting the PORV passes into containment, resulting in an increase in containment pressure.Containment sprays actuate once containment pressure reaches the containment spray setpoint.For these analyses, nooperator actions are modeled except for securing containment sprays. Regarding the actual expected operator response for a loss of all feedwater event,the operators would enter E-0, "Reactor Trip or Safety Injection" (e.g.,verify reactor and turbine trip, verify mitigative system availabilities and alignments), transition to ES-0.1, "Reactor TripResponse" (e.g.,attempt to establish feed flow, control pressurizer pressure), and later enter FR-H.1,"Response to Loss of Secondary Heat Sink" (e.g.,establish feed from condensate system, manually initiate bleed and feed) based on the associated critical safety function status tree. For the purpose of determining the effectiveness of a single PORV for removing decay heat, the lack of operator action is conservative(i.e.,delayedinitiation of HHSI). However, these resultsshould be used with cautionfor determining the time to RWST depletion (and thereby switchover to recirculation) because for that aspectthis assumption may benonconservative(i.e.,earlier initiation of HHSI may lead to earlier RWST depletion depending on the interplay with containment spray actuation).The cause of the reactor trip is varied for three cases to scope the effect of the different trip criteria that exist for the set ofhigh-head three-loop Westinghouse plants in operation. In all cases, safety injection(SI)does not start until an auto-SI signal occurs due to high containment pressure. The power level is also varied to scope the effect of higher decaypower,becauseSurry has the lowestpower level of the high-head three-loop Westinghouse plants in operation. The cases that useda power level of 13.9percenthigher than Surry's power levelcorrespond to 21a power level of 2,900megawatts thermal(MWt),which corresponds to the upper range of the three-loop plants.The analysis performed here demonstratesthat one PORV provides a sufficient bleed path to maintain quasi-steady conditions on the primary side.

11Further, it is not necessary for the operators to manually open the PORV, as the HHSI at Surry will cause the valve to automatically open due to high pressure. Even in the absence of operator action, the capacity ofone HHSI pump is sufficient to remove decay heat for either the Surry or elevated (e.g.,VirgilC.SummerNuclear Station) power levels. Nevertheless, it is important to note that other differences between Surry and the higherpower-level three-loop plants (most notably the type of steam generator (SG)) have not been addressed.In the absence of further operator action, thesecasesdo eventually proceed to core damage in these analyses because HHSI recirculation (which would actuate upon RWST depletion) is not modeled.However, at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is available prior to RWST depletion, and an additional 3.5 to 4hours is available until core damage occurs. This timing information can be used to inform related sequences that include human failure events associated with refilling the RWST or aligning the HHSI water source to the containment sump.In addition to the resultsand key timings inTable9and Table10below, plots for various results of interestare provided in Appendix A, Section A.3.Key assumptionsand operator actionsin these calculations include the following:Prior to RWST depletion, operators secure containment sprays (and reset to allow subsequent actuation) in accordance with the EOPs after containment pressure drops below 12psia (0.083MPa).HHSI recirculation is not modeled;thus,the time to core damage is driven by RWST depletion (the timing of which is affectedby the assumption that operators do not take early action to start HHSI).The PORV is aligned for automatic operation and opens when the RCS pressure increases above thehigh pressure setpoint (i.e.,no manual operator action).For all calculations performed, the PORV had cycled roughly 200 times as of the time of core damage (i.e., fewer times than required for the valve to stick open for the cycling failure model used in this report). Were high-head recirculation to have been modeled, the valve would have eventually reached the required number of cycles for failure.Such a failure (if treated) would probably not impact the ability of the high-head safety injection pumps to maintain primary-side inventory, based on a qualitative assessment of the results in this section, as well as Cases 27 and 29 of Section 6.5.Manual RCS depressurization and cooldown is not modeled.RCPs trip at 10-percentvoiding; in actuality, FunctionRestorationProcedure FR-H.1would have the operators stop allRCPs. See additional information in Appendix D regarding the sensitivity to this assumption.

11Note that for Cases2 and 3, SRV1 briefly lifts because ofthe actuation of HHSI (PORV2 was disabled for the calculation). This brief actuation is judged to be inconsequential to the overall progression of the event.

22Table9Surry Feed-and-Bleed PORV Success Criteria Results CasePower Level 1Cause of Reactor Trip 2 Cause of SI# HHSI Pumps#of PressurizerPORVs Core Uncovery(hr)Core Damage 1 100%MFW tripHigh Cont. Press.11 No 3 No 3 2Low SG level + feed/steam mismatch1.65No 33113.9%Low-low SG level1.60No 3 1100percentequals 2,546MWt (Surry),and113.9percentequals 2,900MWt (Beaver Valley, Harris, and Summer); 2,900MWt is the highest present power level of the three-loop Westinghouse plants.

2Low SG level is <19percentof narrow-range span, while low-low SG level is <16percentof narrow-range span, based onTechnical Specification2.3-3 (January2008).

3Coreuncovery anddamagelate in the simulationareartifactsofthe assumed unavailability of HHSI recirculation.Table10Surry Feed-and-Bleed PORV Success Criteria Key TimingsEvent 1Case1 (hr)Case2 (hr)Case3 (hr)MFW, MD-AFW, TD-AFW unavailable000Reactor trip00.008(29 s)0.008(27 s)SGdryout1.110.630.58PRT rupture disk open1.560.970.93SI signal (containment pressure>1.22 bars)1.961.361.29RCP trip (10% void)2.051.431.35First actuation of containment sprays(containment pressure>1.72 bars)3.843.243.17RWST depletion (<13.5%)9.438.358.24Core uncovery10.90 21.65/ 9.54 21.60/ 9.42 2Core damage (max.temp.>2,200°F)13.5311.8011.68 11.22bars=0.122MPa;1.72 bars = 0.172 MPa;2,200°F=1,204°C.

2For Case 1, the core comes close to uncovering around the time of SI actuation, then later does uncover after the loss of HHSI.For Cases 2 and 3,the coreuncovers earlyin the accident,recovers prior to significant heatup, and later uncovers again (due to the loss of HHSI).6.3Steam Generator Tube Rupture Event TreeTiming (Surry)These calculations assessthe time available to take corrective actions for events involving spontaneous(as opposed to accident-induced orconsequential)tube rupture events. In addition to the resultsand key timingsin Table11andTable12below, plots for various results of interestare provided in Appendix A, Section A.4. For reference, the effective leak size of a one-tube rupture is about a1-in.(2.5cm)effective diameter. Past operating experience for SGTRevents suggests that, in some cases, the time between the initiating event and initiation of RHR can be significant (e.g.,this timing ranges from 3.25hours to 21.5hours forthe events covered in a study conducted in the mid-1990s) 12.Here, very few operator actions are assumed. In reality, the operators would be expected to enter E-0, "Reactor Trip or Safety Injection" (e.g.,verify reactor and turbine trip, verify mitigative system availabilities and alignments),transition to E-3, "Steam Generator Tube Rupture" (e.g.,initiate RCS cooldown, depressurize RCS and terminate SI to minimize primary-to-secondary leakage),and later transition to one of three post-SGTR procedures (based on plant conditions).

12"Steam Generator Tube Failures," NUREG/CR-6365, April1996.

23Even with fewoperator actionsassumed, the results provided below show that the availability of secondary-side heat removalallows asubstantial amount of time for corrective actions. At 24hours, thefuel temperatures for all fivecases are stable at less than 550degreesF(288degreesC), although additional actions would be eventually required (e.g., refilling the CST).For thefirst threecases, the faulted SGrelief valves are not allowed to stick open, despite cycling a large number of times (e.g., >15,000).For Cases 4 and 5,the faulted SG relief valve sticks open after 119 cycles(seeTable 5), which occurs within the first hour for both cases. Even in these cases, the availability of SIearly in the accident and MD-AFWlater in the accident results in times to core damage greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.Key assumptionsand operator actionsin these calculations include the following:Main steamlineisolation valves close on reactor trip.Operators secure eitheroneor twoHHSI pumpsat 15minutes (depending on the case)and manually controlauxiliary feedwater to maintain SGlevel (standard practice).For Cases 1 through 3, the faulted SGPORV does not stick openregardless of the number of lifts and regardless of whether it passes water. In all other situations, theSGPORVsstick open after119 cycles(seeTable 5).

HHSI recirculation is not modeled.RCPs trip at 10-percentvoiding.Manual isolation of the faulted SG is not assumed (i.e.,operators fail to perform this action).Manual actions to model long-term heat removal (EOP Emergency Contingency Action (ECA)3.1/3.2) are not modeled.Table11Surry SGTR Results Case No. Tubes HHSIPumpsSG PORV TreatmentTD-AFWMD-AFWNominal Break Flow Prior to Loss of HHSI (kg/sec)Core Uncovery(hr)Core Damage(hr)11 3/2Does not stick open 1 Yes 230No 3 No 32550-60No 3 No 3 3 13/123No 3 No 3 4 3/2Sticks open after 119 lifts30-40No 3 No 35560-70No 3 No 3 1Logic was added to address numerical instability (by limiting the flow area to smooth the liquid flow through the faulted SG PORV).

2TD-AFW is lost within the first hour for all cases due to flooding of the steamline.

3The response is based on a 24-hour mission time.

24Table12Surry SGTR Key TimingsEventCase 1 (hr)Case 2 (hr)Case 3 (hr)Case 4 (hr)Case 5 (hr)Reactor Trip0.0480.0120.0480.0480.012HHSI initiates (3 pumps)0.0510.0130.0510.0510.0131 of 3 HHSI pumps secured0.250.25N/A0.250.252 of 3 HHSI pumps securedN/AN/A0.25N/AN/ATD-AFW shut down 10.700.320.750.700.32Faulted SG PORV stuck openN/AN/AN/A0.760.35RWST depletion (<13.5%)

210.685.5814.068.414.69Accumulator injectionN/AN/AN/A8.620.94RCP trip (10% void)17.8111.7120.2012.445.02Emergency CST empty 3> 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s> 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s> 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s> 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s22.20Core damage> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1TD-AFW shuts down due to filling of the steamlineand flooding of the pump.

2Recall that since the RCS leak location is the ruptured SGtube(s),a substantial amount of water is expelled from the system via the SG relief valves (rather than into containment) and is thus unavailable for containment sumprecirculation.

3Depletion of the emergency CST(96,000 gal(363 m 3)), which is the normal injection source for AFW,stops MD-AFW.6.4Pressurized-Water ReactorStation Blackout (Surry)A number of simulations were run for station blackout sequences to investigate the effects of RCP seal failures, SRV operation, and TD-AFW availability and operation on the time available to recover ac power and re-establish core cooling.Along with the above variations in system conditions and responses, some other factorsthat affect the time to core damage are the time to battery depletion (loss of direct current (dc) power), the time to depletion of the emergency CST tank (for cases with TD-AFW available), the system pressure, and the occurrence of natural circulation (Case4).Cases4 and 6 assume dcpoweris always available, which mimics successful "blind feeding" of the SGs using TD-AFW following the loss of dc(see (West., 2008)for more information on this topic).Meanwhile, Cases9 and 10 assume the loss of TD-AFW at 4hours,which equals thestation blackout coping time for Surry from NUREG-1776, "Regulatory Effectiveness of the Station Blackout Rule," issued August2003(NRC,2003a).In theEOPs, the operators would first enter E-0, "Reactor Trip or SafetyInjection"(e.g.,verify reactor and turbine trip, verify ac emergency buses energized),which would direct them to ECA-0.0, "Loss of All AC Power"(verify AFW flow, try to restore power to any ac emergency bus).If acpower is recovered,the operators will transition to ECA-0.1, "Loss of All AC Power Recovery without SI Required" and/or ECA-0.2, "Loss of All AC Power Recovery with SI Required" (e.g., restore necessary injection sources, restore component cooling). If ac power is not recovered and the core-exit thermocouples rise past 1,200degreesF (649degreesC), the operators will transition to SACRG-1, "Severe Accident Control Room Guideline Initial Response" (e.g., check if RCS should be depressurized, determine containment spray requirements).The Surry SPAR model does not credit operation of auxiliary feedwater following battery depletion. Further, the SPAR model assumes core damage at the time of battery depletion(i.e.,no further opportunity for recovering ac power and averting core damage).This assumption exists because dc power is an integral part of ac power recovery, in that it provides the control power to operate electrical distribution systembreakers in order to bring electrical power into the power block following a station blackout. Alternate sources of dc control power 25are required once batteries are depleted in a station blackoutsequence, but this issue is not further explored in this report.The RCP seal leakage rates and timing are taken from the Westinghouse Owners Group (WOG) 2000 seal leakage model for "new" high-temperature seals used in the current Surry SPAR model, which is described in WCAP 15603, "WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs," issued May 2003 (West., 2003), as modified by the NRC staff's associated April 2003 safety evaluation report (NRC, 2003b).

13Table13The safety evaluation report for WCAP-15603 makes a few modifications to theWCAP-15603model, including the disallowance of credit for the third RCP seal. The resulting model has outcomes associated withfour possible leakage rates for use in PRAs, with the onset of increased leakage occurring at 13minutes in all cases.reproduces the leakage ratesand their conditional probabilities, along with some associated timings from the Westinghouse Emergency Response Guidelines as reproduced in the Surry SPARv3.52model documentationof July2008.The current analysis rancases for three of theseleakage sizes(21gpmper pump(0.079m 3/min), 182gpm per pump (0.689m 3/min),and500gpm per pump (1.89m 3/min)).14Table13Reactor Coolant Pump Seal Leakage DetailsSeq.#Leak Rate at >13 Minutes (gpm)2 Conditional ProbabilityTime to Core Uncovery Based on Westinghouse Emergency Response Guidelines 1Without DepressurizationWith Depressurization1210.79~13hours~22hours3760.01~7hours~9hours21820.1975~3hours~5hours44800.0025~2hours~2.5hours 1Assumes availability of TD-AFW 221gpm=0.079 m 3/min; 76gpm=0.29 m 3/min; 182gpm=0.689 m 3/min; 480gpm=1.82m 3/minThe results of the present analysis are in good agreement with thosefrom the Westinghouse Emergency Response Guidelines(Table13). For analogous cases (i.e.,those with TD-AFW available and no secondary-side depressurization)the following conditions apply:Time to core uncovery is about 1.5hours for the largest leakage rateof 500gpm/RCP(1.89m 3/min/RCP), as compared to 2hours in the Westinghouse calculations.Time to core uncovery is about 4hours for the intermediate leakage rate of 182gpm/RCP (0.68m 3/min/RCP), as compared to 3hours in the Westinghouse calculations.Time to core uncovery is about 13hours for the normal leakagerateof21gpm/RCP(0.079m 3/min/RCP), which is identical to the Westinghouse calculations.

13This is the same model that is invoked in a later PRA guidance topical report, WCAP-16141, "WOG2000 RCP Seal Leakage PRA Model Implementation Guidelines for Westinghouse PWRs," issued August2003.

14In accordance withconvention, these leak rates correspond to full system pressure. Actual leak rates will be substantially lower once system pressure decreases. Note that the figures for RCP seal leakage in AppendixA are designed to demonstrate this fact. An unfortunate side effect of plotting these leakage rates as a volumetric flow rate (as opposed to a mass flow rate) is that the plots go offscale once the flow becomes twophase.

26The current MELCOR calculations demonstrate an additional 0.5 to 3hours between the time of core uncovery and the time of core damage.Topical report WCAP-16396-NP, "WOG2000 Reactor CoolantPump Seal Performance for AppendixR Solutions," issued January2005(West.,2005),discusseswhy the NRC's safety evaluation of the WOG2000 model-and the WOG2000 model itself-result in conservative estimatesof RCP seal leak rates. These conservatismsare associated with both the leak rates assumedandthe timing of seal failure (which is reported to vary from 8minutes to 40minutes, as compared with the 13minutes used in the WOG2000 model). This topical reportquantitatively assessesthe effects of these conservatisms on accident progression timings (specifically,the time for loss of pressurizerlevel and core uncovery). The topical reportconcludesthat the conservatisms can substantiallyaffect the assessment of coping strategies, but that the conservatisms are "unlikely to affect any conclusions drawn from PRA models for internal events from at-power conditions"(West.,2005)These conclusions led to the decision not to request NRC review of a less conservative model. If applied in this case,these conclusions suggest that the timings to core damage calculated are conservative, but that these conservatisms will not affect the overall conclusions drawn from the models. Even so, the potential conservatisms could affect intermediate PRA results,such as the human error probability associatedwith aparticular action.Forthetiming of ac power recovery needed to avert core damage, two sensitivity cases were run for Case1:recovery of HHSI at 2.14hours (i.e.,at the onset of core damage based on a PCT of2,200degreesF(1,204degreesC))recovery of HHSI at 1.64hours (i.e.,half anhour beforecore damage)As shown in Figure5, the sensitivity case in which HHSI was recovered at 2.14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> occurred too late to avert fuel melting.For the case in which HHSI was recovered at 1.64 hours7.407407e-4 days <br />0.0178 hours <br />1.058201e-4 weeks <br />2.4352e-5 months <br />, recovery of injection was sufficient to avert fuel melting.

15Table14A best-estimate time could be developed by running calculations using an intermediate time (e.g., 15 minutes) for this case, as well as running similar sensitivities for other cases.In addition to the resultsand key timings in through Table17,and Figure4below, plots for various results of interestare provided in Appendix A, Section A.5. Key assumptionsand operator actionsin these calculations include the following:Operators manually controlauxiliary feedwater to maintain SG level (standard practice).DCpower is always available for control of TD-AFWfor Cases4 and 6 (i.e.,mimics successful blind feeding).Operatoractions to refill the emergency CST are not modeled.SRV sticks open on the firstlift for some cases (as specified below).

15Note that the sensitivity studies correspond to a case in whichthe seals fail at 13 minutes. As such, failure of the seals fromthermal shock upon recovery of seal cooling is not pertinent to this particular case.In addition, Surry has proceduralized isolation of the RCP seals as part of ECA-0.0, "Loss of All AC Power."

27For cases with RCP seal failure, failure is assumed to occur at 13minutes.

16Manual operator actions for rapid secondary-sidedepressurization are not modeled.Table14Surry Station Blackout Results CaseSeal Leakage Rate 1after Failure (gpm 3per pump)Seal Failure Time (min)SRV StuckOpenTD-AFWac/dc Core Uncovery (hr)Core Damage (hr)150013 N/A 2Fails to start-1.42.1 1aacrecovery at 2.1hours1.42.1 1bacrecovery at 1.6hours1.4-2Available

-1.62.3 321-Fails to start2.33.4 4Available; successful blind feeding13.316.3 5 1 stliftFails to start2.12.6 6Available; successful blind feeding13.013.8 718213 N/A 2Fails to start2.03.18Available3.94.8 921-Available; lost at 4hoursdclost at 4hours8.410.9101 s tlift8.18.8 1The leakage rate provided here is the leakage rate at full system pressure. As the system depressurizes, the leak rate decreases.

2The model is set to stick the valve open after 256lifts, but the valve does not lift that many times for these calculations.

3500gpm=1.89m 3/min;182gpm=0.689m 3/min; 21gpm=0.076m 3/min.

16Note that this differs from the seal failure model used inthe SOARCA project, which employed a more mechanistic approach (saturated conditions at the pump) to model the timing of seal failure.

28Figure4PCT Signaturesfor all Surry Station Blackout CasesTable15Surry Station Blackout Key Timings (Cases1-2)Event 1Case1 (hr)Case1a(hr)Case1b (hr)Case2 (hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Seal failure (500gpm/pump)0.220.220.220.22Primary side SG tubes water level starts to decrease0.520.520.520.52Primary side SG tubes dry0.960.960.960.98SG dryout1.161.161.16-Core uncovery 1.401.401.401.63Gap release1.921.92-2.15Core damage (max.temp.>2,200°F)2.142.14-2.25 1500gpm=1.89m 3/min; 21gpm=0.076m 3/min; 2,200°F=1,204°C.

