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{{#Wiki_filter:ANF-87-126REVIStON1AD~MHCSDo HUCIt.EAR FUSMCORPORATION SUSQUEHANNA UNIT2CYCLE3RELOADANALYSISDESIGNANDSAFETYANALYSES.
{{#Wiki_filter:ANF-87-1 26 REVIStON 1 AD~MHCSDo HUCIt.EAR FUSM CORPORATION SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS DESIGN AND SAFETY ANALYSES.NOVEMBER 198787i2310i58 87i223 POR ADOCK 0500058]~AN AFFII.IATE OF KRAF TWERK UNION Q~KRU ADVANCED NUCLEAR FUELS CORPORATION ANF-87-126 Revision 1 Issue Date: 11/25/87 SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Prepared By: J.A.White BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services AII AFFILIATE OF KRAFTWERK UNION Qxsvu CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued.Accordingly, except as otherwise expressly pro-vided In such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained In this document, or that the use of any information, apparatus, method or process disclosed ln this document will not infringe privately owned rights: or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.The information contained herein is for the sole use of Customer.In order to avoid Impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term)of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (6)months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement.
NOVEMBER198787i2310i58 87i223PORADOCK0500058]~ANAFFII.IATE OFKRAFTWERKUNIONQ~KRU ADVANCEDNUCLEARFUELSCORPORATION ANF-87-126 Revision1IssueDate:11/25/87SUSQUEHANNA UNIT2CYCLE3RELOADANALYSISDesignandSafetyAnalysesPreparedBy:J.A.WhiteBWRSafetyAnalysisLicensing andSafetyEngineering FuelEngineering andTechnical ServicesAIIAFFILIATE OFKRAFTWERK UNIONQxsvu CUSTOMERDISCLAIMER IMPORTANT NOTICEREGARDING CONTENTSANDUSEOFTHISDOCUMENTPLEASEREADCAREFULLY AdvancedNuclearFuelsCorporation's warranties andrepresentations con-cemingthesubjectmatterofthisdocumentarethosesetforthintheAgreement betweenAdvancedNuclearFuelsCorporation andtheCustomerpursuanttowhichthisdocumentisissued.Accordingly, exceptasotherwise expressly pro-videdInsuchAgreement, neitherAdvancedNuclearFuelsCorporation noranypersonactingonitsbehalfmakesanywarrantyorrepresentation, expressed orimplied,withrespecttotheaccuracy, completeness, orusefulness oftheinfor-mationcontained Inthisdocument, orthattheuseofanyinformation, apparatus, methodorprocessdisclosed lnthisdocumentwillnotinfringeprivately ownedrights:orassumesanyliabilities withrespecttotheuseofanyinformation, ap-paratus,methodorprocessdisclosed inthisdocument.
No rights or licenses In or to any patents are implied by the furnishing of this docu-ment.XN NF F00.765 (1 ANF-87-126 Revision 1 TABLE OF CONTENTS Section 1.0 2.0 Pacae INTRODUCTION.
Theinformation contained hereinisforthesoleuseofCustomer.
InordertoavoidImpairment ofrightsofAdvancedNuclearFuelsCorporation inpatentsorinventions whichmaybeincludedintheinformation contained inthisdocument, therecipient, byitsacceptance ofthisdocument, agreesnottopublishormakepublicuse(inthepatentuseoftheterm)ofsuchinformation untilsoauthorized inwritingbyAdvancedNuclearFuelsCorporation oruntilaftersix(6)monthsfollowing termination orexpiration oftheaforesaid Agreement andanyextension thereof,unlessotherwise expressly providedintheAgreement.
NorightsorlicensesInortoanypatentsareimpliedbythefurnishing ofthisdocu-ment.XNNFF00.765(1 ANF-87-126 Revision1TABLEOFCONTENTSSection1.02.0PacaeINTRODUCTION.
~..............,....,................................
~..............,....,................................
1FUELMECHANICAL DESIGNANALYSIS...................................
1 FUEL MECHANICAL DESIGN ANALYSIS...................................
23.03.23.2.13.2.33.2.5THERMALHYDRAULIC DESIGNANALYSIS..............
2 3.0 3.2 3.2.1 3.2.3 3.2.5 THERMAL HYDRAULIC DESIGN ANALYSIS..............
~..................
~..................
3HydraulicCharacteri zation........................................