29Table16Surry Station Blackout Key Timings (Cases3-6)Event 1Case3(hr)Case4(hr)Case5(hr)Case6(hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Primary side SG tubes water level starts to decrease1.925.381.525.42Emergency CST depleted-7.97-7.97Primary side SG tubes dry2.0311.301.6611.30SG dryout1.1911.771.1911.80SRV sticksopenN/AN/A1.4512.71Core uncovery 2.2813.312.0613.03Gap release2.9614.832.4213.60Core damage (max.temp.>2,200°F)3.4016.332.5713.80 121gpm=0.076m 3/min; 2,200°F=1,204°C.Table17Surry Station Blackout Key Timings (Cases7-10)Event 1Case7(hr)Case8(hr)Case9(hr)Case10(hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Seal failure (182gpm/pump)0.220.22--TD-AFW assumed lost at battery depletion--44Primary side SG tubes water level starts to decrease1.041.015.625.63Primary side SG tubes dry1.522.226.586.58SG dryout1.22-7.137.12SRV sticks openN/AN/AN/A7.67Core uncovery 1.983.888.378.10Gap release2.634.009.488.59Core damage (max.temp.>2,200°F)3.094.7710.858.77 1182gpm = 0.689m 3/min; 21gpm = 0.076m 3/min; 2,200°F = 1,204°C.

30Figure5Surry Injection Recovery Sensitivity Cases6.5Pressurized-Water ReactorMedium-and Large-Break Loss-of-Coolant AccidentInitial Response (Surry)The final set of Surry sequences investigatedcombinations of accumulators, HHSI,and LHSIfora spectrum of LOCA break sizesfor the early phaseof the accident(e.g.,the first fewhours).Break sizes from 2in.(5.1cm) to a double-ended break were analyzed, as shown in Table19.Although some calculations are simulated into the long-term cooling phase,the calculations are only intended to inform successcriteriafor the early injectionphase of the accident.By convention, the breakdown in the LOCA spectrum for most Westinghouse PWRs is 0.5in.(1.3cm) to 2in.(5.1cm) (SBLOCA), 2in.(5.1cm) to 6in.(15.2cm) (medium-break LOCA(MBLOCA)),and 6in.(15.2cm)and greater (LBLOCA). The break location for the current analyses is always the horizontal section of the cold leg in the pressurizer loop.Very few operator actions are modeled. In reality, the operators would enter E-0, "Reactor Tripor Safety Injection" (e.g.,verify reactor and turbine trip, verify mitigative system availabilities and alignments)and transition to E-1, "Loss of Reactor or Secondary Coolant" (e.g., check if containment sprays should be secured and reset, check if accumulators should be isolated). Depending on the course of the accident, theoperatorswould then transition to one of several ES-1.x series supplemental emergency procedures.As can be seenbelow, some of these accidents progress very quickly, with core uncovery taking place within the first minute (for LBLOCAs). Since quicklyevolving accidents can be more challenging to simulate from a thermal-hydraulic standpoint, it is of interest to look at the degree of margin between the PCT (for cases that are deemed successful) and the core damage definition being used.Table18presents these figures,demonstrating that the highest MBLOCA PCT (for a success case) is 483degreesF (268degreesC) fromthe core damage 0 500 1000 1500 2000 2500 300000.511.522.5 Time (hr)Maximum Core Temperature (K)

Case 1 Case 1a Case 1b 1477 K (2200 F)Injection recovery for Case 1bInjection recovery for Case 1a 31definition used here, and the highest LBLOCA PCT (for a success case) is 706degreesF(392degreesC) fromthe core damage definition. This demonstrates that there is significant marginin these cases, which helps tocounteract the additional model uncertainty that might be expected for these quicklyevolving accidents.In addition to the key timingsin Table20through Table26below, plots for various results of interestare provided in Appendix A, Section A.6. Table18PCT Ranges for Accumulator Success CasesRange of Break SizeRange of PCT for Success CasesRange of Margin: 2,200°FPCT(1,204°CPCT)MBLOCA (2in.to 6in.)575-1,717°F(302-936°C)483-1,625°F(268-902°C)LBLOCA (6in.to double-ended)575-1,494°F(302-812°C)706-1,625°F(392-902°C)The results in Table19are distilled hereto identify the minimal equipment needed to avoid core damageduring the injection phase.For MBLOCAs, the minimal equipment is the following:For 6-in.(15.2-cm)breaks,the analysesdemonstrate that anytwo of the followingthreewould be adequate:one HHSI, oneaccumulator in an intact loop, and one LHSI,with or without AFW.For 4-in.(10.2-cm) breaks, Case13 demonstrates that one accumulator in an intact loopandoneLHSI are not adequate, leaving tworemaining success pathsthat are successful for this break size:oneHHSI and one accumulator in an intact loop,orone HHSI and one LHSI, with or without AFW.For 2-in.(5.1-cm) breaks, both of the above criteria are sufficient,withor withoutAFW.The resulting minimal equipment successcriteria for the injection phasefor MBLOCAs is oneHHSI and either oneaccumulator in an intact looporoneLHSI.Note thatthe former criterionwould not be sufficient for the recirculation phase because LHSI is necessary to accomplish HHSI recirculation.AFW is not needed for success for an MBLOCAfor the injection phase; the break size is large enough to remove decay heat.For LBLOCAs, the minimal equipment is the following:For 6-in.(15.2-cm) breaks, the analyses demonstrate that any twoof the following threewould be adequate:oneHHSI, oneaccumulator in an intact loop, and oneLHSI,with or without AFW.For 8-in.(20.3-cm) breaks, Cases3,18, and 23confirm the above.For 10-in.(25.4-cm) breaks, Cases4,19,and 24confirm the above.For adouble-ended break, Case10 demonstrates that only LHSI is necessary. A case was not runto determine if one HHSI and oneaccumulator in an intact loop would have been sufficient.As noted above, such a combinationwould not permit recirculation.The resulting minimal equipment success criteria for the injection phaseforLBLOCAs areoneLHSI and either oneaccumulator in an intact looporoneHHSI.AFW is not needed for 32success for anLBLOCA; the break size is large enough to remove decay heat and the system fully depressurizes. Key assumptions and operator actionsin these calculations include the following:The break is in the horizontal section of the cold leg in the pressurizer loop.The RCPs trip at 10-percentvoiding.HHSI recirculation is not modeled.Operator actions to depressurize and perform secondary side cooldown are not modeled.Containment sprays are available for all cases(same actuation pressure and operator actions to secure as inSection 6.1and 6.2).Table19Surry MBLOCA and LBLOCA Results CaseBreak Size (in.)4# HHSIPumps#Accum.# LHSIPumpsAFW?1Time of Initial Core Uncovery (hr)Core Damage During InjectionPhase? (hr) 9 2100 Yes0.42No 2150210.410.73201100.42No 2211010.42No 327110 No0.38No 2291010.38No 3 1 4101 Yes0.09No111000.09No 2120010.100.27130110.100.27140210.10No221100.09No 225101 No0.09No281100.09No 2 2 6101 Yes0.04No50010.040.1660110.04No71000.070.2881100.08No 216101 No0.04No171100.06No 2260110.04No 3 8101 Yes0.02No181100.01No 2230110.03No 4 101010.01No191100.01No 2240110.02No 10Double-ended0010.02No 1Conventionally, AFW is not needed for success for an LBLOCA; the break size is large enough to remove decay heat and the system fully depressurizes.

2Note that core damage eventually occurs (or would occur, in cases in whichthe calculation was terminated early) because ofthe inability to go to HHSI recirculation (due to the unavailability of LHSI) 33or,more directly,from the lack of a low-pressure injection source.Recall that the present calculations are focused only on the injection phase success criteria.

3For these cases, core damage eventually occurs because HHSI recirculation is not modeled, and the pressure is not sufficiently low prior to core damage to allow for LHSI recirculation.

42in.=5.1cm; 4in.=10.2cm; 6in.=15.2cm; 8in.=20.3cm; 10in.=25.4cm.Table20Surry MBLOCA and LBLOCA Key Timings (2-in.Breaks)EventCase9(hr)Case15(hr)Case20(hr)Case21(hr)Case27(hr)Case29(hr)Reactor trip0.010.0030.010.010.010.01HHSI injection0.01-0.010.010.010.01RCP trip (10% void)0.280.070.280.280.180.17First actuation of containment sprays1.14-1.211.140.940.94Core uncovery (water<TAF)0.420.410.420.420.380.38LHSI injection---6.39-6.17Maximum claddingtemperature timing (max. temperature) 0.44(592K)0.73(1,477K 1)0.44(592K)0.44(592K)0.40(592K)0.40(592K)Core covered0.87N/A0.80.870.750.75 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage-2,200°F(1,204°C).Table21Surry MBLOCA and LBLOCA Key Timings (4-in.Breaks Group1)EventCase1 (hr)Case11(hr)Case12(hr)Case13(hr)Reactor trip0.0030.0030.0030.003HHSI injection0.0030.004--RCP trip (10% void)0.040.040.040.04First actuation of containment sprays0.080.080.070.07Core uncovery (water < TAF)0.090.090.100.10LHSI injection0.29-0.330.45Maximum claddingtemperature timing (max. temperature) 0.34(982K)0.53(1,209K)0.27(1,477K 1)0.27(1,477K 1)Core covered0.38>0.83N/AN/A 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage-2,200°F(1,204°C).

34Table 22Surry MBLOCA and LBLOCA Key Timings (4-in.Breaks Group 2)EventCase 14 (hr)Case 22 (hr)

Case 25 (hr)Case 28 (hr)Reactor trip0.0030.0030.0030.003HHSI injection-0.0040.0040.004RCP trip (10% void)0.040.040.040.03First actuation of containment sprays0.070.080.080.07Core uncovery (water < TAF)0.100.090.090.09LHSI injection0.73-0.30-Maximum claddingtemperature timing (max.

temperature) 0.73(1,183K)0.21(807K)0.32(1,054K)0.26(721K)Core covered0.790.390.390.41Table 23Surry MBLOCA and LBLOCA Key Timings (6-in.Breaks Group 1)EventCase 2 (hr)Case 5 (hr)Case 6 (hr)Case 7 (hr)Reactor trip0.0020.0020.0020.002HHSI injection0.002--0.002RCP trip (10% void)0.020.020.020.02First actuation of containment sprays0.030.030.030.03Core uncovery (water < TAF)0.040.040.040.07LHSI injection0.130.140.18-Maximum claddingtemperature timing(maximum temperature) 0.15(774K)0.16(1,477K 1)0.16(990K)0.28(1,477K 1)Core covered0.19N/A0.20N/A 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage-2,200

°F(1,204°C).

35Table 24Surry MBLOCA and LBLOCA Key Timings (6-in.Breaks Group 2)EventCase 8 (hr)Case 16 (hr)Case 17 (hr)Case 26 (hr)Reactor trip0.0020.0020.0020.002HHSI injection0.0020.0020.002-RCP trip (10% void)0.020.020.020.02First actuation of containment sprays0.030.030.030.03Core uncovery (water < TAF)0.080.040.060.04LHSI injection-0.13-0.18Maximum claddingtemperature timing(maximum temperature) 0.04(592K)0.152(775K)0.04(575K)0.13(931K)Core covered0.100.190.120.22Table25Surry MBLOCA and LBLOCA Key Timings (8-in.Breaks)EventCase3(hr)Case18(hr)Case23(hr)Reactor trip0.0020.0020.002HHSI injection0.0020.002-RCP trip (10% void)0.0090.0090.01First actuation of containment sprays0.010.010.01Core uncovery (water < TAF)0.020.010.03LHSI injection0.07-0.08Maximum claddingtemperature timing (maximum temperature) 0.10(851K)0.40(1,085K)0.07(792K)Core covered0.140.910.11Table26Surry MBLOCA and LBLOCA Key Timings (-in.Breaks)EventCase4 (hr)Case19(hr)Case24(hr)Case10(hr)Reactor trip0.0010.0010.0010.001HHSI injection0.0010.001--RCP trip (10% void)0.0080.0080.0060.001First actuation of containment sprays0.0080.0080.0080.005Core uncovery (water < TAF)0.010.0080.020.022LHSI injection0.04-0.050.005Maximum claddingtemperature timing (maximum temperature) 0.08(850K)0.30(835K)0.04(640K)0.036(1,043K)Core covered0.120.870.060.053 366.6Inadvertent Open Relief Valve Success Criteria (Peach Bottom)The first scenario of interestfor Peach Bottomdeals with an inadvertent/stuck-open relief valve. For most of thesesimulations, at time zero the reactor trips,feedwater trips, and a safety relief valve (SRV1) opens.In actuality, the plant would not be expected to trip, but would instead be manually tripped by the operators due to high suppression pool temperature if they were unable to reclose the stuck-open valve. Two sensitivity cases scope the effects of the assumption that the reactor and feedwater trip at time zero. LPCI is available for all casesand the main steam isolation valves (MSIVs)close shortly after reactor trip(see Table28)

17. The availability of RCIC, HPCI, and CRD injection is varied to assess their effects.This analysis modelsvery fewoperator actions. In reality, the operators would execute their procedures. A number of different procedure paths are possible, depending on availableequipment. In general, the following procedures would apply:Conditions will prompt the operators to attempt to reclose the open SRV.High suppression pool temperature will prompt the operators to start the residual heat removal system in suppression pool cooling mode in accordance with T-102, "Primary Containment Control."Low vessel level will prompt the alignment or recovery of frontline injection sources (e.g.,RCIC), and,if insufficient, alternative injection sources (e.g.,high-pressure service water) in accordance with T-101, "RPV Control,"and T-111, "Level Restoration,"along with supporting procedures.If conditions continue to degrade, the operators will perform an emergency depressurization to allow low-pressure injection.The calculationssummarized in Table27and Table28demonstratethatany of the injection options considered will prevent heatup beforedepressurization to LPCI entry. In the case of HPCI, the injection capacity is such that depressurization to LPCIentry doesnot occur for 9hours. For cases with only CRD injection, CRD prevents significant heatup even when the secondCRD pump is not started until 20minutes after the initiating event. For cases with no high-pressure injection, the system still depressurizes to LPCI entry conditions beforecore damagewould occur, with a maximum cladding temperature of 939 degreesC(1,722degreesF).Theaboveresults are subject to the assumption that suppression pool temperature is not significantly elevated by the time of natural depressurization to LPCI conditions,such thatlow-pressure injection drawing from the suppression pool would be unavailable due to NPSH concerns(i.e., the somewhat stylized nature of assuming the reactor trips at time zero).To investigate this assumption,twosensitivity caseswererun in which the reactor continues to operate at power until anautomatic trip signalis reached.These sensitivity cases were run for the more limiting of the CRD cases (Case 4). In Case 4a,feedwater is tripped at time zero,andin Case 4b,feedwatercontinues to run.For Case 4a, the reactor automatically tripped at eight seconds,leading to a PCT that is 110degrees Chigherthan in Case 4, but still more than 500degrees Cbelow the onsetof core damage.For Case 4b, because of the continued

17Due to the way the MSIVs are modeled in the MELCOR model, MSIVs effectively close when the pressure in the main steamline drops below 994 psia (6.85 MPa),ordue to lowRPV water level (for Case 4a).

37availability of feedwater, there is no core cooling concern and the reactor does not automatically trip until 46minutes.By the time the reactor trips (on high drywell pressure), the suppression pooltemperature has already exceeded multiple technicalspecificationlimits (as prescribedin Section3.6.2.1)that would have prompted operator action.Specifically, at 95degreesF(35degreesC),the technicalspecifications initiate increased monitoring and action to reduce temperature. At 110degreesF(43degreesC), the technicalspecifications require that the reactor be tripped "immediately." Finally, at 120degreesF(49degreesC), the technicalspecifications require the reactor to be depressurizedwith a completion time for this action of 12hours(NRC, 2003c).The significance of this case is that despite the assumption that the operators do not take the above actions, the suppression pool temperature does not reach the NPSH limit until greater than 5hours (the CRD continues to provide sufficient injection after this point).In addition to the key timing tables below, plots for various results of interestare provided in Appendix B, Section B.1. Key assumptions and operator actionsin these calculations include the following:Operator actions to reclose the SRV, start RHRin suppression pool cooling mode, and perform an emergency depressurization are either not initiated or are unsuccessful.Reactor trip(mimicking an early manual scram), feedwater trip,andoneSRVstuck open occurs at time zero(except for Cases4aand 4b).RCIC is run in inventory control mode.Post-scram CRD flow ranges from 110gpm (0.416m 3/min)at high pressure(1,020psia (7.0MPa)) to 180gpm (0.681m 3/min) at low pressure(14.7psia (0.1MPa)) for one pump, or 210gpm (0.795m 3/min) to 300gpm (1.14m 3/min) for twopumps 18.RCIC and HPCI isolate on low steamline pressure of 75psig (0.52MPa).

18No operator action is requiredto achieve the flow rate cited for one pump.Thepre-SCRAM flow rate is 60 gpm (0.227 m 3/min) for onepump. SCRAM results in the automatic opening of inlet valves on the individual CRD hydraulic control units, which increases the flow to 110 gpm (0.416 m 3/min) for onepump. The operators can also open throttle valves to increase the flow further, but this action is notconsidered in thisanalysis.

38Table27Peach Bottom Inadvertent Open SRV ResultsCaseRCICHPCICRDLPCILPCSac/dc FW, SPC, ADS Core Uncovery (hr)Core Damage (hr)1YesNo NoYesNoac/dc NoNoNo 2 NoYesNoNo 3 No1att=0 and 2att=10min0.41No 41att=0 and 2at20minafter scram0.37No 4a 10.29No 4b 1FWNoNo5NoNo0.32No 1For this case, the reactor was allowed to scram based on a reactor protection system trip signal, rather than at time t=0.Table28Peach Bottom Inadvertent Open SRV Key Timings (Cases1-5)EventCase1(hr)Case2(hr)Case3(hr)Case4(hr)Case4a(hr)Case4b (hr)Case5(hr)SRV1 open0000000Reactor trip0000< 0.01 10.760MSIVs close< 0.01< 0.01< 0.01< 0.01< 0.010.790Downcomer level first reaches

L20.070.070.070.070.03N/A0.07RCIC/HPCI first started (CST injection mode)0.080.08-----

2 n dCRD pump started--0.170.330.331.09-Downcomer level reaches L10.378.930.320.320.24N/A0.26Downcomer level below TAF0.378.930.350.330.25N/A0.28Suppression pool temp. >110degreesF 30.400.610.420.420.410.300.40LPCI first started0.518.930.590.580.53N/A0.57RCIC/HPCI pump isolation:

low steamline pressure

<0.52MPa(75psig)0.825.59-----HCTL limit reached 3(no action taken)4.54.0> 1 2> 1 25.00.57> 1 2RHR pump isolation -NPSH9.611.1> 1 2> 1 2> 10 25.4> 1 2Maximum cladding temperature timing (max.

temperature)

No heatup No heatup 0.78 (786K)0.76(830K)0.67(941K)Noheatup 0.75 (1,212K)1Reactor trips at 8secondson low RPV level.

2The simulation was stopped prior to reaching this condition.