3 Hydraul i c Characteri zati on........................................
3HydraulicCompatibility...........................................
3 Hydraul i c Compatibility...........................................
3FuelCenterline Temperature.......................................
3 Fuel Centerline Temperature.......................................
3BypassFlowe~~~~~~~~~~~~~~~~~~~~~~~~\~~~~~~~~~~~~~~~~~~~~~~~~~~~~33.33.3.13.3.2.3.34.0MCPRFuelCladdingIntegrity SafetyLimit...........
3 B ypass Flowe~~~~~~~~~~~~~~~~~~~~~~~~\~~~~~~~~~~~~~~~~~~~~~~~~~~~~3 3.3 3.3.1 3.3.2.3.3 4.0 MCPR Fuel Cladding Integrity Safety Limit...........
CoolantThermodynamic Conditions DesignBasisRadialPowerDistribution.
Coolant Thermodynamic Conditions Design Basis Radial Power Distribution.
DesignBasisLocalPowerDistribution.
Design Basis Local Power Distribution.
NUCLEARDESIGNANALYSIS..
NUCLEAR DESIGN ANALYSIS..
~~~~~~~~~0~~~~33~~~~~~~~~~~~~4~~~~~~~\~~~~~~~~~~~~~~~54.14.24.2.14.2.24.2.45.0FuelBundleNuclearDesignAnalysis.......
~~~~~~~~~0~~~~3 3~~~~~~~~~~~~~4~~~~~~~\~~~~~~~~~~~~~~~5 4.1 4.2 4.2.1 4.2.2 4.2.4 5.0 Fuel Bundle Nuclear Design Analysis.......
CoreNuclearDesignAnalysisCoreConfiguration.......
Core Nuclear Design Analysis Core Configuration.......
CoreReactivity Characteristics...,....,..
Core Reactivity Characteristics...,....,..
CoreHydrodynamic Stability.....
Core Hydrodynamic Stability.....
~.ANTICIPATED OPERATIONAL OCCURRENCES......,
~.ANTICIPATED OPERATIONAL OCCURRENCES......,~~~~~~~~~~~~~~~~~~~~~~~~5~~~~~~~5 5 6~~~~~~~~~~~~~~~~~~~~~~~~7 5.1 5.2 5.3 5.4 5.5 5.6 5.7 6.0 6.1 F 1.1 itlons~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~9~~~~~~~~~~~~~~~~~~~~~~~~9 10 10 Loss-Of-Coolant Accident....,,......
~~~~~~~~~~~~~~~~~~~~~~~~5~~~~~~~556~~~~~~~~~~~~~~~~~~~~~~~~75.15.25.35.45.55.65.76.06.1F1.1itlons~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~9~~~~~~~~~~~~~~~~~~~~~~~~91010Loss-Of-Coolant Accident....,,......
Break Location Spectrum........
BreakLocationSpectrum........
10 Analysis Of Plant Transients At Rated Cond Analyses For Reduced Flow Operation.......
10AnalysisOfPlantTransients AtRatedCondAnalysesForReducedFlowOperation.......
Analyses For Reduced Power Operation......
AnalysesForReducedPowerOperation......
ASME Overpressurization Analysis..........
ASMEOverpressurization Analysis..........
Control Rod Withdrawal Error (CRWE)Fuel Loading Error........
ControlRodWithdrawal Error(CRWE)FuelLoadingError........
Determination Of Thermal Margins..........
Determination OfThermalMargins..........
POSTULATED ACCIDENTS...
POSTULATED ACCIDENTS...
ANF-87-1RevisionTABLEOFCONTENTS(Continued)
ANF-87-1 Revision TABLE OF CONTENTS (Continued)
Section6.1.26.1.36.27.07.1T.1.17.1.27.27.2.17.2.27.2.3737.3.17.3.28.0LimitingSafetySystemSettings......
Section 6.1.2 6.1.3 6.2 7.0 7.1 T.1.1 7.1.2 7.2 7.2.1 7.2.2 7.2.3 73 7.3.1 7.3.2 8.0 Limiting Safety System Settings......
HCPRFuelCladdingIntegrity SafetyLSteamDomePressureSafetyLimitLimitingConditions ForOperation.
HCPR Fuel Cladding Integrity Safety L Steam Dome Pressure Safety Limit Limiting Conditions For Operation.