3The HCTL limit is based on suppression pool temperature, suppression pool level, and RPV pressure.6.7Boiling-Water ReactorStation Blackout (Peach Bottom)These calculations investigate variations in the availability of injection sources,the behavior of theSRVs(failure to close), manual operator actions to implementheat capacity temperature 39limit (HCTL)-based depressurization 19, and the time to battery depletion.For reference, the Peach Bottom copingtime listed in NUREG-1776 is 8hours(NRC,2003a).Here, very fewoperator actions aremodeled. In reality, the operators would enter special event procedure SE-11, "Station Blackout,"based on plant conditions. This procedure would have the operators attempt to recover ac power from the grid and diesel generators and request configuration of the Conowingostation blackout line. The procedure would also direct the operators to shed loads to extend battery availability, take steps to extend HPCIor RCIC operation, and depressurize once plant conditions permitted. Concurrently, the EOPs would direct the operators to maintain level, stabilize pressure,and cooldown, as achievable.A sensitivity case was performedto look at the effect of recovery, similar to the Surry station blackout sensitivities described in Section6.4.Except as noted, mostcases assume that dcpoweris always available, which is an intentional modeling artifact to investigate timing.NoEOPmanual actions are modeledexcept forHCTL-based depressurization.For cases in whichboth HPCI and RCIC areunavailable, core damage occurs at 0.8 or 1.2hours, depending on the assumption aboutastuck-openSRV.Recovery of injection at the time of core damage was demonstrated to quickly arrest heatup. For cases in whichdcis lost after 2hours, core damage occurs at 4 to 5hours. For cases in whichthe SRV sticks open after 187lifts(see Table 5)or HCTL depressurization is performed, core damage ranges from 7to 11hours. (Note that the operators would initiate HCTL depressurizationto protect containmenteven without a low-pressure injection source.) For cases in whichthe SRV does not stick open and HCTL depressurization is not performed, RCIC or HPCI failsafter approximately 14 or 16hours(depending on which is assumed available)because ofCST depletion, and core damage occurs after 19hours.In these cases, switchover to the suppression pool is not permitted because the NPSH limit has already been exceeded 20Table29.Considering all cases, the time lag from uncovery of the top of active fuel (TAF) to the time of core damage ranges from 0.5 to 1.8hours.In addition to the results and key timingsin to Table32below, plots for various results of interestare provided in Appendix B, Section B.2. Key assumptions and operator actionsin these calculations include the following:RCICandHPCI (when available) are run in inventory control mode.DCpower is always available for control of HPCI and RCIC, except as noted.Post-accident alignment of CRD is not credited.

19For an SRV sticking open due to cycling, the lowest setpoint SRV (SRV/1) is the relevant SRV. For HCTL depressurization, the highest setpoint SRV (SRV/11) is the relevant SRV. Note that in the MELCOR model these valves have the same effectiveleak area.

20For cases such as these where RCIC or HPCI is operated for an extended period of time without room cooling, failure due to pump bearing temperature can become a concern. However, for the analysis performed here, temperatures did not approach the pump bearing rating assumed in the MELCOR model (210 degrees F (99 degrees C)).

40Table29Peach Bottom Station Blackout ResultsCaseRCICHPCIac/dcSRV Sticks Open?HCTL Depress ?Core Uncovery(hr)Core Damage (hr)1 No No-No 1 No0.51.2 1aacrecovery at 1.2hrNo0.51.2 22-At t=00.30.8 3 Yesdcis always available No17.719.44Yes6.07.252hr of dc No3.34.3 6dcis always availableAt 187lifts6.07.2 7NoYes No17.519.38Yes9.310.892hr of dc No3.84.9 10dcis always availableAt 187lifts9.210.7 1For this case, the SRV does not stick open until after core damage, so this assumption does not affect the outcome.

2Recovery of injection upon reaching 2,200degreesF(1,204degreesC) quickly arrestsfurther heatup.Table30Peach Bottom Station Blackout Key Timings (Cases1, 1a, and 2)EventCase1(hr)Case1a(hr)Case2(hr)Reactor trip, MSIVclosure000Downcomer level reaches L20.160.160.16Downcomer level reaches L10.500.500.27Downcomer level below TAF0.500.500.27Gap release: 900°C(1,652°F)1.021.020.69Core damage: max.temp.>1,204°C(2,200°F)1.171.170.79HPCI, RCIC, CRD injection start-1.17-ADS actuated-1.24-Downcomer level recovers above TAF-1.27-SRV sticks open due to high # of cycles1.75--

41Table31Peach Bottom Station Blackout Key Timings (Cases3-6)EventCase3 (hr)Case4 (hr)Case5 (hr)Case6 (hr)Reactor trip, MSIV closure0000Downcomer level first reaches L20.160.160.160.16RCIC started (CST injection mode)0.170.170.170.17RCIC fails due to loss of dc--2.00-HCTL limit reached2.46(no action taken) 2.462.46 (no action taken)2.46 (no action taken)SRV sticks open due to high # of cycles---2.47RCIC NPSH limit exceeded 112.67---RCIC pump isolation: low steam line pressure < 0.52MPa (75psig)-3.90-3.92RCIC injection ends due to CST level < 5ft(1.5m)14.43---Downcomer level reaches L117.685.613.255.62Downcomerlevel below TAF17.685.613.255.62Gap release: 900°C (1,652°F)19.066.994.047.00Core damage max.temp.>1,204°C (2,200°F)19.427.174.257.18Exhaust pressure exceeded:0.35MPa (50psig)20.14---1Switchover to the suppression pool is not permitted after this point.Table32Peach Bottom Station Blackout Key Timings (Cases7-10)EventCase7 (hr)Case8 (hr)Case9 (hr)Case10 (hr)Reactor trip, MSIV closure0000Downcomer level first reaches L20.160.160.160.16HPCI started (CST injection mode)0.170.170.170.17HPCI fails due to loss of dc--2.00-SRV sticks open due to high # of cycles---2.53HCTL limit reached2.67 (no action taken) 2.672.67 (no action taken)2.67 (no action taken)HPCI NPSH limit exceeded 112.07---HPCI pump isolation: low steam line pressure < 0.52MPa (75psig)-5.72-5.61HPCI injection ends due to CST level < 5ft (1.5m)16.05---Downcomer level reaches L117.538.973.828.94Downcomerlevel below TAF17.539.063.828.94Gap release: 900°C (1,652°F)18.9610.594.6310.46Core damage max.temp.>1,204°C (2,200°F)19.3110.84.8510.68Exhaust pressure exceeded: 1.04MPa (150psig)----1Switchover to the suppression poolis not permitted after this point.

437.APPLICATION OF MELCOR RESULTS TO SURRY AND PEACH BOTTOM SPARMODELSTable33andTable34below mapthe MELCOR calculations presented inSection6with the most closely corresponding SPAR model 21sequencesand providethe relative risk contribution of these sequences.Note that at the initiator heading level (e.g., LOMFW), the right-most column gives the relative contribution of all SPAR sequences from that initiator class (e.g.,

9.97percent), while the subsequent rows give the relative contributions from the subset of sequences studied in this report (e.g., LOMFW-16 = 9.32percent). Regarding loss of offsite power/ station blackout, the initiator class relative contribution is for all loss of offsite power events (e.g., switchyardcentered), whereas the analyses inthis report focus on station blackout events. Finally, for the station blackout sequences, the nomenclature of having multiple sequence numbers reflects transfers amongst two or more event trees. For instance, "LOOP-17-45" indicates the sequence with end-state #17 from the LOOP event tree, which transfers to the SBOevent tree and results in end-state #45 from that event tree. Allrelevant event treesare provided in Appendix C.It is also of interest to look at the quantitative timings to core uncovery and ac power recovery used in the Surry SPAR model relative to those from the present analysis (as provided in Section6.4). Table35provides this comparison.A key difference between the SPAR model and the present analyses arises for sequences with AFW available and a stuck-open relief valve. SPAR assumes that the relief valve sticks open early in the event, whereas in the presentanalyses,the relief valves are not challenged (when AFW is available) until much later (e.g.,8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />). This difference results in a very large delta in the timeto core damage. A second key difference is the SPAR assumption that offsite power must be recovered beforebattery depletion(i.e., no opportunity for preventing core damage following battery depletion), as compared to the present analysis in which the calculation is continued beyond battery depletion until the core damage surrogate is reached.

21When the majority of the analysis documented in this report was performed, the active versions of the SPAR models were v3.52 (Surry) and v3.50 (Peach Bottom). These are the models that are discussed in thissection. In the intervening evaluation and documentation phase of the project, these models were updated to the 8.xmodels presently used with SAPHIRE8, which included some data and unrelated modeling changes.As such, the relative risk contribution and sequence numbering for particular sequences would be different for these newer models, but the overall concepts and proposed modifications discussed in this section are unchanged. The actual model changes made as a result of thisproject were implemented in December2010 through February2011.

44Table33Mapping of MELCOR Analyses to the Surry1 & 2 SPAR(v3.52) ModelSPAR Sequence(see App. C)MELCOR CalculationsPercentage as Part of Initiator Class CDF (Internal Events)Percentage as Part of Total Internal Event CDFSBLOCA-Section 6.1of this report2.05%SLOCA-1Cases2b, 6bN/A-Success PathN/A-Success PathSLOCA-9Cases1, 2, 2a, 3, 4, 5, 6, 6a, 7, 81.05%0.02%LOMFW Feed andBleed-Section6.2of this report9.97%LOMFW-16 1All Cases93.39%9.32%SGTR-Section6.3of this report13.83%SGTR-12All Cases37.26%5.15%LOOP / Station Blackout-Section6.4of this report43.69%LOOP-17-42Cases6, 100.11%0.05%LOOP-17-15-7Case4<0.01%<0.01%LOOP-17-15-10Case90.06%0.03%LOOP-17-21Case80.05%0.02%LOOP-17-39Case2<0.01%<0.01%LOOP-17-45Cases1, 3, 5, 76.51%2.85%MBLOCA-Section6.5of this report1.70%MLOCA-6Cases1, 2, 7, 8, 9, 11, 20, 21, 2269.21%1.18%MLOCA-9Cases16, 17, 25, 27, 28, 29<0.01%<0.01%MLOCA-14Cases14, 15<0.01%<0.01%MLOCA-16Cases5, 6, 12, 13, 2617.41%0.30%LBLOCA-Section6.5of this report0.06%LLOCA-8Cases2, 3, 4, 5, 6, 7, 8, 10, 16, 17, 18, 19, 23, 24, 263.50%<0.01%

1The feed-and-bleed fault tree is used for many event trees.The relative contribution of the LOMFWsequence studied to the overall core damage frequency (CDF)is on the same order of magnitude or higher than the frequency associated with other sequences that include a failure of feed and bleed. The only other sequence with a higher CDF is a loss of acbus1J(22percenthigher). In addition, there is a non-station-blackoutLOOPsequence that includes failure of feed and bleed, and the summation of the four types of LOOP (e.g.,switchyard centered) results in a CDF equivalent to the LOMFW sequence.Note that the latter sequence uses a modified fault tree (FAB-L)specific to the LOOP event tree. All other sequences that include failure of feed and bleed are a factor of four or more lower.

45Table34Mapping of MELCOR Analyses to the Peach Bottom2 SPAR(v3.50) ModelSPAR Sequence (See App. C)MELCOR CalculationsPercentage as Part of Initiator Class CDF (Internal Events)Percentage as Part of Total Internal Event CDFInadvertently OpenRelief Valve-Section6.6of this report2.86%IORV-14Cases1, 2N/A-Success PathN/A-Success PathIORV-44Cases3, 4, 4a, 4b, 54.47%0.13%LOOP / Station Blackout-Section 6.7of this report5.75%LOOP-31-9Cases3, 4<0.01%<0.01%LOOP-31-30Case516.86%0.97%LOOP-31-45Case8<0.01%<0.01%LOOP-31-51Cases7, 90.51%0.03%LOOP-31-57Cases1, 1a2.14%0.12%LOOP-31-59-6Cases6, 100.01%<0.01%LOOP-31-59-7Case20.04%<0.01%Table35Comparison of Surry Station Blackout Results to the SPAR ModelSPAR(v3.52) ModelThis ReportConditionsSequence #SPAR Basisfor Timeto Core Uncovery(hr)Required Timefor Power Recovery(hr)Time to Core Uncovery(hr)Time to Core Damage(hr)AFW available w/ stuck-open SRVw/ 21gpm/RCP leakLOOP-17-420.518-139-14AFW available w/o stuck-openSRV w/ 21gpm/RCP leakLOOP-17-15-7/10154 18-1311-16AFW available w/o stuck-openSRV w/ 182gpm/RCP leakLOOP-17-213345AFW available w/o stuck-openSRV w/ 500gpm/RCP leakLOOP-17-39221.62.3AFW unavailableLOOP-17-450.511.4-2.32.1-3.4 1SPAR assumes a maximum time to recover power from station blackoutof 4hours, which is related to assumed battery depletion (and an assumed inability to control AFW or restore offsite power following loss of dc).Table36andTable 37below (1)summarizethe scenarios that have been investigated, (2)recap the boundary and initial condition variations studied using MELCOR, (3)highlight the relevant parts of the existing Surry and Peach Bottom SPAR success criteria, and (4)discusspotential changes to these models based on the MELCOR analysis(including identifying whether these changes have or have not been made).In addition,the table identifies cases in which these results wereapplied to SPAR models for other, similar plants.

46Table36Potential Success Criteria Updates Based on Surry ResultsInitiator/Aspect of InterestMELCOR VariationsAffected Portion of Existing SPAR ModelProposed/ActualChangesSBLOCA(Section6.1)Break size:0.5, 1, 2in.(1.3, 2.5, 5.1cm)# of containment spray pumps operating: 0, 2PORV treatment: sticks open at 247lifts, does not stick openSBLOCA sequence timing and mitigation success criteriaFor sequences without modeling of controlled cooldown via operator action, it has not been demonstrated that all break sizes will depressurize to RHR conditions beforeRWST depletion, or even core damage. Thus, HHSI recirculation is still required. Sensitivity studies have been performed for investigating the effects of controlled cooldown, but these calculations are not sufficient to justify changes to the SPAR models.These calculations demonstrate that the time between RWST depletion and core damage can be substantial, but are subject to the assumption that accumulators are available (see App. D).Feed and Bleed (Section6.2)Power levelReactor trip signalSuccess criteria for feedandbleed:2PORVsand 1HHSI trainThe analysis supports reduction of the number of required PORVs for Surry and similar plants 1from 2 to 1. Thischange(which has been made)alignsthe SPAR success criteria with the significance determination process notebooks and the licensee PRAs for all of these plants.SGTR(Section6.3)# of tubes ruptured: 1, 5# of HHSI pumps secured: 1, 2Faulted SG PORV treatment: sticks open at 119lifts, does not stick openSGTR event tree timingThe analysis performed demonstrates that (1)a single HHSI pump is sufficient for adequate injection and (2)significant time (>24hours) exists before core damage will occur(for the conditions studied),even with very littleoperator action and even though the RWST is depleted much earlier.The former item confirms the current treatment of HHSI in the success criteria. The latter item suggests that some specific sequences for which the failure to refill the RWST is an important factor may warrant revisitingin the context of a specific application,particularly in light of the fact that some of these sequences include human error probabilities that are driven by time-sensitive performance shaping factors.

47Initiator/Aspect of InterestMELCOR VariationsAffected Portion of Existing SPAR ModelProposed/ActualChangesStation Blackout (Section6.4)RCP seal leakage rate: 21, 182, 500gpm/pump (0.076, 0.689, 1.89m 3/min)SRV stuck-open:

1 stlift, neverTD-AFW: available, unavailable, blind-feeding successdc power: unavailable, depletes at 4 hr, always availableTime to recover acpower (and re-establish AFW cooling and RCS makeup capability)Table35provides a comparison of the timings between SPAR and the MELCOR analyses. In many cases, the MELCOR results confirm the current modeling assumption. In othercases, the timings suggest a potential toreduce conservatismin the context of a specific application.Sensitivity cases for this scenario suggest that recovery of acpower at 30minutes or more prior to core damage provides adequate time (from a thermal-hydraulic standpoint, as opposed to a system alignment standpoint) to establish high-pressureinjection and stop fuel heatup.The timing window needed for low-pressure sequences would be expected to be shorter, owing to the higher capacity injection that would be available.MBLOCA(Section6.5) 2Break size: 2, 4, 6, 8, 10in.,double-ended (5.1, 10.2, 15.2, 20.3, 25.4cm)# of HHSI pumps:

0, 1# of LHSI pumps:

0, 1# of accumulators:0, 1, 2AFW availabilitySuccess criteria for the injection phasefor the MBLOCA event tree:1HHSItrain and (1accumulator in eachintact loop or1AFWtrain)The MELCOR analysis suggests possible refinement of the success criteria. Nonetheless, the decision was made to retainthe existing SPAR success criteria.LBLOCA (Section6.5) 3Success criteria for inventory control during injection phase for the LBLOCA event tree: 1accumulator in eachintact loop and 1low-pressure injectiontrainBased on the MELCOR analyses, thesuccess criteria for the early stages of anLBLOCA has been changed for Surry and similar plants 1from twoaccumulators to oneaccumulator (in an intact loop) or oneHHSI train.

1 In this case, similar plants arethose with high-volume/high-head SI (chemical and volume control system)pumps(150gpm (0.568m 3/min) at2,500psi (17.2MPa)), large-volume SGs (series 51 and F) and core thermal power 2,900MWt; plants in this category are Beaver Valley1&2, J.M.Farley1 & 2, North Anna1 & 2, Harris, Summer, and Surry1 & 2.

2 Historically. MBLOCAs have been2-in.(5.1-cm) to 6-in.(15.2-cm) equivalent diameter(NRC,1990)and (NRC,1999)AppendixJ

.3 Historically, LBLOCAs have beengreater than 6-in.(15.2-cm) equivalent diameter (NRC,1990)and (NRC,1999)AppendixJ.

48Table 37Potential Success Criteria Updates Based on Peach Bottom ResultsClassMELCOR VariationsAffected Portion of Existing SPAR ModelProposed ChangesIORV(Section 6.6)Injection source: RCIC, HPCI, CRD, noneTiming of 2 nd CRDpump initiation:

10min,20minReactor/feedwatertrip: time zero vs. automaticEffectiveness of injection source for core cooling until low-pressure pumps can provide makeupFor RCICand HPCI,thecalculation confirms the treatment intheSPAR models.For CRD, the SPAR models for BWR Mark I and Mark II plants have been changed to credit:(1)two CRD pump flow for adequate core cooling following initial HPCI/RCIC success and (2)one CRD pump flow for adequate core cooling late in the event (if not already credited).Station Blackout(Section 6.7)Injection: HPCI, RCIC, noneOperator actions: HCTL depress.,

noneSRV behavior: stuck open at t=0, stuck open at 187lifts, never sticksRecovery time: 1.2hours, neverdcpower: unavailable, depletes at2hours, always availableTime to recover acpower (and reestablish core cooling)In general, the SPAR modeling conventions related to battery depletion limit the incorporation of insights from the MELCOR analyses. Some specific observations are provided below for consideration in the context of a specific application:For complete loss of acand dc, calculations suggest that credit for recovery of offsite power can be extended to 1hour (currently, the SPAR models givecredit for 30minutes).For complete loss of acand dccommensurate with a stuck-open SRV, calculations suggest that credit for recovery of offsite power can be extended to 30 minutes(currently,none of the SPAR models givecredit).For cases in which dcpower is always availableand RCIC/HPCI is lostbecause ofNPSH, current SPAR models agree with these results.For cases with 2hours of dc, calculations suggest that 2hours can be credited for boiloff (currently, none of the SPAR models gives anycredit for boiloff).Forthe maximum time for injection without suppression pool cooling (HCTL depressurization cases), the SPAR models agree with these calculations, with the exception of Grand Gulf and Nine Mile Point2 for RCIC.

498.CONCLUSIONThis projectdefined arealistically conservative core damage definition surrogate based on accident simulations. The project performed MELCOR analyses for two plants(Surry and Peach Bottom), looking at a range of initiating events and sequences. These results have been mapped to specific, realized changes for relevant SPAR models. TheNRCis continuing to work in this area and continues to seek opportunities to engage internal and external stakeholders.

519.REFERENCES(10 CFR,2007)U.S.Code ofFederal Regulations, "Domestic Licensing of Production and Utilization Facilities,Part50, ChapterI, Title10, "Energy."(Adams,1985)Adams, J.P. et al., "Quick Look Report on OECD LOFT Experiment LP-FP-2,"OECD LOFT-T-3804, September 1985.(Agencywide Documents Access and Management System (ADAMS) Accession No.ML071940358)(ASME/ANS,2009)American Society of Mechanical Engineers/American Nuclear Society, ASME/ANSRA-Sa-2009, "Standard for Level1/Large Early Release Frequency ProbabilisticRisk Assessment for Nuclear Power Plant Applications,", American Nuclear Society, LaGrange Park, IL,March2009.(Dementiev,1977)Dementiev, B.A. "Studying Hydrodynamics of Steam-water Media in Steady-state and Non steady-state Conditions for Nuclear Power Installations," Doctor-of-Science Thesis, Moscow Power Engineering Institute, Moscow,1977(in Russian).(Hering,2007)Hering, W.et al., "Results of Boil-off Experiment QUENCH-11," FZKA 7247, SAMLACOMERA-D-18, Forschungszentrum Karlsruhe, June 2007.(Huhtinemi, 1993)Huhtinemi, I.K. and Corradini, M.L.,"Condensation in the Presence of Noncondensable Gases,"Nucl. Engrg.Design,No.141, 429-446,1993.(NRC,1980)U.S. Nuclear RegulatoryCommission, "Three Mile Island:A Report to the Commissioners and to the Public," NUREG/CR-1250, January 1980.(NRC,1981)U.S. Nuclear Regulatory Commission, "BWR Refill-Reflood Program Task 4.8-Model Qualification Task Plan,"NUREG/CR-1899/EPRINP-1527/GEAP-24898, August 1981.(NRC,1988)U.S.Nuclear Regulatory Commission, "Decay Heat Removal Using Feedand Bleed for U.S. Pressurized-Water Reactors," NUREG/CR-5072, June1988.(NRC,1990)U.S.Nuclear Regulatory Commission, "Severe Accident Risks: An Assessment for Five U.S. Nuclear PowerPlants," NUREG-1150,December1990.(NRC,1992)U.S. Nuclear Regulatory Commission, "Boil-Off Experiments with the EIR-NEPTUN Facility: Analysis and Code Assessment Overview Report,"

NUREG/IA-0040, March 1992.(NRC,1999)U.S.Nuclear Regulatory Commission,"Rates of Initiating Events at U.S.Nuclear Power Plants: 1987-1995," NUREG/CR-5750,February1999.

52(NRC,2003a)U.S.Nuclear Regulatory Commission, "Regulatory Effectiveness of the Station Blackout Rule,"NUREG-1776, August2003.(ADAMS Accession No.ML032450542)(NRC,2003b)Memorandum from Herbert N. Berkowto Robert H. Bryan, "Safety Evaluation of Topical Report WCAP-15603, Rev.1, 'WOG2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs,'"May20,2003.(ADAMS Accession No.ML031400376)(NRC, 2003c)U.S. Nuclear Regulatory Commission, "Exelon Generation Company, LLC/PSEG Nuclear LLC Docket No. 50-277, Peach Bottom Atomic Power Station, Unit 2 Renewed Facility Operating License, Appendix A: Technical Specifications," May 2003.(ADAMS Accession No.ML052720266)(NRC,2005)U.S. Nuclear Regulatory Commission,"MELCOR Computer Code Manuals, Vol.1: Primer and User's Guide, Version1.8.6.2005,"NUREG/CR-6119, Rev.3, 2005.(NRC, 2009)U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic RiskAssessment Results for Risk-Informed Activities," Rev.2, March2009.(NRC, 2011)U.S.Nuclear Regulatory Commission, "Industry Performance of Relief Valves atU.S. Commercial Nuclear Power Plantsthrough 2007," NUREG/CR-7037, March 2011.(NUPEC,1993)Nuclear Power Engineering Corporation, "Specification of ISP-35-NUPEC's Hydrogen Mixing and Distribution Test M-5-5,"ISP35-027, Rev.1, November 3-4, 1993.(SNL,2008)Sandia National Laboratories, "An Assessment of MELCOR 1.8.6: Design Basis Accident Tests of the Carolinas Virginia Tube Reactor (CVTR) Containment (Including Selected Separate Effects Tests)," SAND2008-1224, February2008.(West.,2003)WestinghouseOwners Group,"WOG2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs,"WCAP-15603, Rev.1, Pittsburgh, PA, May2003.(ADAMS Accession No.ML021500485)(West.,2005)WestinghouseOwners Group, "WOG2000Reactor Coolant Pump Seal Performance for AppendixR Solutions,"WCAP-16396-NP,Pittsburgh,PA, January2005.(ADAMS Accession No.ML050320187)(West., 2008)Westinghouse Electric Company, "PRA Model for Blind Feeding a Steam Generator," B. Baron and R. Schneider, PSA 2008, Knoxville, TN, September 7-11, 2008.

APPENDIXASURRY MELCOR ANALYSES

A-1A.1Summary of Surry Model ChangesA.1.1Surry ModelThe Surry input deck is generally consistent with the model used in the State-of-the-Art Reactor Consequence Analyses (SOARCA) project. For the present application, however, a number of modifications are made to control the reactor trip and the engineeredsafety features,including the emergency core cooling system (ECCS)and containment sprays. In addition, the treatment of reactor coolant pump (RCP)seal leakage is alsodifferent. The general setpoints and operation of the systems are described below.Reactor TripTable 1indicates the conditions for reactor trip (i.e.,if any condition becomes true, then the reactor is tripped).Table 1Conditions for Reactor TripConditionComments1Loss of power2ECCS actuationSee ECCS signals3MFW trip 1Scram or loss of power or manual4TCV closure5RCP tripLoss of power or loop void >10%6HHSI activationECCS signal + power available7LHSI activationECCS signal + power available8High RCS pressure>2,400psia (16.55MPa)9Low RCS pressure<1,815psia (12.51MPa)10High PRZ level>44.97ft (26.26m in MELCOR model)11Low PRZ level<12.51ft (16.36m in MELCOR model)12High loop Dt>75°F (41.67°C)13ManualTime based 1 Several different configurations were found just among the three-loop high-head Westinghouse plants in terms of whether a main feedwater(MFW)trip would result in a turbine trip and subsequent reactor trip. The trip in the Surry MELCOR model is potentially dated.For these reasons, additional calculations were run in the loss of all feedwater section of this report to address the (perhaps more common)situation in whicha reactor trip would not occur until the reactor protection system trip signal(s) related to steam generator water level.Emergency Core Cooling SystemThe high-head safety injection (HHSI),as well as the low-head safety injection (LHSI)andcontainment sprays in the injection mode, draw water from the refueling water storage tank (RWST). Once the RWST is depleted, the LHSI suction is switched over to the containment sump.The ECCS actuation signals for HHSI and LHSI areas follows(i.e.,if any of the conditions are satisfied, then both systems are activated):pressurizer (PRZ)pressure (less than 1,775pounds per square inch gage (psig)(12.2megapascals (MPa)))

A-2high steamline differential pressure (greater than 120pounds per square inch differential (psid)(0.83 MPa))high containment pressure (greater than 17.7pounds per square inch absolute(psia)(0.122 MPa))manualoperator actionIn addition, power must be available.

Thecondition for high steam flow and either low steamline pressure (less than 525psig(3.62MPa)) or low average temperature (Tavg)(543degreesFahrenheit(F)(284degreesCelsius(C))) is not modeled.High-Head Safety Injection HHSI flows are delivered to the cold legs of the Surry model (control volumes240/340/440). All three HHSI pumps at Surry are assumed to start on HHSI activation.

1Table 2Total HHSI flow is portioned equally between the threecold legs. HHSI pump performance isgiven inbelowas HHSI flow per pump (gallons per minute (gpm))(Byron Jackson Test T-30705-3, 5-13-69).Table 2High-Head Safety Injection Flow per PumpFeet(Meters)gpm(m 3/min)Comment0(0)615(2.33)Runout1,600(488)550(2.08)2,500(762)500(1.89)3,275(998)450(1.70)3,950(1,204)400(1.51)4,500(1,372)350(1.32)4,950(1,509)300(1.14)5,300(1,615)250(0.946)5,600(1,707)200(0.757)5,800(1,768)150(0.568)Rated5,900(1,798)100(0.379)5,905(1,800)0(0)ShutoffLow-Head Safety InjectionLHSI flows are delivered to the cold legs of the Surry model (control volumes240/340/440). ThetwoLHSI pumps at Surry are assumed to start on LHSI activation. Total LHSI flow isportioned equally between the threecold legs.LHSI pump performanceis given in Table 3belowas LHSI flow per pump (gpm) (Byron Jackson Test T-31192-1, 11-10-69).

1The presentMELCOR model assumes that all threeHHSI pumps inject upon receiving asafety injection(SI)signal(onepump on the Hbus and two pumps on the Jbus).This (perhaps atypical) capability is based on interactions with the licensee and is corroborated by particular references (e.g.,the emergency operating procedures). The model does not account for the potential reduction in overall flow injection created by three pumps injecting through two trains.However, this modeling assumption is actually conservative in the present analysis because the three-pump alignment is only used for the small-break loss-of-coolant accident (LOCA)and steam generator tube rupture scenarios, in whichthe effect of RWST depletion and lack of system depressurization are more relevant than the core cooling (becauseadequate core cooling would be provided by fewer pumps).

A-3Table 3Low-Head Safety Injection Flow per PumpFeet(Meters)gpm(m 3/min)Comment0(0)4,000(15.1)Runout188(57.3)4,000(15.1)213(64.9)3,500(13.2)240(73.2)3,000(11.4)269(82.0)2,500(9.46)296(90.2)2,000(7.57)321(97.8)1,500(5.68)342(104)1,000(3.79)356(109)500(1.89)365(111)0(0)ShutoffThe RWST level must also be above 13.5percent(RWST-to-sump switchover starts at 13.5percentand takes 2.5minutes). After LHSI from the RWST is terminated, a model is activated for LHSI from the reactor sumpusing the same pump curve.Sump water availability and water temperature are checked.AccumulatorsAccumulators are also modeled as mass and enthalpy injected into cold leg component control volumes240,340, and 440.The initial water volume per accumulator is 975cubic feet(ft 3)(27.6 cubic meters (m 3))with an initialnitrogen cover gas volume of 475ft 3(13.5 m 3). The minimum operating pressure is given as 600psig (4.137MPa). All three accumulators are assumed to behave identically in that they are all modeled by a single set of control functions.Containment SpraysThe injection sprays use two pumps that can operate at 2,900gpm each (arated flow of 3,200gpm per pump minus bleed-off flow of 300gpm per pump) (11 m 3/min). The droplet size released by the spray headers is 1millimeter (mm). The pumps deliver water from the RWST at 45degreesF (280.4Kelvin(K)), themaximum temperatureallowedby the technical specifications,until the RWST water reaches 13.5percent. Thefollowingthree headers are in the dome:(1)the first at95.50feet(ft) (29.1 meter (m))elevation with 88nozzles(2)the second at142.40ft(43.4 m) elevation with 73nozzles(3)the third at143.75ft(43.82)elevation with 73nozzlesThe delay from spray signal to full operation is less than 15seconds. The recirculation sprays are modeled by twopumps identical to the injection mode. The cooler duty is 55,534,520British thermalunits per hour (BTU/hr)each (twoper pump,four total), which translates to 16.276megawatts (MW)per cooler (65.1MW total). Headers are common with those of the injection system. The containment sprays are initiated at a pressure of 25psia(0.17 MPa)andare secured(while in injection mode) when the pressure is less than 12psia(0.0827 MPa).Containment Fan Coolers A-4The recirculation system has three 75,000-standard cubic feet per minute(scfm)recirculation fans(35.4 cubic meters per second (m 3/s). The total volumetric flow rate is 225,000scfm (106.2cubic meters per second(m 3/s)). The system is supplied with 2,000gpm(135.91kilograms per second (kg/s)) at 70degreesF(294K) component cooling water until containment temperature and pressure are high enough or until pumps become submerged, at which point the systems goes to chilled cooling water. MELCOR monitors the liquid level, vapor temperature, and pressure in the lower dome (control volume50) and initiates the fans when the correct parameters are met. The fan inlet and discharge is within the "basement" (control volume5). The MELCOR input model containsa "low capacity" fan and a "high capacity" fan. Both recognize the same parameters, with the difference being thatthe high-capacity fan is turned on at a higher temperature. The low-capacity fans have a secondary coolant mass flow rate of 300 pounds per second (lb/s) (135.9kg/s),and the high-capacity fan has a secondary coolant mass flow rate of 831lb/s(377.1kg/s).ReactorCoolant Pumps The pumps operate at arated head of 280ft(85.3 m) of water at 650degreesF(343degreesC), 2,235psig, which is 6.80x10 5pascals. The pumps are tripped on either loss of power or high void (assumedto be10percentin the MELCOR model).Appendix Dprovides more information with respect to the sensitivity of the calculations to this latter assumption.Power-Operated Relief Valve and Safety Relief Valve Setpoints Theopening and closing pressures for the group of pressurizer power-operated relief valves (PORVs) and safety relief valves (SRVs) modeled in the MELCOR inputare given inTable 4below.Table 4Opening and Closing Pressures for PORV and SRVOpening Pressure in MPa (psi)Closing Pressure in MPa (psi)PORV-116.2(2,350)15.55(2,255)PORV-216.3(2,364)15.65(2,270)SRV-117.23(2,499)16.54(2,399)SRV-217.33(2,514)16.64(2,413)SRV-317.43(2,528)16.74(2,428)

A-5A.2Small-Break Loss-of-Coolant Accident Dependency on Sump RecirculationAnalysis SummaryTable 5through Table 7provide results for this portion of the analysis.Table 5Surry SBLOCA Sump Recirculation Results CaseSize (inch)5 HHSIPumpsPORV TreatmentSpraysSecondary-SideCooldown Core Uncovery(hr)Core Damage(hr)1 1 3 N/A 0 No 9.2 1 11.9 1 2 2 7.3 1 9.9 1 2a 23/17.9 1 10.0 1 2b 33/1/0YesNo 4 No 4 3 2 3 0 NoNoNo42NoNo 5 0.5Sticks open after 247lifts0NoNo 6 2NoNo 6a 23/18.8 1 9.6 1 6b 33/1/0N/AYesNo 4 No 4 7 3Does not stick open 0 No 17.8 1 25.1 18214.4 1 21.4 1 1Core damage is an artifact ofthe assumedunavailability of HHSI recirculation.

2It is assumed that twoHHSI pumps are secured at 15minutes.

3It is assumed that twoHHSI pumps are secured at 15minutes, and the third pump is secured at 30minutes, followed by secondary-side cooldown at 100degreesF(55.6degreesC) per hour).

4These cases reach RHR entry conditions (both temperature and pressure) before heatup.

51in.=2.54cm; 2in.=5.1cm; 0.5in.=1.3cm.Table 6Surry SBLOCA Sump Recirculation Key Timings (Cases1-4)EventCase1(hr)Case2(hr)Case2a(hr)Case2b (hr)Case3(hr)Case4(hr)Reactor trip0.030.030.030.030.010.01HHSI injection0.030.030.030.030.010.01LHSI injection---2.02--First actuation of contain. sprays-2.653.29--1.76RWST depletion (<13.5%)5.834.305.80-3.122.63Spray recirculation-4.305.80--2.63LHSI recirculation----3.382.86Accumulator starts to inject6.004.525.830.820.230.23RCP trip (10% void)7.385.766.731.41--Core uncovery9.237.327.9---Core damage (max.temp.>2,200°F) 111.99.9310.0---

12,200°F=1,204 °C.

A-6Table 7Surry SBLOCA Sump Recirculation Key Timings (Cases5-8)EventCase5(hr)Case6(hr)Case6a(hr)Case6b(hr)Case7(hr)Case8(hr)Reactor trip0.010.010.010.010.010.01HHSI injection0.010.010.010.010.010.01LHSI injection---3.49--PORV stuck open0.830.834.65---First actuation of contain.

sprays-2.205.30--3.23RWST depletion (<13.5%)4.143.437.45-8.175.52Spray recirculation-3.437.45--5.53LHSI recirculation4.723.97--26.6-Accumulator starts to inject4.143.437.141.118.285.65RCP trip (10% void)-4.685.0013.811.710.3Core uncovery--8.77-17.814.4Core damage (max.temp.>2,200°F) 1--9.61-25.121.4 12,200°F=1,204 °C.NotesFor Cases5and 6, PORV1cycles initially and then gets stuck open because ofthe number of cycles(247cycles). The equivalent diameter for the PORV is 1.387inches, so it depressurizes and goes to LHSI recirculation mode.Cases2a,2b,6a, and 6bare sensitivity calculations to demonstrate the impact of HHSI injection and secondary cooldown on reactor coolant system (RCS)pressureand to determine the residual heat removal (RHR)entry conditions. They maynot represent actual plant operating procedures.

A-7A.2.1Case1: 1-InchBreakLOCA without Sprays 00.5 11.5 22.5 301000020000300004000050000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 1801000020000300004000050000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-8 0 50 100 150 200 25001000020000300004000050000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 45000001000020000300004000050000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-9 00.20.40.6 0.8 11.21.4 1.61.8 201000020000300004000050000 time [sec]Containment Water Level [m]SumpCavity 0 5 10 15 20 2501000020000300004000050000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-10 300 800 1300 1800 2300 2800 330001000020000300004000050000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-11A.2.2Case2: 1-InchBreakLOCA with Sprays 00.20.40.6 0.8 11.21.4 1.61.8 20500010000150002000025000300003500040000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-12 0 10 20 30 40 50 60 70 800500010000150002000025000300003500040000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 00.050.10.150.20.250.30.350.40500010000150002000025000300003500040000 time [sec]Spray Flow (m3/s)Inj Recir A-13 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000500010000150002000025000300003500040000 time [sec]

Water (gal)RWST SPRLHSI HHSI 00.20.40.6 0.8 11.21.4 1.61.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]SumpCavity A-14 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 17000500010000150002000025000300003500040000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-15A.2.2.1Case2a: 1-InchBreakLOCA with Sprays and Secure Two HHSIPumps 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4 6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]

Pressure [MPa]

SG-A SG-B SG-C PRZ 450 psigRHR Entry Pressure (450 psig)

A-16 300 350 400 450 500 550 600 6500500010000150002000025000300003500040000 time [sec]

Average RCS Temperature [K]RHR Entry Temperature (350 F) 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Average CET Subcooling [K]

A-17 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Hot Leg Subcooling [K]

HL-a HL-b HL-c 0 10 20 30 40 50 60 70 800500010000150002000025000300003500040000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK A-18 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000500010000150002000025000300003500040000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]

SumpCavity A-19 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 28000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-20A.2.2.2Case2b: 1-InchBreakLOCA with Sprays, Secure HHSI Pumps, and Secondary Cooldown 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4 6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Pressure [MPa]

SG-A SG-B SG-C PRZ 450 psigRHR Entry Pressure (450 psig)Start 100 F/hr SG Depressurization A-21 300 350 400 450 500 550 600 650050001000015000200002500030000 time [sec]

Average RCS Temperature [K]RHR Entry Temperature (350 F) 0 10 20 30 40 50 60 70 80 90 100050001000015000200002500030000 time [sec]

Average CET Subcooling [K]

A-22 0 10 20 30 40 50 60 70 80 90 100050001000015000200002500030000 time [sec]

Hot Leg Subcooling [K]

HL-a HL-b HL-c 0 20 40 60 80 100 120 140 160 180 200050001000015000200002500030000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK A-23 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity A-24 0 5 10 15 20 25050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 2800 3300050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-25A.2.3Case3: 2-InchBreakLOCA without Sprays 00.20.40.60.8 11.21.41.61.80500010000150002000025000300003500040000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-26 0 50 100 150 200 250 300 350 400 450 5000500010000150002000025000300003500040000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000500010000150002000025000300003500040000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-27 00.20.40.6 0.8 11.21.4 1.61.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]SumpCavity 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-28 300 500 700 900 1100 1300 1500 170001000020000300004000050000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-29A.2.4Case4: 2-InchBreakLOCA with Sprays 00.20.40.60.8 11.21.41.61.80500010000150002000025000300003500040000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-30 0 50 100 150 200 2500500010000150002000025000300003500040000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 00.050.10.150.20.250.30.350.40500010000150002000025000300003500040000 time [sec]Spray Flow (m3/s)Inj Recir A-31 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000500010000150002000025000300003500040000 time [sec]