AveragePlanarLinearHeatGeneration MinimumCriticalPowerRatio~~~~~~~~~~~~~~~~~~~~~~~~~~~~~imitRateimits.................
Average Planar Linear Heat Generation Minimum Critical Power Ratio~~~~~~~~~~~~~~~~~~~~~~~~~~~~~imit Rate imits.................
L'HGRLlmlts~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~Surveillance Requirements......
L'HGR Llml ts~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~Surveillance Requirements......
.....ScramInsertion TimeSurveillance....
.....Scram Insertion Time Surveillance....
Stability Surveillance..........
Stability Surveillance..........
METHODOLOGY REFERENCES..........
METHODOLOGY REFERENCES..........
Line 57: Line 53:
==9.0 ADDITIONAL==
==9.0 ADDITIONAL==
REFERENCES.......
REFERENCES.......
reakSizeSpectrum...............................................
reak Size Spectrum...............................................
BAPLHGRAnalyses.............................'.....................
B APLHGR Analyses.............................'.....................
HControlRodDropAccident........,...
H Control Rod Drop Accident........,...
TECHNICAL SPECIFICATIONS........
TECHNICAL SPECIFICATIONS........
Pacae101011121212121212111414141516APPENDICES A.SINGLELOOPOPERATION.............
Pacae 10 10 11 12 12 12 12 12 12 1 1 14 14 14 15 16 APPENDICES A.SINGLE LOOP OPERATION.............
A-1B.SEISMIC-LOCA EVALUATION....,.................,.........,..........
A-1 B.SEISMIC-LOCA EVALUATION....,.................,.........,..........
B-1 ANF-87-126 Revision1LISTOFTABLESTablePacae4.1Neutronic DesignValues...........................................
B-1 ANF-87-126 Revision 1 LIST OF TABLES Table Pacae 4.1 Neutronic Design Values...........................................
23B.1Comparison OfPhysicalAndStructural Characteristics For8x8And9x9FuelAssemblies.........................
23 B.1 Comparison Of Physical And Structural Characteristics For 8x8 And 9x9 Fuel Assemblies.........................
.....B-2LISTOFFIGURESFiciure3.1Susquehanna Unit2Powervs.Flow....Cycle3Hydraulic DemandCurvePacae173.23.33.5Susquehanna Unit2DesignBasisLocalDesignBasisLocalDesignBasisLocalFuel.Cycle3DesignBasisRadialPower..............
.....B-2 LIST OF FIGURES Ficiur e 3.1 Susquehanna Unit 2 Power vs.Flow....Cycle 3 Hydraulic Demand Curve Pacae 17 3.2 3.3 3.5 Susquehanna Unit 2 Design Basis Local Design Basis Local Design Basis Local Fuel.Cycle 3 Design Basis Radial Power..............
18PowerDistribution
18 Power Distribution
-ANFXN-29x9Fuel.........
-ANF XN-2 9x9 Fuel.........
19PowerDistribution
19 Power Distribution
-ANFXN-19x9Fuel.........
-ANF XN-1 9x9 Fuel.........
20PowerDistribution
20 Power Distribution
-GE8x8R(Central) 213.64.14.24.34.45.15.2~~......2224~~~~~~~~25~~~~~~~~~26272829DesignBasisLocalPowerDistribution
-GE 8x8R (Central)21 3.6 4.1 4.2 4.3 4.4 5.1 5.2~~......22 24~~~~~~~~25~~~~~~~~~26 27 28 29 Design Basis Local Power Distribution
-GE(Peripheral) 8x8RFuel..........
-GE (Peripheral) 8x8R Fuel..........
~~0~~~~Susquehanna Unit2Cycle3Enrichment Distribution ForANF92-344L-9G4 XN-2FuelLattice.Susquehanna Unit2Cycle3Enrichment Distribution ForANF92-344L-10G5 XN-2FuelLattice.Susquehanna Unit2Cycle3Reference CoreLoadingPlan...Susquehanna Unit2Cycle3-CorePowervs.CoreFlow......
~~0~~~~Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-9G4 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-10G5 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Reference Core Loading Plan...Susquehanna Unit 2 Cycle 3-Core Power vs.Core Flow......
Susquehanna Unit2Cycle3ControlRodWithdrawal ErrorAnalysisLimitingInitialControlRodPattern..
Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern..Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit.......  
Susquehanna Unit2Cycle3FlowMCPROperating Limit.......  
~fj ANF-87-126 Revision 1
~fj ANF-87-126 Revision


==11.0INTRODUCTION==
==1.0 INTRODUCTION==


Thisreportprovidestheresultsoftheanalysesperformed byAdvancedNuclearFuelsCorporation (ANF)*insupportoftheCycle3reloadforSusquehanna Unit2,whichisscheduled tocommenceoperation inthespringof1988.Thisreportisintendedtobeusedinconjunction withANFtopicalreport~XN-Np--191A,914,R111,Nppti1111NCompanyMethodology toBWRReloads,"
This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF)*in support of the Cycle 3 reload for Susquehanna Unit 2, which is scheduled to commence operation in the spring of 1988.This report is intended to be used in conjunction with ANF topical report~XN-Np--191 A, 91 4, R 11 1, Nppti 1 1 1 1 N Company Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list.However, LHGR mechanical design limits (Reference 9.1)and plant transient simulation model developments (Reference 9.141 b 1 dbyANF b 4 t NRN P 1 F~F-Volume 4, Revi'sion 1.Both References 9.1 and 9.2 have been approved by the NRC for use in referencing in license applications.
whichdescribes theanalysesperformed insupportofthisreload,identifies themethodology usedforthoseanalyses, andprovidesagenericreference list.However,LHGRmechanical designlimits(Reference 9.1)andplanttransient simulation modeldevelopments (Reference 9.141b1dbyANFb4tNRNP1F~F-Volume4,Revi'sion 1.BothReferences 9.1and9.2havebeenapprovedbytheNRCforuseinreferencing inlicenseapplications.
Section numbers in this 9 t 1 9 dtd tt b 1 X-N--fNJ, olume 4, Revision 1.The Susquehanna Unit 2 Cycle 3 core will comprise a total of 764 fuel assemblies, including 236 unirradiated ANF XN-2 9x9 assemblies, 324 irradiated ANF XN-1 9x9 assemblies, 112 irradiated General Electric 8x8R fuel assemblies (central region), and 92 irradiated GE 8x8R assemblies in the peripheral region.The reference core configuration is described in Section 4.2.The design and safety analyses reported in this document were based on the design and operational assumptions in effect for Susquehanna Unit 2 during the previous operating cycle.Additional information and the results of design studies covering the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.3.
Sectionnumbersinthis9t19dtdttb1X-N--fNJ,olume4,Revision1.TheSusquehanna Unit2Cycle3corewillcompriseatotalof764fuelassemblies, including 236unirradiated ANFXN-29x9assemblies, 324irradiated ANFXN-19x9assemblies, 112irradiated GeneralElectric8x8Rfuelassemblies (centralregion),and92irradiated GE8x8Rassemblies intheperipheral region.Thereference coreconfiguration isdescribed inSection4.2.Thedesignandsafetyanalysesreportedinthisdocumentwerebasedonthedesignandoperational assumptions ineffectforSusquehanna Unit2duringthepreviousoperating cycle.Additional information andtheresultsofdesignstudiescoveringthedevelopment of9x9fuelassemblies forBWRreloadsarecontained inReference 9.3.
f ANF-87-126 Revision 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ANF Fuel Design Report: Reference 9.1 To assure that the expected power history for the fuels to be irradiated during Cycle 3 of Susquehanna Unit 2 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits (Figure 3.3 of Reference 9.1)have been specified.
f ANF-87-126 Revision12.0FUELMECHANICAL DESIGNANALYSISApplicable ANFFuelDesignReport:Reference 9.1Toassurethattheexpectedpowerhistoryforthefuelstobeirradiated duringCycle3ofSusquehanna Unit2isboundedbytheassumedpowerhistoryinthefuelmechanical designanalysis, LHGRoperating limits(Figure3.3ofReference 9.1)havebeenspecified.
In addition, an LHGR transient operating'imit for Anticipated Operating Occurrences (Figure 3.4 of Reference 9.1)has been specified for ANF 9x9 fuel.Additional information on rod bow, as requested in the NRC's safety evaluation report for Reference 9.1, has been transmitted in Reference 9.4.
Inaddition, anLHGRtransient operating'imit forAnticipated Operating Occurrences (Figure3.4ofReference 9.1)hasbeenspecified forANF9x9fuel.Additional information onrodbow,asrequested intheNRC'ssafetyevaluation reportforReference 9.1,hasbeentransmitted inReference 9.4.