Water (gal)RWST SPRLHSI HHSI 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-32 300 500 700 900 1100 1300 1500 17000500010000150002000025000300003500040000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-33A.2.5Case5: 0.5-InchBreakLOCA without Sprays 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205000100001500020000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1805000100001500020000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZPORV Stuck Open A-34 0 20 40 60 80 100 12005000100001500020000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK+PORV 0 50000 100000 150000 200000 250000 300000 350000 400000 45000005000100001500020000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-35 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205000100001500020000 time [sec]Containment Water Level [m]

SumpCavity 0 5 10 15 20 2505000100001500020000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-36 300 500 700 900 1100 1300 1500 170005000100001500020000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-37A.2.6Case6: 0.5-InchBreakLOCA with Sprays 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205000100001500020000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1805000100001500020000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZPORV Stuck Open A-38 0 20 40 60 80 100 12005000100001500020000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK+PORV 0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.405000100001500020000 time [sec]

Spray Flow (kg/s)

InjRecir A-39 0 50000 100000 150000 200000 250000 300000 350000 400000 45000005000100001500020000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205000100001500020000 time [sec]Containment Water Level [m]

SumpCavity A-40 0 5 10 15 20 2505000100001500020000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 170005000100001500020000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-41A.2.6.1Case6a: 0.5-InchBreakLOCA with Sprays and Secure HHSI Pumps 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4 6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]

Pressure [MPa]

SG-A SG-B SG-C PRZ 450 psigRHR Entry Pressure (450 psig)

A-42 300 350 400 450 500 550 600 6500500010000150002000025000300003500040000 time [sec]

Average RCS Temperature [K]RHR Entry Temperature (350 F) 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Average CET Subcooling [K]

A-43 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Hot Leg Subcooling [K]

HL-a HL-b HL-c 0 10 20 30 40 50 60 70 800500010000150002000025000300003500040000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK A-44 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000500010000150002000025000300003500040000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]

SumpCavity A-45 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 28000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-46A.2.6.2Case6b: 0.5-InchBreakLOCA with Sprays, Secure HHSI Pumps, and Secondary Cooldown 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4 6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Pressure [MPa]

SG-A SG-B SG-C PRZ 450 psigRHR Entry Pressure (450 psig)Start 100 F/hr SG Depressurization A-47 300 350 400 450 500 550 600 650050001000015000200002500030000 time [sec]

Average RCS Temperature [K]RHR Entry Temperature (350 F) 0 10 20 30 40 50 60 70 80 90 100050001000015000200002500030000 time [sec]

Average CET Subcooling [K]

A-48 0 10 20 30 40 50 60 70 80 90 100050001000015000200002500030000 time [sec]

Hot Leg Subcooling [K]

HL-a HL-b HL-c 0 10 20 30 40 50 60 70 80 90 100050001000015000200002500030000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK A-49 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity A-50 0 5 10 15 20 25050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-51A.2.7Case7: 0.5-InchBreakLOCA without Sprays and PRZ PORV Not Stuck Open 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5020000400006000080000100000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18020000400006000080000100000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZRWST empty A-52 0 20 40 60 80 100 120 140 160 180 200020000400006000080000100000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK+PORV 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-53 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2020000400006000080000100000 time [sec]Containment Water Level [m]

SumpCavity 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-54 300 800 1300 1800 2300 2800020000400006000080000100000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-55A.2.8Case8: 0.5-InchBreakLOCA with Sprays and PRZ PORV Not Stuck Open 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 201000020000300004000050000600007000080000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1801000020000300004000050000600007000080000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZRWST empty A-56 0 20 40 60 80 100 120 140 160 180 20001000020000300004000050000600007000080000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK+PORV 0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.401000020000300004000050000600007000080000 time [sec]

Spray Flow (kg/s)

InjRecir A-57 0 50000 100000 150000 200000 250000 300000 350000 400000 45000001000020000300004000050000600007000080000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 201000020000300004000050000600007000080000 time [sec]Containment Water Level [m]

SumpCavity A-58 0 5 10 15 20 2501000020000300004000050000600007000080000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 280001000020000300004000050000600007000080000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-59A.3Feed-and-Bleed PORV Success CriteriaAnalysis SummaryTable 8andTable9provide results for this portion of the analysis.Table 8Surry Feed-and-Bleed PORV Success Criteria Results CasePower Level 1Cause of Reactor Trip 2 Cause of SI# HHSI Pumps#of PressurizerPORVs Core Uncovery(hr)Core Damage 1 100%MFW tripHigh Cont. Press.11 No 3 No 3 2Low SG level + feed/steam mismatch1.65No 33113.9%Low-low SG level1.60No 3 1100%equals 2,546MWt (Surry) and 113.9% equals 2,900MWt (Beaver Valley, Harris, and Summer); 2,900MWtis the highest present power level of the three-loop Westinghouse plants.

2Low SG level is <19% of narrow-range span, while low-low SG level is <16% of narrow-range span, based onTechnical Specification2.3-3(NRC, 2003).

3Coreuncovery anddamagelatein the simulationare artifacts ofthe assumed unavailability of HHSI recirculation.Table9Surry Feed-and-Bleed PORV Success Criteria Key TimingsEvent 1Case1 (hr)Case2 (hr)Case3 (hr)MFW, MD-AFW, TD-AFW unavailable000Reactor trip00.008 (29 s)0.008 (27 s)SGdryout1.110.630.58PRT rupture disk open1.560.970.93SI signal (containment pressure>1.22 bars)1.961.361.29RCP trip (10% void)2.051.431.35First actuation of containment sprays(containment pressure>1.72 bars)3.843.243.17RWST depletion (<13.5%)9.438.358.24Core uncovery10.90 21.65/ 9.54 21.60/ 9.42 2Core damage (max.temp.>2,200°F)13.5311.8011.68 11.22bars=0.122MPa;1.72 bars = 0.172 MPa;2,200°F=1,204°C.

2For Case 1, the core comes close to uncovering around the time of SI actuation and then later does uncover after the loss of HHSI. For Cases 2 and 3, the core uncovers early in the accident,recovers prior to significant heatup, and later uncovers again (due to the loss of HHSI).

A-60A.3.1Case1: 100-PercentPower, Reactor Trip at Time Equals Zero 0 2 4 6 8 10 12 14 16 1801000020000300004000050000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 10 12 14 16 18 20 220200040006000800010000 time [sec]SG Boiler Water Level [m]

A B C A-61 12 13 14 15 16 17 1801000020000300004000050000 time [sec]Reactor Pressure [MPa]

PRZCL-A0.0E+001.0E+05 2.0E+05 3.0E+054.0E+055.0E+056.0E+0501000020000300004000050000 time [sec]Integral Water Mass [kg]PORV1 (16.2/15.55 MPa)PORV2 (16.3/15.65 MPa)SRV1 (17.23/16.54 MPa)SRV2 (17.33/16.64 MPa)SRV3 (17.43/16.74 MPa)

A-62 0 2 4 6 8 10 12 14 16 1801000020000300004000050000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)0.0E+005.0E+041.0E+051.5E+052.0E+052.5E+053.0E+053.5E+054.0E+054.5E+0501000020000300004000050000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-63 0 5 10 15 20 2501000020000300004000050000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 170001000020000300004000050000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-64 00.20.40.6 0.8 11.21.4 1.61.8 201000020000300004000050000 time [sec]Containment Pressure [bar]SG-ADome 00.20.40.6 0.8 11.21.4 1.61.8 2050001000015000200002500030000350004000045000 time [sec]Containment Water Level [m]SumpCavity A-65A.3.2Case2: 100-PercentPower, Anticipatory Reactor Trip 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000350004000045000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 11 12 13 14 15 16 17 18 190200040006000800010000 time [sec]

SG Boiler Water Level [m]

A B C A-66 12 13 14 15 16 17 18050001000015000200002500030000350004000045000 time [sec]

Reactor Pressure [MPa]

PRZ CL-A 0.0E+00 1.0E+05 2.0E+05 3.0E+05 4.0E+05 5.0E+05 6.0E+05050001000015000200002500030000350004000045000 time [sec]Integral Water Mass [kg]

PORV1 (16.2/15.55 MPa)

PORV2 (16.3/15.65 MPa)

SRV1 (17.23/16.54 MPa)

SRV2 (17.33/16.64 MPa)

SRV3 (17.43/16.74 MPa)

A-67 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000350004000045000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj) 0.0E+00 5.0E+04 1.0E+05 1.5E+05 2.0E+05 2.5E+05 3.0E+05 3.5E+05 4.0E+05 4.5E+05050001000015000200002500030000350004000045000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-68 0 5 10 15 20 25050001000015000200002500030000350004000045000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700050001000015000200002500030000350004000045000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-69 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000350004000045000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]

SumpCavity A-70A.3.3Case3: 113.9-Percent PowerLevel, Reactor Trip on Low-Low SG Level 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000350004000045000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 11 12 13 14 15 16 17 18 190200040006000800010000 time [sec]

SG Boiler Water Level [m]

A B C A-71 12 13 14 15 16 17 18050001000015000200002500030000350004000045000 time [sec]

Reactor Pressure [MPa]

PRZ CL-A 0.0E+00 1.0E+05 2.0E+05 3.0E+05 4.0E+05 5.0E+05 6.0E+05050001000015000200002500030000350004000045000 time [sec]Integral Water Mass [kg]

PORV1 (16.2/15.55 MPa)

PORV2 (16.3/15.65 MPa)

SRV1 (17.23/16.54 MPa)

SRV2 (17.33/16.64 MPa)

SRV3 (17.43/16.74 MPa)

A-72 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000350004000045000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj) 0.0E+00 5.0E+04 1.0E+05 1.5E+05 2.0E+05 2.5E+05 3.0E+05 3.5E+05 4.0E+05 4.5E+05050001000015000200002500030000350004000045000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-73 0 5 10 15 20 25050001000015000200002500030000350004000045000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700050001000015000200002500030000350004000045000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-74 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000350004000045000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]Containment Water Level [m]

SumpCavity A-75A.4Steam Generator Tube Rupture Event Tree TimingAnalysis SummaryForSection A.4.1through SectionA.4.5,operators fail to (1)isolate faulted steam generator (SG), (2)depressurize and cool RCS, and (3)extend ECCS duration by refilling RWST or cross-connection to other unit's RWST. LoopA has the faulted SG.Table10and Table11provide results for this portion of the analysis.Table10Surry SGTR Results Case No. Tubes HHSIPumpsSG PORV TreatmentTD-AFWMD-AFWNominal Break Flow Prior to Loss of HHSI (kg/sec)Core Uncovery(hr)Core Damage(hr)11 3/2Does not stick open 1 Yes 230No 3 No 32550-60No 3 No 3 3 13/123No 3 No 3 4 3/2Sticks open after 119 lifts30-40No 3 No 35560-70No 3 No 3 1Logic was added to address numerical instability (by limiting the flow area to smooth the liquid flow through the faulted SG PORV).

2TD-AFW is lost within the first hour for all cases due to flooding of the steamline.

3Based on a 24-hour mission time.Table11Surry SGTR Key TimingsEventCase 1 (hr)Case 2 (hr)Case 3 (hr)Case 4 (hr)Case 5 (hr)Reactor Trip0.0480.0120.0480.0480.012HHSI initiates (3 pumps)0.0510.0130.0510.0510.0131 of 3 HHSI pumps secured0.250.25N/A0.250.252 of 3 HHSI pumps securedN/AN/A0.25N/AN/ATD-AFW shut down 10.700.320.750.700.32Faulted SG PORV stuck openN/AN/AN/A0.760.35RWST depletion (<13.5%)

210.685.5814.068.414.69Accumulator injectionN/AN/AN/A8.620.94RCP trip (10% void)17.8111.7120.2012.445.02Emergency CST empty 3>24 hours>24 hours>24 hours>24 hours22.20Core damage>24 hours 1TD-AFW shuts down due to filling of the steamline and flooding of the pump.

2Recall that since the RCS leak location is the ruptured SGtube(s),a substantial amount of water is expelled from the system via the SG relief valves (rather than into containment) and is thus unavailable for containment sump recirculation.

3Depletion of the emergency condensate storage tank (CST) (96,000 gal (363 m 3)), which is the normal injection source for auxiliary feedwater (AFW), stops MD-AFW.

A-76A.4.1Case1: One Tube and Secure One HHSI Pump 0 2 4 6 8 10 12 14 16 18020000400006000080000100000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 0 10 20 30 40 50 60 70 80 90 100020000400006000080000100000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)SGTR A-77 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPRLHSI HHSI 0 5 10 15 20 25020000400006000080000100000 time [sec]Flow Rate (kg/s)MFW (SG-A)MFW (SG-B)MFW (SG-C)TD-AFW (SG-A)TD-AFW (SG-B)TD-AFW (SG-C)MD-AFW (SG-A)MD-AFW (SG-B)MD-AFW (SG-C)

A-78 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700020000400006000080000100000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-79A.4.2Case2: Five Tubes and Secure One HHSI Pump 0 2 4 6 8 10 12 14 16 18020000400006000080000100000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 0 10 20 30 40 50 60 70 80 90 100020000400006000080000100000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)SGTR A-80 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPRLHSI HHSI 0 5 10 15 20 25020000400006000080000100000 time [sec]Flow Rate (kg/s)MFW (SG-A)MFW (SG-B)MFW (SG-C)TD-AFW (SG-A)TD-AFW (SG-B)TD-AFW (SG-C)MD-AFW (SG-A)MD-AFW (SG-B)MD-AFW (SG-C)

A-81 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700020000400006000080000100000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-82A.4.3Case3: One Tubeand Secure Two HHSI Pumps 0 2 4

6 8 10 12 14 16 18020000400006000080000100000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 0 10 20 30 40 50 60 70 80 90 100020000400006000080000100000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

SGTR A-83 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 5 10 15 20 25020000400006000080000100000 time [sec]

Flow Rate (kg/s)MFW (SG-A)MFW (SG-B)MFW (SG-C)TD-AFW (SG-A)TD-AFW (SG-B)TD-AFW (SG-C)MD-AFW (SG-A)MD-AFW (SG-B)MD-AFW (SG-C)

A-84 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700020000400006000080000100000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-85A.4.4Case4: One Tube,Secure One HHSI Pump and Stuck-Open SG PORV 0 2 4

6 8 10 12 14 16 18020000400006000080000100000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 0 10 20 30 40 50 60 70 80 90 100020000400006000080000100000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

SGTR A-86 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 5 10 15 20 25020000400006000080000100000 time [sec]

Flow Rate (kg/s)MFW (SG-A)MFW (SG-B)MFW (SG-C)TD-AFW (SG-A)TD-AFW (SG-B)TD-AFW (SG-C)MD-AFW (SG-A)MD-AFW (SG-B)MD-AFW (SG-C)

A-87 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700020000400006000080000100000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-88A.4.5Case5:FiveTubes,Secure One HHSI Pump, and Stuck-Open SG PORV 0 2 4

6 8 10 12 14 16 18020000400006000080000100000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 0 10 20 30 40 50 60 70 80 90 100020000400006000080000100000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

SGTR A-89 0 50000 100000 150000 200000 250000 300000 350000 400000 450000020000400006000080000100000 time [sec]

Water (gal)RWST SPR LHSI HHSI 0 5 10 15 20 25020000400006000080000100000 time [sec]

Flow Rate (kg/s)MFW (SG-A)MFW (SG-B)MFW (SG-C)TD-AFW (SG-A)TD-AFW (SG-B)TD-AFW (SG-C)MD-AFW (SG-A)MD-AFW (SG-B)MD-AFW (SG-C)

A-90 0 5 10 15 20 25020000400006000080000100000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 1700020000400006000080000100000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-91A.5Pressurized-Water Reactor Station BlackoutAnalysis SummaryThe station blackout sequence is similar to the SOARCA analysis. In all cases, there is 21gpmexisting leakage, but in some cases seal failure at 13minutesleads to either 182gpm or 500gpm leakage. Note that this is different from how RCP seal failure is modeled in the SOARCA project.Forthe modeling of stuck-open pressurizer SRV, there are two choices:(1)SRV sticks open based on number of cycles or (2)the valve does not recluse after the first lift-off.Note that none of the cases reach the 256-lift criterionbeforecore damage.SectionA.5.1reports two sensitivity calculations:(1)initiation of threeHHSI pumps at 2.14hours(when core damage occurs)and (2)initiation ofthreeHHSI pumps at 1.64hours (half an hour earlier).SectionA.5.1.1shows that core damage would continuefor the former case, while SectionA.5.1.2shows thatthere is sufficient time andinjection flow rateto avert fuel melting and arrest core heatupin the latter case.Table12through Table15below providesresults for this portion of the analysis.Table12Surry Station Blackout Results CaseSeal Leakage Rate 1after Failure (gpm 3per pump)Seal Failure Time (min)SRV StuckOpenTD-AFWac/dc Core Uncovery (hr)Core Damage (hr)150013 N/A 2Fails to start-1.42.1 1aac recovery at 2.1hours1.42.1 1bac recovery at 1.6hours1.4-2Available

-1.62.3 321-Fails to start2.33.4 4Available; successful blind feeding13.316.3 5 1 stliftFails to start2.12.6 6Available; successful blind feeding13.013.8 718213 N/A 2Fails to start2.03.18Available3.94.8 921-Available; lost at 4hoursdc lost at 4hours8.410.9101 s tlift8.18.8 1The leakage rate provided is the leakage rate at full system pressure. As the system depressurizes, the leakrate decreases.

2The model is set to stick the valve open after 256lifts, but the valve does not lift that many times for these calculations.

3500gpm=1.89m 3/min;182gpm=0.689m 3/min; 21gpm=0.076m 3/min.

A-92Table13Surry Station Blackout Key Timings (Cases1-2)Event 1Case1 (hr)Case1a(hr)Case1b (hr)Case2 (hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Seal failure (500gpm/pump)0.220.220.220.22Primary-side SG tubes water level starts to decrease0.520.520.520.52Primary-side SG tubes dry0.960.960.960.98SG dryout1.161.161.16-Core uncovery 1.401.401.401.63Gap release1.921.92-2.15Core damage (max.temp.>2,200°F)2.142.14-2.25 1500gpm=1.89m 3/min; 21gpm=0.076m 3/min; 2,200°F=1,204°C.Table14Surry StationBlackout Key Timings (Cases3-6)Event 1Case3(hr)Case4(hr)Case5(hr)Case6(hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Primary-side SG tubes water level starts to decrease1.925.381.525.42Emergency CST depleted-7.97-7.97Primary-side SG tubes dry2.0311.301.6611.30SG dryout1.1911.771.1911.80SRV sticks openN/AN/A1.4512.71Core uncovery 2.2813.312.0613.03Gap release2.9614.832.4213.60Core damage (max.temp.>2,200°F)3.4016.332.5713.80 121gpm=0.076m 3/min; 2,200°F=1,204°C.Table15Surry StationBlackout Key Timings (Cases7-10)Event 1Case7(hr)Case8(hr)Case9(hr)Case10(hr)Reactor trip, RCP trip, MFW/TD-AFW/MD-AFW0000Seal leakage (21gpm/pump)0000Seal failure (182gpm/pump)0.220.22--TD-AFW assumed lost at battery depletion--44Primary-side SG tubes water level starts to decrease1.041.015.625.63Primary-side SG tubes dry1.522.226.586.58SG dryout1.22-7.137.12SRV sticks openN/AN/AN/A7.67Core uncovery 1.983.888.378.10Gap release2.634.009.488.59Core damage (max.temp.>2,200°F)3.094.7710.858.77 1182gpm = 0.689m 3/min; 21gpm = 0.076m 3/min; 2,200°F = 1,204°C.