ANF-87-126 Revision 1 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 H draul i c Char aeter i zat i on 3.2.1 H draulic Com atibilit Component hydraulic resistances for the constituent fuel types in the Susquehanna Unit 2 Cycle 3 core have been determined in single phase flow tests of full scale assemblies.
ANF-87-126 Revision13.0THERMALHYDRAULIC DESIGNANALYSIS3.2HdraulicCharaeterization3.2.1HdraulicComatibilitComponent hydraulic resistances fortheconstituent fueltypesintheSusquehanna Unit2Cycle3corehavebeendetermined insinglephaseflowtestsoffullscaleassemblies.
Figure 3.1 shows the hydraulic demand curves for ANF 9x9 fuel and GE 8x8R fuel in the Susquehanna Unit 2 core.The similar hydraulic performance indicates compatibility for co-residence in'he Susquehanna Unit 2 core.Applicable Generic Report 3.2'Fuel Centerline Tem erature~~Reference 9.1.2.2~21 Calculated Bypass Flow Fraction at 104%Power/100%
Figure3.1showsthehydraulic demandcurvesforANF9x9fuelandGE8x8RfuelintheSusquehanna Unit2core.Thesimilarhydraulic performance indicates compatibility forco-residence in'heSusquehanna Unit2core.Applicable GenericReport3.2'FuelCenterline Temerature~~Reference 9.1.2.2~21Calculated BypassFlowFractionat104%Power/100%
Flow 10.1%3.3 MCPR Fuel Claddin Inte rit Safet Limit Safety Limit MCPR=1.06 3.3.1 Coolant Thermod namic Condition Rated Thermal Power Feedwater Flowrate (at SLMCPR)Core Pressure (at SLMCPR)Feedwater Temperature 3293 Mwt 16.1 Mlbm/hr 1042.9 psia 383'F ANF-87-1 Revision 3.3.2 Desi n Basis Radial Power Distribution See Figure 3.2 3.3.3 Desi n Basis Local Power Distribution See Figures 3.3 through 3.6 ANF-87-126 Revision 1 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fue Bund e Nuclea Desi n Anal sis Assembly Average Enrichment Radial Enrichment Distribution Axial Enrichment Distribution Burnable Poisons Note: Burnabl e poi sons are distributed uniformly over the enriched length of the designated rods.The natural urania axial blanket sections do not contain burnable absorber material.Non-Fueled Rods Neutronic Design Parameters 3.33%Figure 4.1 and 4.2 Uniform 3.44%with 6" natural uranium top blanket Figure 4.1 and 4.2 Figure 4.1 and 4.2 Table 4.1 4.2 Core Nuclear Desi n Anal sis';2.1 Core Confi oration Figure 4.3 Core Exposure at EOC2, HWd/HTU Core Exposure at BOC3, MWd/HTU Core Exposure at EOC3, HWd/MTU Maximum Cycle 3 Licensing Exposure Limit, HWd/MTU 18350.7 10911.2 21740.8 22076 ANF-87-12 Revision 4.2.2 ore Reactiv't Characteris ics BOC Cold K-effective, All Rods Out BOC Cold K-effective, Strongest Rod Out 1.11353 0.98524 Reactivity Defect (R-Value)0.00%rho 4.2.4 Standby Liquid Control System Reactivity, Cold Conditions, 660 ppm I Core H drod namic Stabilit 0.98348 Power/flow Map Figure 4.4 Power Flow State Points 64/42*69/47**66/45**Deca Ratio COTRA 0.82 0.75 0.75*Two pump minimum flow-APRN Rod Block intercept point.Extended operation at lower flow is not allowed by Technical Specifications.
Flow10.1%3.3MCPRFuelCladdinInteritSafetLimitSafetyLimitMCPR=1.063.3.1CoolantThermodnamicCondition RatedThermalPowerFeedwater Flowrate(atSLMCPR)CorePressure(atSLMCPR)Feedwater Temperature 3293Mwt16.1Mlbm/hr1042.9psia383'F ANF-87-1Revision3.3.2DesinBasisRadialPowerDistribution SeeFigure3.23.3.3DesinBasisLocalPowerDistribution SeeFigures3.3through3.6 ANF-87-126 Revision14.0NUCLEARDESIGNANALYSIS4.1FueBundeNucleaDesinAnalsisAssemblyAverageEnrichment RadialEnrichment Distribution AxialEnrichment Distribution BurnablePoisonsNote:Burnablepoisonsaredistributed uniformly overtheenrichedlengthofthedesignated rods.Thenaturaluraniaaxialblanketsectionsdonotcontainburnableabsorbermaterial.
**Operation at less than 45%flow requires APRH/LPRN surveillance.
Non-Fueled RodsNeutronic DesignParameters 3.33%Figure4.1and4.2Uniform3.44%with6"naturaluraniumtopblanketFigure4.1and4.2Figure4.1and4.2Table4.14.2CoreNuclearDesinAnalsis';2.1CoreConfiorationFigure4.3CoreExposureatEOC2,HWd/HTUCoreExposureatBOC3,MWd/HTUCoreExposureatEOC3,HWd/MTUMaximumCycle3Licensing ExposureLimit,HWd/MTU18350.710911.221740.822076 ANF-87-12 Revision4.2.2oreReactiv't Characteris icsBOCColdK-effective, AllRodsOutBOCColdK-effective, Strongest RodOut1.113530.98524Reactivity Defect(R-Value) 0.00%rho4.2.4StandbyLiquidControlSystemReactivity, ColdConditions, 660ppmICoreHdrodnamicStabilit0.98348Power/flow MapFigure4.4PowerFlowStatePoints64/42*69/47**66/45**DecaRatioCOTRA0.820.750.75*Twopumpminimumflow-APRNRodBlockintercept point.Extendedoperation atlowerflowisnotallowedbyTechnical Specifications.
In addition, operation at power/flow, combinations above and to the left of the line connecting these two points requires APRH/LPRtl surveillance.
**Operation atlessthan45%flowrequiresAPRH/LPRN surveillance.
See Figbre 4.4.
Inaddition, operation atpower/flow, combinations aboveandtotheleftofthelineconnecting thesetwopointsrequiresAPRH/LPRtl surveillance.
ANF-87-126 Revision 1 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Methodology Report References 9.5 5 9.7 5.1 Anal sis Of Plant Transients At Rated Conditions Reference 9.6 Limiting Transient(s):
SeeFigbre4.4.
Load Rejection Without Bypass (LRWB)Feedwater Controller Failure (FWCF)Loss of Feedwater Heating (LFWH)Event LRWB Power*100%FWCF 100%100%1 16.8%LFWH 100%100%121'%233%123%1179 1078%Rated%Rated Maximum Maximum Maximum Pressure Flow Heat Flux Power ,~aia I 100%116.2%267%1194 Del ta CPR**0.24 0.23 0.16 Model COTRANSA/XCOBRA-T COTRANSA/XCOBRA-T PTSBWR3/XCOBRA Single Loop Operation:
ANF-87-126 Revision15.0ANTICIPATED OPERATIONAL OCCURRENCES Applicable GenericTransient AnalysisMethodology ReportReferences 9.559.75.1AnalsisOfPlantTransients AtRatedConditions Reference 9.6LimitingTransient(s):
Appendix A 5.2 Anal ses For Reduced Flow 0 eration Reference 9.6 Limiting Transient(s):
LoadRejection WithoutBypass(LRWB)Feedwater Controller Failure(FWCF)LossofFeedwater Heating(LFWH)EventLRWBPower*100%FWCF100%100%116.8%LFWH100%100%121'%233%123%11791078%Rated%RatedMaximumMaximumMaximumPressureFlowHeatFluxPower,~aiaI100%116.2%267%1194DeltaCPR**0.240.230.16ModelCOTRANSA/
Recirculation Flow Increase Transient (RFIT)*104%power used in analysis as design bases.**Delta-CPR results for most limiting fuel type.
XCOBRA-TCOTRANSA/
ANF-87-1 Revision 5.3 Anal ses For Reduced Power 0 eration Reference 9.6 Limiting Transient(s):
XCOBRA-TPTSBWR3/XCOBRASingleLoopOperation:
Feedwater Controller Failure (FWCF)%Power Transient Delta CPR ANF 9x9 GE 8x8R 104 80 65 40 FWCF FWCF FWCF FWCF 0.23 0.25 0.28 0.31 0.20 0.23 0.26 0.28 5.4 ASME Over ressurization Anal sis Reference 9.6 Limiting Event Worst Single Failure Maximum Pressure Maximum Steam Dome Pressure Full MSIV Isolation Direct Sera 1297 psig 1281 psig 5.5 Control Rod Withdrawal Error CRWE Starting Control Rod Pattern for Analysis Figure 5.1 Rod Block Settin 105 106*107 108*100%Flow Distance Withdrawn~ft 4.0 4.5 5.0 5.0 Delta CPR 0.22 0.24 0.26 0.26*Rod Block Monitor settings recommended for Cycle 3 operation.