A-93A.5.1Case1: Station Blackout without Turbine-Driven Auxiliary Feedwater (500gpm)0 2 4 6 8 10 12 14 16 180200040006000800010000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 10 12 14 16 18 20 220200040006000800010000 time [sec]SG Boiler Water Level [m]

A B C A-94 0 50 100 150 200 250 300 350 400 450 5000200040006000800010000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow 00.20.40.6 0.8 11.21.4 1.61.8 20200040006000800010000 time [sec]Containment Pressure [bar]SG-ADome A-95 0 2 4 6 8 10 12 14 16 18 200200040006000800010000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 2800 33000200040006000800010000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-96A.5.1.1Case1a: Three HHSI Pumps at 2.14Hours 0 10 20 30 40 50 60 70 80 90 1000200040006000800010000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj) 0 2 4 6 8 10 12 14 16 18 200200040006000800010000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-97 300 800 1300 1800 2300 28000200040006000800010000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-98A.5.1.2Case1b: Three HHSI Pumps at 1.64Hours 0 10 20 30 40 50 60 70 80 9002000400060008000100001200014000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj) 0 2 4 6 8 10 12 14 16 18 20020004000600080001000012000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-99 300 500 700 900 1100 1300 1500 1700020004000600080001000012000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-100A.5.2Case2:Station Blackout with Turbine-Driven Auxiliary Feedwater (500gpm) 0 2 4 6 8 10 12 14 16 180200040006000800010000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 10 12 14 16 18 20 22 240200040006000800010000 time [sec]SG Boiler Water Level [m]

A B C A-101 0 50 100 150 200 250 300 350 400 450 5000200040006000800010000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow 00.20.40.6 0.8 11.21.4 1.61.8 20200040006000800010000 time [sec]Containment Pressure [bar]SG-ADome A-102 0 2 4 6 8 10 12 14 16 18 200200040006000800010000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 28000200040006000800010000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-103A.5.3Case3: Station Blackout without Turbine-Driven Auxiliary Feedwater (21gpm) 0 2 4 6 8 10 12 14 16 18 2002000400060008000100001200014000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ Hot leg creep rupture 10 12 14 16 18 20 220200040006000800010000 time [sec]

SG Boiler Water Level [m]

A B C A-104 0 5 10 15 20 25 30 35 40 45 5002000400060008000100001200014000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow 0.0E+00 2.0E+04 4.0E+04 6.0E+04 8.0E+04 1.0E+05 1.2E+05 1.4E+0502000400060008000100001200014000 time [sec]Integral Water Mass [kg]

PORV1 (16.2/15.55 MPa)

PORV2 (16.3/15.65 MPa)

SRV1 (17.23/16.54 MPa)

SRV2 (17.33/16.64 MPa)

SRV3 (17.43/16.74 MPa)

Hot leg creep rupture A-105 0 0.5 1 1.5 2 2.5 3 3.5 402000400060008000100001200014000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 5 10 15 20 2502000400060008000100001200014000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-106 300 800 1300 1800 2300 280002000400060008000100001200014000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-107A.5.4Case4: Station Blackout with Turbine-Driven Auxiliary Feedwater (21gpm) 0 2 4 6 8 10 12 14 16 18 200100002000030000400005000060000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 10 12 14 16 18 20 22 240100002000030000400005000060000 time [sec]SG Boiler Water Level [m]

A B C A-108 0 20000 40000 60000 80000 100000 1200000100002000030000400005000060000time [sec]Water Volume (gal)eCST 0 5 10 15 20 25 30 35 40 45 500100002000030000400005000060000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow A-1090.0E+001.0E+042.0E+043.0E+044.0E+045.0E+046.0E+047.0E+048.0E+049.0E+040100002000030000400005000060000 time [sec]Integral Water Mass [kg]PORV1 (16.2/15.55 MPa)PORV2 (16.3/15.65 MPa)SRV1 (17.23/16.54 MPa)SRV2 (17.33/16.64 MPa)SRV3 (17.43/16.74 MPa) 00.20.40.6 0.8 11.21.4 1.61.8 20100002000030000400005000060000 time [sec]Containment Pressure [bar]SG-ADome A-110 0 5 10 15 20 250100002000030000400005000060000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 17000100002000030000400005000060000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-111A.5.5Case5: Station Blackout without Turbine-Driven Auxiliary Feedwater (21gpm); Stuck-Open Relief Valve 0 2 4 6 8 10 12 14 16 18 200200040006000800010000 time [sec]Pressure [MPa]SG ASG BSG C PRZ 10 12 14 16 18 20 220200040006000800010000 time [sec]SG Boiler Water Level [m]

A B C A-112 0 5 10 15 20 25 30 35 40 45 500200040006000800010000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow0.0E+002.0E+044.0E+046.0E+048.0E+041.0E+051.2E+051.4E+051.6E+051.8E+050200040006000800010000 time [sec]Integral Water Mass [kg]PORV1 (16.2/15.55 MPa)PORV2 (16.3/15.65 MPa)SRV1 (17.23/16.54 MPa)SRV2 (17.33/16.64 MPa)SRV3 (17.43/16.74 MPa)SRV Stuck Open A-113 00.20.40.6 0.8 11.21.4 1.61.8 20200040006000800010000 time [sec]Containment Pressure [bar]SG-ADome 0 5 10 15 20 250200040006000800010000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-114 300 800 1300 1800 2300 28000200040006000800010000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-115A.5.6Case6: Station Blackout with Turbine-Driven Auxiliary Feedwater (21gpm); Stuck-Open Relief Valve 0 2 4 6 8 10 12 14 16 180100002000030000400005000060000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-116 10 12 14 16 18 20 22 240100002000030000400005000060000 time [sec]SG Boiler Water Level [m]

A B C 0 20000 40000 60000 80000 100000 1200000100002000030000400005000060000time [sec]Water Volume (gal)eCST A-117 0 20 40 60 80 100 120 1400100002000030000400005000060000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow0.0E+002.0E+044.0E+046.0E+04 8.0E+041.0E+051.2E+051.4E+051.6E+050100002000030000400005000060000 time [sec]Integral Water Mass [kg]PORV1 (16.2/15.55 MPa)PORV2 (16.3/15.65 MPa)SRV1 (17.23/16.54 MPa)SRV2 (17.33/16.64 MPa)SRV3 (17.43/16.74 MPa)

A-118 00.5 11.5 22.5 30100002000030000400005000060000 time [sec]Containment Pressure [bar]SG-ADome 0 5 10 15 20 250100002000030000400005000060000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-119 300 800 1300 1800 2300 28000100002000030000400005000060000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-120A.5.7Case7: Station Blackout without Turbine-Driven Auxiliary Feedwater (182gpm)0 2 4 6 8 10 12 14 16 18 20020004000600080001000012000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 12 14 16 18 20 22020004000600080001000012000 time [sec]

SG Boiler Water Level [m]

A B C A-121 0 20 40 60 80 100 120 140 160 180 200020004000600080001000012000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2020004000600080001000012000 time [sec]

Containment Pressure [bar]

SG-A Dome A-122 0 5 10 15 20 25020004000600080001000012000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 2800020004000600080001000012000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-123A.5.8Case8: Station Blackout with Turbine-Driven Auxiliary Feedwater (182gpm) 0 2 4

6 8 10 12 14 16 1805000100001500020000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 12 14 16 18 20 22 240200040006000800010000 time [sec]

SG Boiler Water Level [m]

A B C A-124 0 20 40 60 80 100 120 140 160 180 20005000100001500020000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205000100001500020000 time [sec]

Containment Pressure [bar]

SG-A Dome A-125 0 2 4 6 8 10 12 14 16 18 2005000100001500020000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 280005000100001500020000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-126A.5.9Case9: Station Blackout with Turbine-Driven Auxiliary Feedwater (21gpm) and 4-Hour Direct Current 0 2 4 6 8 10 12 14 16 18 200500010000150002000025000300003500040000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 12 14 16 18 20 22 240500010000150002000025000300003500040000 time [sec]

SG Boiler Water Level [m]

A B C A-127 0 20000 40000 60000 80000 100000 1200000500010000150002000025000300003500040000 time [sec]

Water Volume (gal) eCST 0 5 10 15 20 25 30 35 40 45 500500010000150002000025000300003500040000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow A-128 0.0E+00 2.0E+04 4.0E+04 6.0E+04 8.0E+04 1.0E+05 1.2E+050500010000150002000025000300003500040000 time [sec]Integral Water Mass [kg]

PORV1 (16.2/15.55 MPa)

PORV2 (16.3/15.65 MPa)

SRV1 (17.23/16.54 MPa)

SRV2 (17.33/16.64 MPa)

SRV3 (17.43/16.74 MPa) 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

SG-A Dome A-129 0 5 10 15 20 250500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 500 700 900 1100 1300 1500 17000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-130A.5.10Case10: Station Blackout with Turbine-Driven Auxiliary Feedwater (21gpm) and 4-Hour Direct Current; Stuck-Open Relief Valve 0 2 4 6 8 10 12 14 16 18 2005000100001500020000250003000035000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ 10 12 14 16 18 20 22 2405000100001500020000250003000035000 time [sec]

SG Boiler Water Level [m]

A B C A-131 0 20000 40000 60000 80000 100000 12000005000100001500020000250003000035000 time [sec]

Water Volume (gal) eCST 0 20 40 60 80 100 120 14005000100001500020000250003000035000 time [sec]

Seal Leakage (gpm)

Loop A Loop B Loop CStart of 2-phase flow A-132 0.0E+00 2.0E+04 4.0E+04 6.0E+04 8.0E+04 1.0E+05 1.2E+05 1.4E+0505000100001500020000250003000035000 time [sec]Integral Water Mass [kg]

PORV1 (16.2/15.55 MPa)

PORV2 (16.3/15.65 MPa)

SRV1 (17.23/16.54 MPa)

SRV2 (17.33/16.64 MPa)

SRV3 (17.43/16.74 MPa) 0 0.5 1 1.5 2 2.5 305000100001500020000250003000035000 time [sec]

Containment Pressure [bar]

SG-A Dome A-133 0 5 10 15 20 2505000100001500020000250003000035000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 280005000100001500020000250003000035000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F

A-135A.6Pressurized-Water Reactor Medium-and Large-BreakLOCAInitial ResponseAnalysis SummaryFor all LOCA scenarios, containmentfancoolersand containment spraysare available. The breakis assumedto occurinthehorizontal part of the cold leg in LoopA (pressurizer loop).Table16through Table23below provide results for this portion of the analysis.Table16SurryMBLOCA and LBLOCAResults CaseBreak Size (inch)4# HHSIPumps#Accum.# LHSIPumpsAFW?1Time of Initial Core Uncovery (hr)Core Damage During InjectionPhase? (hr) 9 2100 Yes0.42No 2150210.410.73201100.42No 2211010.42No 327110 No0.38No 2291010.38No 3 1 4101 Yes0.09No111000.09No 2120010.100.27130110.100.27140210.10No221100.09No 225101 No0.09No281100.09No 2 2 6101 Yes0.04No50010.040.1660110.04No71000.070.2881100.08No 216101 No0.04No171100.06No 2260110.04No 3 8101 Yes0.02No181100.01No 2230110.03No 4 101010.01No191100.01No 2240110.02No 10Double-ended0010.02No 1Conventionally, AFW is not needed for success for an LBLOCA; the break size is large enough to remove decay heat and the system fully depressurizes.

2Note that core damage eventually occurs (or would occur, in cases in whichthe calculation was terminated early) because ofthe inability to go to HHSI recirculation (due to the unavailability of LHSI) or,more directly,from the lack of a low-pressure injection source.Recall that the present calculations are focused only on the injection phase success criteria.

3For these cases, core damage eventually occurs because HHSI recirculation is not modeled, and the pressure is not sufficiently low prior to core damage to allow for LHSI recirculation.

42in.=5.1cm; 4in.=10.2cm; 6in.=15.2cm; 8in.=20.3cm; 10in.=25.4cm.

A-136Table17Surry MBLOCA and LBLOCA Key Timings (2-Inch Breaks)EventCase9(hr)Case15(hr)Case20(hr)Case21(hr)Case27(hr)Case29(hr)Reactor trip0.010.0030.010.010.010.01HHSI injection0.01-0.010.010.010.01RCP trip (10% void)0.280.070.280.280.180.17First actuation of containment sprays1.14-1.211.140.940.94Core uncovery (water<TAF)0.420.410.420.420.380.38LHSI injection---6.39-6.17Maximum claddingtemperature timing (max. temperature) 0.44(592K)0.73(1,477K 1)0.44(592K)0.44(592K)0.40(592K)0.40(592K)Core covered0.87N/A0.80.870.750.75 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage,2,200°F(1,204°C).Table18Surry MBLOCA and LBLOCA Key Timings (4-Inch Breaks Group1)EventCase1 (hr)Case11(hr)Case12(hr)Case13(hr)Reactor trip0.0030.0030.0030.003HHSI injection0.0030.004--RCP trip (10% void)0.040.040.040.04First actuation of containment sprays0.080.080.070.07Core uncovery (water <TAF)0.090.090.100.10LHSI injection0.29-0.330.45Maximum claddingtemperature timing (max.

temperature) 0.34(982K)0.53(1,209K)0.27(1,477K 1)0.27(1,477K 1)Core covered0.38>0.83N/AN/A 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage,2,200°F(1,204°C).Table 19Surry MBLOCA and LBLOCA Key Timings (4-InchBreaks Group 2)EventCase 14 (hr)Case 22 (hr)

Case 25 (hr)Case 28 (hr)Reactor trip0.0030.0030.0030.003HHSI injection-0.0040.0040.004RCP trip (10% void)0.040.040.040.03First actuation of containment sprays0.070.080.080.07Core uncovery (water <TAF)0.100.090.090.09LHSI injection0.73-0.30-Maximum claddingtemperature timing (max. temperature) 0.73(1,183K)0.21(807K)0.32(1,054K)0.26(721K)Core covered0.790.390.390.41 A-137Table 20Surry MBLOCA and LBLOCA Key Timings (6-InchBreaks Group 1)EventCase 2 (hr)Case 5 (hr)Case 6 (hr)Case 7 (hr)Reactor trip0.0020.0020.0020.002HHSI injection0.002--0.002RCP trip (10% void)0.020.020.020.02First actuation of containment sprays0.030.030.030.03Core uncovery (water <TAF)0.040.040.040.07LHSI injection0.130.140.18-Maximum claddingtemperature timing(maximum temperature) 0.15(774K)0.16(1,477K 1)0.16(990K)0.28(1,477K 1)Core covered0.19N/A0.20N/A 1Actual peak temperature would be higher; this value corresponds to the surrogate used in this project for core damage,2,200

°F(1,204 °C).Table 21Surry MBLOCA and LBLOCA Key Timings (6-InchBreaks Group 2)EventCase 8 (hr)Case 16 (hr)Case 17 (hr)Case 26 (hr)Reactor trip0.0020.0020.0020.002HHSI injection0.0020.0020.002-RCP trip (10% void)0.020.020.020.02First actuation of containment sprays0.030.030.030.03Core uncovery (water <TAF)0.080.040.060.04LHSI injection-0.13-0.18Maximum claddingtemperature timing(maximum temperature) 0.04(592K)0.152(775K)0.04(575K)0.13(931K)Core covered0.100.190.120.22Table22Surry MBLOCA and LBLOCA Key Timings (8-Inch Breaks)EventCase3(hr)Case18(hr)Case23(hr)Reactor trip0.0020.0020.002HHSI injection0.0020.002-RCP trip (10% void)0.0090.0090.01First actuation of containment sprays0.010.010.01Core uncovery (water <TAF)0.020.010.03LHSI injection0.07-0.08Maximum claddingtemperature timing (maximum temperature) 0.10(851K)0.40(1,085K)0.07(792K)Core covered0.140.910.11Table23Surry MBLOCA and LBLOCA Key Timings (10-Inch Breaks)EventCase4 (hr)Case19(hr)Case24(hr)Case10(hr)Reactor trip0.0010.0010.0010.001HHSI injection0.0010.001--RCP trip (10% void)0.0080.0080.0060.001First actuation of containment sprays0.0080.0080.0080.005Core uncovery (water <TAF)0.010.0080.020.022LHSI injection0.04-0.050.005Maximum claddingtemperature timing (maximum temperature) 0.08(850K)0.30(835K)0.04(640K)0.036(1,043K)Core covered0.120.870.060.053 A-138A.6.1Case1: 4-Inch Break LOCA, One HHSI, One LHSI, and No ACC 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-139 0 200 400 600 800 1000 1200050010001500200025003000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-140 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-141 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-142A.6.2Case2: 6-Inch Break LOCA, One HHSI, One LHSI, and No ACC 00.5 11.5 22.50100020003000400050006000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 180100020003000400050006000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-143 0 500 1000 1500 2000 25000100020003000400050006000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000100020003000400050006000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-144 00.20.40.6 0.8 11.21.4 1.61.8 20100020003000400050006000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 200100020003000400050006000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-145 300 500 700 900 1100 1300 1500 17000100020003000400050006000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-146A.6.3Case3: 8-InchBreakLOCA, One HHSI, One LHSI, and No ACC 00.5 11.5 22.5050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-147 0 500 1000 1500 2000 2500 3000 3500 4000 4500050010001500200025003000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-148 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-149 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-150A.6.4Case4: 10-Inch Break LOCA, One HHSI, One LHSI, and No ACC 00.5 11.5 22.5 3050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-151 0 1000 2000 3000 4000 5000 6000050010001500200025003000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-152 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-153 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-154A.6.5Case5: 6-Inch Break LOCA, No HHSI, One LHSI, and No ACC 00.20.40.6 0.8 11.21.4 1.61.8 2010002000300040005000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18010002000300040005000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-155 0 100 200 300 400 500 600 700 800 900 1000010002000300040005000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000010002000300040005000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-156 00.20.40.6 0.8 11.21.4 1.61.8 2010002000300040005000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20010002000300040005000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-157 300 800 1300 1800 2300 2800 3300010002000300040005000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-158A.6.6Case6: 6-InchBreakLOCA, No HHSI, One LHSI, and One ACC 00.20.40.6 0.8 11.21.4 1.61.8 2010002000300040005000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18010002000300040005000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-159 0 100 200 300 400 500 600 700 800 900 10000100020003000400050006000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000010002000300040005000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-160 00.20.40.6 0.8 11.21.4 1.61.8 2010002000300040005000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20010002000300040005000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-161 300 500 700 900 1100 1300 1500 1700010002000300040005000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-162A.6.7Case7: 6-InchBreakLOCA, One HHSI, No LHSI, and No ACC 00.5 11.5 22.5010002000300040005000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18010002000300040005000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-163 0 100 200 300 400 500 600 700 800 900 1000010002000300040005000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000010002000300040005000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-164 00.20.40.6 0.8 11.21.4 1.61.8 2010002000300040005000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20010002000300040005000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-165 300 800 1300 1800 2300 2800010002000300040005000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-166A.6.8Case8: 6-Inch Break LOCA, One HHSI, No LHSI, and One ACC 00.5 11.5 22.502004006008001000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 1802004006008001000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-167 0 100 200 300 400 500 600 700 800 900 100002004006008001000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 45000002004006008001000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-168 00.20.40.6 0.8 11.21.4 1.61.8 202004006008001000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 2002004006008001000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-169 300 500 700 900 1100 1300 1500 170002004006008001000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-170A.6.9Case9: 2-Inch Break LOCA, One HHSI, No LHSI, and No ACC 00.20.40.60.8 11.21.41.61.8050001000015000200002500030000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050001000015000200002500030000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-171 0 50 100 150 200 250050001000015000200002500030000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-172 00.20.40.6 0.8 11.21.4 1.61.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-173 300 800 1300 1800 2300 2800050001000015000200002500030000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-174A.6.10Case10: Double-EndedCold-LegBreak LOCA, No HHSI, One LHSI, and No ACC 00.5 11.5 22.5 33.5 44.50100200300400500600 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 180100200300400500600 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-175 0 200 400 600 800 1000 1200 1400 1600 1800 20000100200300400500600 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000100200300400500600 time [sec]

Water (gal)RWST SPRLHSI HHSI A-176 00.20.40.6 0.8 11.21.4 1.61.8 20100200300400500600 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 200100200300400500600 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-177 300 500 700 900 1100 1300 15000100200300400500600 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-178A.6.11Case11: 4-Inch Break LOCA, One HHSI, No LHSI, and No ACC 00.5 11.5 22.5050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Pressure [MPa]SG ASG BSG C PRZ A-179 0 50 100 150 200 250 300 350 400 450 500050010001500200025003000 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-180 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-181 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-182A.6.12Case12: 4-Inch Break LOCA, No HHSI, One LHSI, and No ACC 00.20.40.6 0.8 11.21.4 1.61.8 2030060090012001500 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18030060090012001500 time [sec]Reactor Pressure [MPa]