AppendixA5.2AnalsesForReducedFlow0erationReference 9.6LimitingTransient(s):
ANF-87-126 Revision 1 5.6 Fuel Loadin Error Maximum Delta CPR 0.16 5.7 Determination Of Thermal Har ins Summary of Thermal Margin Requirements Event LRWB FWCF LFWH CRWE Power 1P0%**1PP%**1PP%9c*100%Flow 100%100%100%100%Delta CPR*0.24 0.23 0.16 0.24 at 106%RBH 0.26 at 108%RBM MCPR Limit 1.30 1.29 1.22 1.30 1.32 HCPR Operating Limits at Rated Conditions MCPR 0 eratin Limit 1.30 at 106%RBM 1.32 at 108%RBM Reduced Flow MCPR Limits Figure 5.2 Power Dependent HCPR Operating Limit Results for Cycle 3: 100*+/100 80/100 65/100 40/100 Limiting Transient LRWB FWCF FWCF FWCF ANF 9x9 1.30 1.31 1.34 1.37 GE 8x8R 1.27 1.29 1.32 1.34 ,i*Delta CPR results for most limiting fuel type.**104%power used in analysis as design bases.  
Recirculation FlowIncreaseTransient (RFIT)*104%powerusedinanalysisasdesignbases.**Delta-CPR resultsformostlimitingfueltype.
ANF-87-1Revision5.3AnalsesForReducedPower0erationReference 9.6LimitingTransient(s):
Feedwater Controller Failure(FWCF)%PowerTransient DeltaCPRANF9x9GE8x8R104806540FWCFFWCFFWCFFWCF0.230.250.280.310.200.230.260.285.4ASMEOverressurization AnalsisReference 9.6LimitingEventWorstSingleFailureMaximumPressureMaximumSteamDomePressureFullMSIVIsolation DirectSera1297psig1281psig5.5ControlRodWithdrawal ErrorCRWEStartingControlRodPatternforAnalysisFigure5.1RodBlockSettin105106*107108*100%FlowDistanceWithdrawn
~ft4.04.55.05.0DeltaCPR0.220.240.260.26*RodBlockMonitorsettingsrecommended forCycle3operation.
ANF-87-126 Revision15.6FuelLoadinErrorMaximumDeltaCPR0.165.7Determination OfThermalHarinsSummaryofThermalMarginRequirements EventLRWBFWCFLFWHCRWEPower1P0%**1PP%**1PP%9c*100%Flow100%100%100%100%DeltaCPR*0.240.230.160.24at106%RBH0.26at108%RBMMCPRLimit1.301.291.221.301.32HCPROperating LimitsatRatedConditions MCPR0eratinLimit1.30at106%RBM1.32at108%RBMReducedFlowMCPRLimitsFigure5.2PowerDependent HCPROperating LimitResultsforCycle3:100*+/100 80/10065/10040/100LimitingTransient LRWBFWCFFWCFFWCFANF9x91.301.311.341.37GE8x8R1.271.291.321.34,i*DeltaCPRresultsformostlimitingfueltype.**104%powerusedinanalysisasdesignbases.  


10ANF-87-126 Revision16.0POSTULATED ACCIDENTS 6.1Loss-Of-Coolant AccidentSeismic-LOCA:AppendixB6.1.1BreakLocationSectrumReference 9.86.1.2BreakSizeSectrumReference 9.86.1.3MAPLHGRAnalsesANF9x9FuelReference 9.9LimitingBreak:Double-ended guillotine pipebreakRecirculation pumpdischarge line0.4Discharge Coefficient BundleAverageExposureGWDMTU0510152025303540MAPLHGR~kwft10.210.210.210.210.29.68.98.27.5PeakCladTemperature*
10 ANF-87-126 Revi si on 1 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident Sei smi c-LOCA: Appendix B 6.1.1 Break Location S ectrum Reference 9.8 6.1.2 Break Size S ectrum Reference 9.8 6.1.3 MAPLHGR Anal ses ANF 9x9 Fuel Reference 9.9 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 Discharge Coefficient Bundle Average Exposure GWD MTU 0 5 10 15 20 25 30 35 40 MAPLHGR~kw ft 10.2 10.2 10.2 10.2 10.2 9.6 8.9 8.2 7.5 Peak Clad Temperature*
~F206020692121214021472016183917521676PeakLocalMWR**~Percent3.93.73.74.85.22.71.00.70.5*Peakcladtemperatures forXN-1andXN-2fuelareboundedbytheseresults.**MetalWaterReaction.