PRZCL-A A-183 0 100 200 300 400 500 600 700 800 900 1000030060090012001500 time [sec]Total Flow (kg/s)LHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000030060090012001500 time [sec]

Water (gal)RWST SPRLHSI HHSI A-184 0 2 4 6 8 10 12 14 16 18 200200400600800100012001400 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF 300 800 1300 1800 2300 2800 3300030060090012001500 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-185A.6.13Case13: 4-Inch Break LOCA, No HHSI, One LHSI, and One ACC 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Reactor Pressure [MPa]

PRZCL-A A-186 0 50 100 150 200 250 300 350 400 450 500050010001500200025003000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-187 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-188 300 800 1300 1800 2300 2800050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-189A.6.14Case14: 4-Inch Break LOCA, No HHSI, One LHSI, and Two ACC 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Reactor Pressure [MPa]

PRZCL-A A-190 0 50 100 150 200 250 300 350 400 450 500050010001500200025003000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-191 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-192 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-193A.6.15Case15: 2-Inch Break LOCA, No HHSI, One LHSI, and Two ACC 00.20.40.6 0.8 11.21.41.6050010001500200025003000 time [sec]Containment Pressure [bar]SG-ADome 0 2 4 6 8 10 12 14 16 18050010001500200025003000 time [sec]Reactor Pressure [MPa]

PRZCL-A A-194 0 50 100 150 200 250050010001500200025003000 time [sec]Total Flow (kg/s)

ACCLHSI (inj)LHSI (rec)HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPRLHSI HHSI A-195 00.20.40.6 0.8 11.21.4 1.61.8 2050010001500200025003000 time [sec]Containment Water Level [m]SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-196 300 800 1300 1800 2300 2800050010001500200025003000 time [sec]Maximum Core Temperature [K]TCLTFU 2200 F A-197A.6.16Case16: 6-Inch Break LOCA, One HHSI, One LHSI, and No ACC, without Auxiliary Feedwater 0 0.5 1 1.5 2 2.50100020003000400050006000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 180100020003000400050006000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-198 0 500 1000 1500 2000 25000100020003000400050006000 time [sec]

Total Flow (kg/s)

LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 4500000100020003000400050006000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-199 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20100020003000400050006000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 200100020003000400050006000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-200 300 500 700 900 1100 1300 1500 17000100020003000400050006000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-201A.6.17Case17: 6-Inch Break LOCA, One HHSI, No LHSI, and One ACC, without Auxiliary Feedwater 0 0.5 1 1.5 2 2.502004006008001000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1802004006008001000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-202 0 100 200 300 400 500 600 700 800 900 100002004006008001000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 45000002004006008001000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-203 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 202004006008001000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 2002004006008001000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-204 300 500 700 900 1100 1300 1500 170002004006008001000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-205A.6.18Case18: 8-InchBreakLOCA, One HHSI, No LHSI, and One ACC 0 0.5 1 1.5 2 2.505001000150020002500300035004000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1805001000150020002500300035004000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-206 0 200 400 600 800 1000 1200 1400 160005001000150020002500300035004000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 45000005001000150020002500300035004000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-207 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 2005001000150020002500300035004000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-208 300 500 700 900 1100 1300 1500 170005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-209A.6.19Case19: 10-Inch Break LOCA, One HHSI, No LHSI,and One ACC 0 0.5 1 1.5 2 2.5 305001000150020002500300035004000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 1805001000150020002500300035004000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-210 0 200 400 600 800 1000 1200 140005001000150020002500300035004000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 45000005001000150020002500300035004000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-211 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 2005001000150020002500300035004000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-212 300 500 700 900 1100 1300 1500 170005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-213A.6.20Case20: 2-Inch Break LOCA, One HHSI, No LHSI, and One ACC 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-214 0 50 100 150 200 250050001000015000200002500030000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-215 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-216 300 500 700 900 1100 1300 1500 1700 1900050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-217A.6.21Case21: 2-Inch Break LOCA, One HHSI, One LHSI, and No ACC 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Reactor Pressure [MPa]

PRZ CL-A A-218 0 50 100 150 200 250050001000015000200002500030000 time [sec]

Total Flow (kg/s)

LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-219 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-220 300 800 1300 1800 2300 2800050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-221A.6.22Case22: 4-Inch Break LOCA, One HHSI, No LHSI, and One ACC 0 0.5 1 1.5 2 2.5050010001500200025003000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050010001500200025003000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-222 0 50 100 150 200 250 300 350 400 450 500050010001500200025003000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-223 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-224 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-225A.6.23Case23: 8-Inch Break LOCA, No HHSI, One LHSI, and One ACC 0 0.5 1 1.5 2 2.5050010001500200025003000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050010001500200025003000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-226 0 50 100 150 200 250 300 350 400 450 500050010001500200025003000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-227 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-228 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-229A.6.24Case24: 10-Inch Break LOCA, No HHSI, One LHSI, and One ACC 0 0.5 1 1.5 2 2.5 3050010001500200025003000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050010001500200025003000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-230 0 1000 2000 3000 4000 5000 6000050010001500200025003000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-231 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-232 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-233A.6.25Case25: 4-Inch Break LOCA, One HHSI, One LHSI, No ACC, without Auxiliary Feedwater 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050010001500200025003000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-234 0 200 400 600 800 1000 1200050010001500200025003000 time [sec]

Total Flow (kg/s)

LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-235 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-236 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-237A.6.26Case26: 6-Inch Break LOCA, No HHSI, One LHSI, and One ACC, without Auxiliary Feedwater 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2010002000300040005000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18010002000300040005000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-238 0 100 200 300 400 500 600 700 800 900 1000010002000300040005000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000010002000300040005000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-239 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2010002000300040005000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20010002000300040005000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-240 300 500 700 900 1100 1300 1500 1700010002000300040005000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-241A.6.27Case27: 2-Inch Break LOCA, One HHSI, No LHSI, and One ACC, without Auxiliary Feedwater 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-242 0 50 100 150 200 250 300050001000015000200002500030000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-243 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-244 300 500 700 900 1100 1300 1500 1700 1900 2100 2300050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-245A.6.28Case28: 4-InchBreakLOCA, One HHSI, No LHSI, and One ACC, without Auxiliary Feedwater 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]

Containment Pressure [bar]

SG-A Dome 0 2 4

6 8 10 12 14 16 18050010001500200025003000 time [sec]

Pressure [MPa]

SG A SG B SG C PRZ A-246 0 100 200 300 400 500 600050010001500200025003000 time [sec]

Total Flow (kg/s)

ACC LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050010001500200025003000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-247 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050010001500200025003000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050010001500200025003000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-248 300 500 700 900 1100 1300 1500 1700050010001500200025003000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F A-249A.6.29Case29: 2-Inch Break LOCA, One HHSI, One LHSI, and No ACC, without Auxiliary Feedwater 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]

Reactor Pressure [MPa]

PRZ CL-A 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8050001000015000200002500030000 time [sec]

Containment Pressure [bar]

SG-A Dome A-250 0 50 100 150 200 250 300050001000015000200002500030000 time [sec]

Total Flow (kg/s)

LHSI (inj)

LHSI (rec)

HHSI (inj)

BRK 0 50000 100000 150000 200000 250000 300000 350000 400000 450000050001000015000200002500030000 time [sec]

Water (gal)RWST SPR LHSI HHSI A-251 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2050001000015000200002500030000 time [sec]Containment Water Level [m]

SumpCavity 0 2 4 6 8 10 12 14 16 18 20050001000015000200002500030000 time [sec]Reactor Water Level (m)

PRZ DC Ring 1 BAF TAF A-252 300 800 1300 1800 2300 2800050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TCL TFU 2200 F

A-254A.7REFERENCES(NRC, 2003)U.S.Nuclear Regulatory Commission, "Virginia Electric and Power Company Docket No. 50-280 Surry Power Station, Unit No. 1 Renewed Facility Operating License,"Appendix A,"Technical Specifications," March 2003. (Agencywide Documents Access and Management System (ADAMS) Accession No.ML0529103580)

APPENDIXBPEACH BOTTOMMELCOR ANALYSES

B-1B.1 Inadvertent Open Relief Valve Success CriteriaAnalysis SummaryTable1andTable2below provide results for this portion of the analysis.Table1Peach Bottom Inadvertent Open SRV ResultsCaseRCICHPCICRDLPCILPCSac/dc FW, SPC, ADS Core Uncovery (hr)Core Damage (hr)1YesNo NoYesNoac/dc NoNoNo 2 NoYesNoNo 3 No1att=0 and 2att=10min0.41No 41att=0 and 2at20minafter SCRAM0.37No 4a 10.29No 4b 1FWNoNo5NoNo0.32No 1For this case, the reactor was allowed to scram based on a reactor protection system trip signal, rather than at time t=0.Table2Peach Bottom Inadvertent Open SRV Key Timings (Cases1-5)EventCase1(hr)Case2(hr)Case3(hr)Case4(hr)Case4a(hr)Case4b (hr)Case5(hr)SRV1open0000000Reactor trip0000< 0.01 10.760MSIVs close< 0.01< 0.01< 0.01< 0.01< 0.010.790Downcomer level first reaches

L20.070.070.070.070.03N/A0.07RCIC/HPCI first started (CST injection mode)0.080.08-----

2 n dCRD pump started--0.170.330.331.09-Downcomer level reaches L10.378.930.320.320.24N/A0.26Downcomer level below TAF0.378.930.350.330.25N/A0.28Suppression pool temp. >110°F 30.400.610.420.420.410.300.40LPCI first started0.518.930.590.580.53N/A0.57RCIC/HPCI pump isolation:low steamline pressure <0.52MPa(75psig)0.825.59-----HCTL limit reached 3(no action taken)4.54.0> 1 2> 1 25.00.57> 1 2RHR pump isolation -NPSH9.611.1> 1 2> 1 2> 10 25.4> 1 2Maximum cladding temperature timing (max.temperature)

No heatup No heatup 0.78 (786K)0.76(830K)0.67(941K)No heatup 0.75 (1,212K)1Reactor trips at 8seconds on low RPV level.

2The simulation was stopped beforereaching this condition.

3The HCTL limit is based on suppression pool temperature, suppression pool level, and RPV pressure.

B-2B.1.1Case1: Safety Relief Valve Opening and Reactor Core Isolation Cooling 0 1 2 3 4 5 6 7 805000100001500020000 time [sec]Reactor Pressure [MPa]Dome DC 0 10 20 30 40 50 60 70 80 90 10005000100001500020000 time [sec]Mass Flow Rate [kg/s]SV->DWSRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9SRV10SRV11Only SRV1(ADS) Open B-3 0 2 4 6 8 10 12 14 1605000100001500020000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 5 10 15 20 25 30 35 4005000100001500020000 time [sec]Total Mass Flow Rate [kg/s]RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-4 0 100 200 300 400 500 600 700 800 90005000100001500020000 time [sec]Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 200005000100001500020000 time [sec]Maximum Core Temperature [K]TFUTCL 2200F B-5 00.20.40.6 0.8 11.21.4 1.61.8 205000100001500020000 time [sec]Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 410 43005000100001500020000 time [sec]Wetwell Temperature [K]TLIQ TSAT TVAP B-6 44.14.24.3 4.4 4.54.64.7 4.84.9 505000100001500020000 time [sec]WW Water Level [m]SwollenCollapsed B-7B.1.2Case2: Safety Relief ValveOpening and High-Pressure CoolantInjection 0 1 2 3

4 5 6 7

80500010000150002000025000300003500040000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-8 0 2 4 6 8 10 12 14 160500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 50 100 150 200 250 300 3500500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-9 0 100 200 300 400 500 600 700 800 9000500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 20000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-10 0 0.5 1 1.5 2 2.5 30500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 410 430 4500500010000150002000025000300003500040000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-11 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 60500010000150002000025000300003500040000 time [sec]WW Water Level [m]Swollen Collapsed B-12B.1.3Case3: Safety Relief ValveOpening and CRD2 at 10Minutes 0 1 2 3

4 5 6 7

805001000150020002500300035004000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 10005001000150020002500300035004000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-13 0 2 4

6 8 10 12 14 16 1805001000150020002500300035004000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 2 4 6 8 10 12 14 16 18 2005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRDCRD Pump 2 Start B-14 0 100 200 300 400 500 600 700 800 90005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 200005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-15 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 39005001000150020002500300035004000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-16 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 505001000150020002500300035004000 time [sec]WW Water Level [m]Swollen Collapsed B-17B.1.4Case4: Safety Relief ValveOpening and CRD2 at 20Minutes 0 1 2 3

4 5 6 7

805001000150020002500300035004000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 10005001000150020002500300035004000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-18 0 2 4 6 8 10 12 14 1605001000150020002500300035004000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 2 4 6 8 10 12 14 16 18 2005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRDCRD Pump 2 Start B-19 0 100 200 300 400 500 600 700 800 90005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 200005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-20 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 39005001000150020002500300035004000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-21 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 505001000150020002500300035004000 time [sec]WW Water Level [m]Swollen Collapsed B-22B.1.4.1Case4a: Safety Relief Valve Opening, CRD2 at 20Minutes, and No Reactor Trip at t=0 0 1 2 3

4 5

6 7

805001000150020002500300035004000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 10005001000150020002500300035004000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-23 0 2 4

6 8 10 12 14 16 1805001000150020002500300035004000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 2 4 6 8 10 12 14 16 18 2005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRDCRD Pump 2 Start B-24 0 100 200 300 400 500 600 700 800 90005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 200005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-25 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 41005001000150020002500300035004000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-26 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 505001000150020002500300035004000 time [sec]WW Water Level [m]Swollen Collapsed B-27B.1.4.2Case4b:Safety Relief Valve Opening, CRD2 at 20Minutes, and No Reactor or Feedwater Trip at t=0 0 1 2

3 4

5 6

7 80500010000150002000025000300003500040000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 1000500010000150002000025000300003500040000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-28 0 2 4

6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 2 4 6 8 10 12 14 16 18 200500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRDCRD Pump 2 Start B-29 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 10500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 20000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-30 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 50500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 410 4300500010000150002000025000300003500040000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-31 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 60500010000150002000025000300003500040000 time [sec]WW Water Level [m]Swollen Collapsed 0 500 1000 1500 2000 25000500010000150002000025000300003500040000 time [sec]

FW Mass Flow Rate [kg/s]

B-32B.1.5Case5: Safety Relief ValveOpening 0 1 2

3 4

5 6

7 805001000150020002500300035004000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 10 20 30 40 50 60 70 80 90 10005001000150020002500300035004000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-33 0 2 4 6 8 10 12 14 1605001000150020002500300035004000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 0.5 1 1.5 2 2.5 3 3.5 4 4.505001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-34 0 100 200 300 400 500 600 700 800 90005001000150020002500300035004000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 1800 200005001000150020002500300035004000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-35 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 205001000150020002500300035004000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 39005001000150020002500300035004000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-36 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 505001000150020002500300035004000 time [sec]WW Water Level [m]Swollen Collapsed B-37B.2Boiling-Water Reactor Station BlackoutAnalysis SummaryTable3throughTable6below provide results forthis portion of the analysis.Table3Peach Bottom Station Blackout ResultsCaseRCICHPCIac/dcSRV Sticks Open?HCTL Depress?Core Uncovery(hr)Core Damage (hr)1 No No-No 1 No0.51.2 1aacrecovery at 1.2hrNo0.51.2 22-At t=00.30.8 3 Yesdcis always available No17.719.44Yes6.07.252hr of dc No3.34.3 6dcis always availableAt 187lifts6.07.2 7NoYes No17.519.38Yes9.310.892hr of dc No3.84.9 10dcis always availableAt 187lifts9.210.7 1For this case, the SRV does not stick open until after core damage, so this assumption does not affect the outcome.

2Recovery of injection upon reaching 2,200

°F (1,204°C) quickly arrests further heatup.Table4Peach Bottom Station Blackout Key Timings (Cases1, 1a, and 2)EventCase1(hr)Case1a(hr)Case2(hr)Reactor trip, MSIVclosure000Downcomer level reaches L20.160.160.16Downcomer level reaches L10.500.500.27Downcomer level below TAF0.500.500.27Gap release: 900°C (1,652°F)1.021.020.69Core damage: max.temp.>1,204°C (2,200°F)1.171.170.79HPCI, RCIC, CRD injection start-1.17-ADS actuated-1.24-Downcomerlevel recovers above TAF-1.27-SRV sticks open due to high # of cycles1.75--

B-38Table5Peach Bottom Station Blackout Key Timings (Cases3-6)EventCase3 (hr)Case4 (hr)Case5 (hr)Case6 (hr)Reactor trip, MSIV closure0000Downcomer level first reaches L20.160.160.160.16RCIC started (CST injection mode)0.170.170.170.17RCIC fails due to loss of dc--2.00-HCTL limit reached2.46 (no action taken) 2.462.46 (no action taken)2.46 (no action taken)SRVsticks open due to high # of cycles---2.47RCIC NPSH limit exceeded 112.67---RCIC pump isolation: low steamline pressure <0.52MPa (75psig)-3.90-3.92RCIC injection ends due to CST level <5ft (1.5m)14.43---Downcomerlevel reaches L117.685.613.255.62Downcomer level below TAF17.685.613.255.62Gap release: 900°C (1,652°F)19.066.994.047.00Core damage max.temp.>1,204°C (2,200°F)19.427.174.257.18Exhaust pressure exceeded:0.35MPa (50psig)20.14---1Switchover to the suppression pool is not permitted after this point.Table6Peach Bottom Station Blackout Key Timings (Cases7-10)EventCase7 (hr)Case8 (hr)Case9 (hr)Case10 (hr)Reactor trip, MSIV closure0000Downcomer level first reaches L20.160.160.160.16HPCI started (CST injection mode)0.170.170.170.17HPCI fails due to loss of dc--2.00-SRV sticks open due to high # of cycles---2.53HCTL limit reached2.67 (no action taken) 2.672.67 (no action taken)2.67 (no action taken)HPCI NPSH limit exceeded 112.07---HPCI pump isolation: low steamline pressure <0.52MPa (75psig)-5.72-5.61HPCI injection ends due to CST level <5ft (1.5m)16.05---Downcomerlevel reaches L117.538.973.828.94Downcomer level below TAF17.539.063.828.94Gap release: 900°C (1,652°F)18.9610.594.6310.46Core damage max.temp.>1,204°C (2,200°F)19.3110.84.8510.68Exhaust pressure exceeded: 1.04MPa (150psig)----1Switchover to the suppression pool is not permitted after this point.

B-39B.2.1Case1:Station BlackoutwithNo Injection 0 1 2

3 4 5 6 7

8 901000200030004000500060007000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4 6

8 10 12 14 1601000200030004000500060007000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-40 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 101000200030004000500060007000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 101000200030004000500060007000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-41 0 500 1000 1500 2000 2500 300001000200030004000500060007000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 501000200030004000500060007000 time [sec]

Containment Pressure [bar]

DW WW B-42 250 270 290 310 330 350 370 390 41001000200030004000500060007000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 501000200030004000500060007000 time [sec]WW Water Level [m]Swollen Collapsed B-43B.2.1.1Case1a: Station BlackoutwithNo Injection and Power Recovery at Core Damage 0 1 2

3 4 5 6 7

8 9010002000300040005000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4

6 8 10 12 14 16 18010002000300040005000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-44 0 50 100 150 200 250 300 350010002000300040005000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 200 400 600 800 1000 1200010002000300040005000 time [sec]

Total Mass Flow Rate [kg/s]

LPCI(I)LPCI(II)SPC(I)SPC(II)

B-45 0 200 400 600 800 1000 1200 1400 1600010002000300040005000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2010002000300040005000 time [sec]

Containment Pressure [bar]

DW WW B-46 250 270 290 310 330 350 370 390010002000300040005000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 5010002000300040005000 time [sec]WW Water Level [m]Swollen Collapsed B-47B.2.2Case2: Station Blackoutand Safety Relief ValveOpen at t=0 0 1 2

3 4 5 6 7

8 901000200030004000500060007000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 20 40 60 80 100 12001000200030004000500060007000 time [sec]

Mass Flow Rate [kg/s]

SV->DW SRV1(ADS)SRV2(ADS)SRV3(ADS)SRV4(ADS)SRV5(ADS)SRV6 SRV7 SRV8 SRV9 SRV10 SRV11Only SRV1(ADS) Open B-48 0 2 4 6 8 10 12 14 1601000200030004000500060007000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 101000200030004000500060007000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-49 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 101000200030004000500060007000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 500 1000 1500 2000 2500 300001000200030004000500060007000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-50 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 501000200030004000500060007000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 410 43005000100001500020000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-51 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 501000200030004000500060007000 time [sec]WW Water Level [m]Swollen Collapsed B-52B.2.3Case3: Station Blackoutand Reactor Core Isolation CoolingNote:By the time reactor core isolation cooling (RCIC)injection stops from condensate storage tank (CST)depletion at 14.4hours, the RCIC pump net positive suction head (NPSH)limit has already been exceededat 11.6hours.

0 1 2 3 4 5 6 7 8 9010000200003000040000500006000070000 time [sec]Reactor Pressure [MPa]Dome DC B-53 0 2 4

6 8 10 12 14 16 18010000200003000040000500006000070000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 5 10 15 20 25 30 35 40010000200003000040000500006000070000 time [sec]Total Mass Flow Rate [kg/s]RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-54 00.10.20.3 0.4 0.50.60.7 0.80.9 1010000200003000040000500006000070000 time [sec]Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 500 1000 1500 2000 2500 3000010000200003000040000500006000070000 time [sec]Maximum Core Temperature [K]TFUTCL 2200F B-55 00.5 11.5 22.5 33.5 4010000200003000040000500006000070000 time [sec]Containment Pressure [bar]DW(200)DW(201)DW(202)DW(205)DC(210)WW(220)250 270 290 310 330 350 370 390 410 430 450010000200003000040000500006000070000 time [sec]Wetwell Temperature [K]TLIQ TSAT TVAP B-56 44.24.44.6 4.8 55.25.4 5.65.8 6010000200003000040000500006000070000 time [sec]WW Water Level [m]SwollenCollapsed B-57B.2.4Case 4: Station Blackoutand RCICand HCTLDepressurization 0 1 2

3 4 5 6 7

8 9050001000015000200002500030000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 20 40 60 80 100 120050001000015000200002500030000 time [sec]

HCTL Depressurizat ion Flow Rate [kg/s]

B-58 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 5 10 15 20 25 30 35 40050001000015000200002500030000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-59 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1050001000015000200002500030000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 500 1000 1500 2000 2500 3000050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-60 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5050001000015000200002500030000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 410 430050001000015000200002500030000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-61 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 6050001000015000200002500030000 time [sec]WW Water Level [m]Swollen Collapsed B-62B.2.5Case5: Station Blackoutand RCICand 2-Hour Direct Current 0 1 2

3 4 5 6 7

8 90200040006000800010000120001400016000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4

6 8 10 12 14 16 180200040006000800010000120001400016000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-63 0 5 10 15 20 25 30 35 400200040006000800010000120001400016000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 10200040006000800010000120001400016000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-64 0 500 1000 1500 2000 2500 30000200040006000800010000120001400016000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 20200040006000800010000120001400016000 time [sec]

Containment Pressure [bar]

DW WW B-65 250 270 290 310 330 350 370 390 410 4300200040006000800010000120001400016000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 50200040006000800010000120001400016000 time [sec]WW Water Level [m]Swollen Collapsed B-66B.2.6Case6: Station Blackoutand RCICand SRVStuck Open 0 1 2

3 4 5 6 7

8 9050001000015000200002500030000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-67 0 5 10 15 20 25 30 35 40050001000015000200002500030000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1050001000015000200002500030000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-68 0 500 1000 1500 2000 2500 3000050001000015000200002500030000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.5 1 1.5 2 2.5 3 3.5 4050001000015000200002500030000 time [sec]

Containment Pressure [bar]

DW WW B-69 250 270 290 310 330 350 370 390 410 430 450050001000015000200002500030000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 6050001000015000200002500030000 time [sec]WW Water Level [m]Swollen Collapsed B-70B.2.7Case7: Station Blackoutand High-Pressure CoolantInjectionNote: By the time high-pressure coolantinjection (HPCI)injection stops from the CST at 16.05hours, the HPCIpump NPSHlimithas already been exceeded at 12.07hours.

0 1 2 3 4 5 6 7 8 9010000200003000040000500006000070000 time [sec]Reactor Pressure [MPa]Dome DC 0 2 4

6 8 10 12 14 16 18010000200003000040000500006000070000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-71 0 50 100 150 200 250 300 350010000200003000040000500006000070000 time [sec]Total Mass Flow Rate [kg/s]RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 00.10.20.3 0.4 0.50.60.7 0.80.9 1010000200003000040000500006000070000 time [sec]Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-72 0 200 400 600 800 1000 1200 1400 1600 1800 2000010000200003000040000500006000070000 time [sec]Maximum Core Temperature [K]TFUTCL 2200F 00.5 11.5 22.5 33.5 4010000200003000040000500006000070000 time [sec]Containment Pressure [bar]DW(200)DW(201)DW(202)DW(205)DC(210)WW(220)

B-73 250 270 290 310 330 350 370 390 410 430 450010000200003000040000500006000070000 time [sec]Wetwell Temperature [K]TLIQ TSAT TVAP 44.24.44.6 4.8 55.25.4 5.65.8 6010000200003000040000500006000070000 time [sec]WW Water Level [m]SwollenCollapsed B-74B.2.8Case8: Station Blackoutand HPCI and HCTLDepressurization 0 1 2

3 4 5 6 7

8 90500010000150002000025000300003500040000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 20 40 60 80 100 1200500010000150002000025000300003500040000 time [sec]

HCTL Depressurizat ion Flow Rate [kg/s]

B-75 0 2 4

6 8 10 12 14 16 180500010000150002000025000300003500040000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF 0 50 100 150 200 250 300 3500500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD B-76 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 10500010000150002000025000300003500040000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)0 200 400 600 800 1000 1200 1400 1600 18000500010000150002000025000300003500040000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F B-77 0 0.5 1 1.5 2 2.50500010000150002000025000300003500040000 time [sec]

Containment Pressure [bar]

DW WW 250 270 290 310 330 350 370 390 4100500010000150002000025000300003500040000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP B-78 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 60500010000150002000025000300003500040000 time [sec]WW Water Level [m]Swollen Collapsed B-79B.2.9Case9: Station Blackoutand HPCIand 2-Hour Direct Current 0 1 2

3 4 5 6 7

8 905000100001500020000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4

6 8 10 12 14 16 1805000100001500020000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-80 0 50 100 150 200 250 300 35005000100001500020000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 105000100001500020000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-81 0 500 1000 1500 2000 2500 300005000100001500020000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.5 1 1.5 2 2.5 305000100001500020000 time [sec]

Containment Pressure [bar]

DW WW B-82 250 270 290 310 330 350 370 390 410 43005000100001500020000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.1 4.2 4.3 4.4 4.5 4.6 4.7 4.8 4.9 505000100001500020000 time [sec]WW Water Level [m]Swollen Collapsed B-83B.2.10Case10:Station Blackoutand HPCIand SRV Stuck Open 0 1 2

3 4 5 6 7

8 9050001000015000200002500030000350004000045000 time [sec]

Reactor Pressure [MPa]

Dome DC 0 2 4

6 8 10 12 14 16 18050001000015000200002500030000350004000045000 time [sec]Reactor Water Level (m)

JP DC Core (R1)BAF TAF B-84 0 50 100 150 200 250 300 350050001000015000200002500030000350004000045000 time [sec]

Total Mass Flow Rate [kg/s]

RCIC(SP)RCIC(CST)HPCI(SP)HPCI(CST)LPCS CRD 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1050001000015000200002500030000350004000045000 time [sec]

Total Mass Flow Rate [kg/s]LPCI(I)LPCI(II)SPC(I)SPC(II)

B-85 0 500 1000 1500 2000 2500 3000050001000015000200002500030000350004000045000 time [sec]

Maximum Core Temperature [K]

TFU TCL 2200F 0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5050001000015000200002500030000350004000045000 time [sec]

Containment Pressure [bar]

DW WW B-86 250 270 290 310 330 350 370 390 410 430 450050001000015000200002500030000350004000045000 time [sec]

Wetwell Temperature [K]TLIQ TSAT TVAP 4 4.2 4.4 4.6 4.8 5 5.2 5.4 5.6 5.8 6050001000015000200002500030000350004000045000 time [sec]WW Water Level [m]Swollen Collapsed APPENDIXCEVENTTREEMODELSFORSURRYANDPEACHBOTTOM

C-1C.1Surry Event TreesThis section provides the relevant event trees from the Surry (v3.52) Standardized Plant Analysis Risk model dated November 2009. These event trees show the sequences described in the main report.

C-2 C-3 C-4 C-5 C-6 C-7 C-8 C-9

C-11C.2Peach Bottom Event TreesThis section provides the relevant event trees from the Peach Bottom (v3.50) Standardized Plant Analysis Risk model dated October 2009. These event trees show the sequences described in the main report.

C-12 C-13 C-14 C-15

APPENDIXDRESPONSE TO PUBLIC COMMENTS

D-1D.1IntroductionIn November 2010,the U.S. Nuclear Regulatory Commission (NRC)issued a draft of NUREG-1953for a 30-day public comment period. In December 2010,the NRC extended the comment period to February 28, 2011,in response to an external request. Exelon Nuclear (Exelon, 2010)and the Nuclear Energy Institute (NEI, 2011) submitted comments to the agency.The following sections respond to each comment received, and NUREG-1953has been modifiedas necessary to address some concerns. The comments were helpful not only for improving the utility of this report, but also for providing insights for future analyses.

D-3D.2Comments from Exelon Nuclear Dated December 15, 2010D.2.1Plant Representation in MELCORSection 5.1 of the draft NUREG states that the core nodalization assumed 10 axial by 5 radial regions. Further clarification on this investigation to the sensitivity of this assumed nodalization scheme would help demonstrate the impact of this assumption.

NRC ResponseThe nodalization used in the Surry model follows a well-established nodalization convention for the use of MELCOR in reactor applications.Past sensitivity studies have shown this nodalization to reproduce the necessary physics for the types of accidents being considered in this report.D.2.2Stuck Open Safety Relief ValvesThe State-of-the-Art Reactor Consequence Analyses (SOARCA) project identified a significant sensitivity to Safety Relief Valves (SRVs) sticking open due to elevated gas temperatures. Further clarification on how this impactsthe current success criteria analysis would be beneficial.

NRC ResponseRegarding the issue of relief valvessticking open due to elevated gas temperatures, the elevated temperaturesat the valve necessary to prompt this concern were not seen until the time of core damage. Since the present study only considers the phase of the accident up tothe start of core damage, this valve failure mechanism is not believed to be relevant here. Section 5.1 of the main report nowaddressesthisissue.D.2.3Reactor Coolant Pump Manual TripIt appears that a credit was not assumed for operator actions to trip the Reactor Coolant Pumps (RCPs) manually. It is recommended that a sensitivity be included to demonstrate the impact of manual actions to trip the RCPs in accordance with the existing guidance.

NRC ResponseThe staff agrees with the need to assess this impact. To that end, the staff revisited this issue for each of the initiators considered for Surry and developed the following simplified criteria:For small-break loss-of-coolant accidents (SBLOCAs), manual RCP tripis covered in E-0, "Reactor Trip or Safety Injection"and E-1, Loss of Reactor or Secondary Cooling".For the Surry procedures used in this study,the RCPs will be tripped whenreactor coolant systemsubcooling reaches 30degrees Fahrenheit (F)(16.7degrees Celsius (C))if at least onecharging pump is running.Thesituationfor depressurization and cooldown cases would be more complicated.Two cases from the main report were re-run using the above criteria to assess the effect. Table 1andTable 2provide the results of these calculations.

D-4E-0, "Reactor Trip or Safety Injection"and FR-H.1, "Response to Loss of Secondary Heat Sink"cover manual RCP trip for loss-of-all feedwater events.We assume that the RCPs will be tripped between 5 to15 minutes following reactor trip, based on data from a recent Halden study that investigated crew response to this initiator as presented in (Coyne 2009).For simplicity, 10 minutes is used for the trip criteria. One of the cases from the main report was re-run, and results are provided in Table 3.Manual RCP trip for steam generator tube rupture (SGTR)is covered in E-0, "Reactor Trip or Safety Injection"andearly in E-3, "Steam Generator Tube Rupture".

1 Weassume that the RCP trip criteria would not be reached during the E-0 and early E-3execution, and it is not a continuous action in E-3.Based on this,we donot assumethat operators trip the RCPs, even as a sensitivity.Manual RCP trip for station blackoutis not relevant sincethe RCPs require alternating current (ac)power.E-0, "Reactor Trip or Safety Injection"and E-1, "Loss of Reactor or Secondary Coolant"cover manual RCP trip for medium-break loss-of-coolant accidents(MBLOCAs)andlarge-break loss-of-coolant accidents(LBLOCAs). TheRCPs will be tripped when subcoolingreaches 30 degrees F (16.7 degrees C)if at least onecharging pump is running.However, for these cases, the 10-percentvoid criteria assumed in the MELCOR model isreached early (latest case is approximately17 minutes).For the purposes of these calculations, it is judged that additional sensitivities are not required.As cited above, the following sensitivity studies were completed to demonstrate the impact of manual actions to trip the RCPs in accordance with the existing guidance.The impact on the time to key events, including the time to core damage, was very small.Table 1SBLOCA Case 2 withRCP Trip at 30

°F(16.7°C)SubcoolingEventTime (hr)22cReactor trip0.030.03HHSI injection0.030.03First actuation of containment sprays(containment pressure >1.72 bars)2.652.65RWST depletion (<13.5%)4.304.30Spray recirculation4.304.30Accumulator start to inject4.524.52RCP trip (30

°F(16.7°C)subcooled)-5.03RCP trip (10-percentvoid)5.76-Core uncovery7.327.40Core damage (max.temp.>2,200

°F)9.9310.1 11.72 bars = 0.172 MPa;2,200

°F=1,204°C.

1Note that the Surry E-3procedurespecifically directs operatorsto disregard RCP trip criteria once depressurization and cooldown has been initiated.

D-5Table 2SBLOCA Case 8 withRCP Trip at 30

°F(16.7°C)SubcoolingEventTime (hr)88aReactor trip0.010.01HHSI injection0.010.01First actuation of containment sprays(containment pressure >1.72 bars)3.233.23RWST depletion (<13.5%)5.525.52Spray recirculation5.535.53Accumulator start to inject5.655.65RCP trip (30

°F(16.7°C)subcooled)-6.35RCP trip (10% void)10.3-Core uncovery14.414.3Core damage (max.temp.>2,200

°F)21.421.4 11.72 bars = 0.172 MPa;2,200

°F=1,204°C.Table 3LOMFW Case 2 withRCP Trip at 10 MinutesEventTime (hr)22aMFW, MD-AFW, TD-AFW unavailable00Reactor trip0.008(29 sec)0.008(29 sec)MCP trip (10 min)-0.17SG dryout0.630.65PRT rupture disk open0.970.96SI signal (containment pressure >1.22bars)1.361.31RCP trip (10% void)1.43-First actuation of containment sprays(containment pressure >1.72 bars)3.243.10RWST depletion (<13.5%)8.358.10Core uncovery1.65/9.54 29.26Core damage (max.temp.>2,200

°F)11.8011.47 11.22bars=0.122MPa;1.72 bars = 0.172 MPa;2,200

°F=1,204°C.2For Case 2, the core uncovers early in the accident,recovers prior to significant heatup, and later uncovers again (due to the loss of HHSI).D.2.4General CommentsThe draft NUREG summarizes best estimate analyses for Surry and Peach Bottom success criteria. The NUREG provides adequate details to describe the sequences being investigated and provides a clear summary of the results. In addition, the results are summarized in terms of the proposed changes to the currentSPAR model assumptions.A detailed comparison of these results with those from the Modular Accident Analysis Program (MAPP4) code is currently underway as an Electric Power Research Institute (EPRI) sponsored project. That comparison effort may reveal additional insights, the results of which are expected to be communicated to the NRC when they are completed.

D-6 NRC ResponseWe acknowledge this activity and will consider the results when they are made available.

D-7D.3Comments from the Nuclear Energy Institute Dated February 23, 2011D.3.1Description of Major Plant CharacteristicsIn the description of the major plant characteristics in Section 4.1, it is suggested that for Surry, it be noted that successful sump recirculation function requires containment heat removal through the recirculation spray system.

NRC ResponseAnote to this effect has been added in Section 4.1of the main report.D.3.2Plant Representation in MELCORIn describing the plant representation used for the study, it is stated that the core nodalization assumed 10 axial and 5 radial regions. Clarification of the sensitivity of this nodalization assumption would be helpful in illustrating its impact.NRCResponseExelon Nuclear also submitted this comment. Please see the NRC's response in SectionD.2.1.D.3.3Small-Break Loss-of-Coolant AccidentCase AssumptionsIt appears that accumulator injection was credited for all SLOCA cases discussed in this report; however, probabilistic risk assessments normally do not credit accumulator injection for SLOCA mitigation. The impact of this should be explored before issuance of the final NUREG.

NRC ResponseAn examination of the various cases run for SBLOCA determined that the crediting of the accumulators was not expected to affect whether the simulation went to core damage or to a stable end state. However, in some cases the assumption may have affected other aspects of the results. Specifically, in cases 1, 2, 5, 6, and 6a, the accumulators injectedafter refueling water storage tank (RWST)depletion and,therefore,may have affected the time betweenRWST depletion andcore damage. In cases 2b, 3, 4, and 6b,the accumulators injected before RWST depletion and,therefore,may have had an effect on the time to RWST depletion. Crediting the accumulators is notexpected to have made a significant difference in even the intermediate results for cases 2a, 7, and 8. Since no changes were made to the SPAR models based on the SBLOCA results (see Section 7 of the main report), sensitivity calculations were not performed to assess the effect of this assumption.However, cautionary statements now appearin multiple places in the main report to highlight the potential effect the accumulators might have on some of the intermediate results.D.3.4Additional Sensitivities to ConsiderWhile the work described in the draft NUREG involved extensive analysis evaluating sensitivities, the industry suggests two other sensitivities to consider. The first is the impact of D-8crediting manual actions to trip the Reactor Coolant Pumps in accordance with existing guidance, as such credit was not assumed in the analysis. The second suggested sensitivity that the industry suggests evaluating is the impact of the Safety Relief Valves at Peach Bottom sticking open due to elevated gas temperatures, as the State-of-the-Art Reactor Consequence Analysisidentified this as a significant sensitivity.

NRC ResponseExelon Nuclear also submitted thesecomments. Please see Sections D.2.2and D.2.3for the NRC's response.

D-9D.4REFERENCES(Coyne, 2009)Coyne, Kevin A.,"A Predictive Model of Nuclear Power Plant Crew Decision-Making and Performance in a Dynamic Simulation Environment,"pg. 333-334, Doctoral dissertation, University of Maryland, 2009.(Exelon, 2010)Exelon Generation Company, LLC, "Comments Concerning Draft NUREG-1953," December 15, 2010.(Agencywide Documents Access and Management System (ADAMS)Accession No. ML103510341)(NEI,2011)Nuclear Energy Institute, "Industry Comments on Draft NUREG-1953," February 23, 2011.(ADAMS Accession No. ML110680444)

UNITED STATES NUCLEAR REGULATORY COMMISSIONWASHINGTON, DC 20555-0001

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OFFICIAL BUSINESS NUREG-1953 Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models - Surry and Peach Bottom September 2011