~F 2060 2069 2121 2140 2147 2016 1839 1752 1676 Peak Local MWR**~Percent 3.9 3.7 3.7 4.8 5.2 2.7 1.0 0.7 0.5*Peak clad temperatures for XN-1 and XN-2 fuel are bounded by these results.**Metal Water Reaction.
llANF-87-1Revision6.2ControlRodDroAccidentSection8.0DroppedControlRodWorth,mkDopplerCoefficient, 1/kdk/dTEffective DelayedNeutronFractionFour-Bundle LocalPeakingFactormaximumDeposited FuelRodEnthalpy, cal/gmNumberofRodsExceeding 170cal/gm13.5-10.6x(10)60.00581.34205(250 12ANF-87-126 Revision17.0TECHNICAL SPECIFICATIONS 7.1LimitinSafetSstemSettins7.1.1MCPRFuelCladdinInteritSafetLimitMCPRSafetyLimit1.067.1.2SteamDomePressureSafetLimitPressureSafetyLimit(asmeasuredinsteamdome)1325psigAnalysisshowsthatasteamdomepressuresafetylimitof1358psigisallowedbutthe1325psigvalueusedinCycle2istobeconservatively retained.
ll ANF-87-1 Revision 6.2 Control Rod Dro Accident Section 8.0 Dropped Control Rod Worth, mk Doppler Coefficient, 1/k dk/dT Effective Delayed Neutron Fraction Four-Bundle Local Peaking Factor maximum Deposited Fuel Rod Enthalpy, cal/gm Number of Rods Exceeding 170 cal/gm 13.5-10.6 x(10)6 0.0058 1.34 205 (250 12 ANF-87-126 Revision 1 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Safet Limit Pressure Safety Limit (as measured in steam dome)1325 psig Analysis shows that a steam dome pressure safety limit of 1358 psig is allowed but the 1325 psig value used in Cycle 2 is to be conservatively retained.7.2 Limitin Conditions For 0 er ation 7.2.1 Avera e Planar Linear Heat Generation Rate Limits Bundle Average Exposure GWD MT 0 5 10 15 20 25 30 35 40 MAPLHGR Limits kw ft ANF 9x9 Fuel 10.2 10.2 10.2 10.2 10.2 9.6 8.9 8.2 7.5 13 ANF-87-1 Revision 7.2.2 Minimum Critical Power Ratio MCPR Operating Limits at Rated Conditions:
7.2LimitinConditions For0eration7.2.1AveraePlanarLinearHeatGeneration RateLimitsBundleAverageExposureGWDMT0510152025303540MAPLHGRLimitskwftANF9x9Fuel10.210.210.210.210.29.68.98.27.5 13ANF-87-1Revision7.2.2MinimumCriticalPowerRatioMCPROperating LimitsatRatedConditions:
MCPR 0 eratin Limit 1.30 at 106%RBM 1.32 at 108%RBM MCPR Operating Limits at Off-Rated Conditions:
MCPR0eratinLimit1.30at106%RBM1.32at108%RBMMCPROperating LimitsatOff-Rated Conditions:
At Reduced Flow Figure 5.2 Total Core Recirculation Flow%Rated 100 96 92 83 76 60 50 40 Reduced Flow MCPR 0 eratin Limit 1.12 1.14 1.16 1,20 1.23 1.31 1.44 1.61 At Reduced Power Power Level%Rated 100*80 65 40 Reduced Power MCPR 0 eratin Limit 1.30 1.31 1.34 1.37*104%power used in analysis as design bases.
AtReducedFlowFigure5.2TotalCoreRecirculation Flow%Rated10096928376605040ReducedFlowMCPR0eratinLimit1.121.141.161,201.231.311.441.61AtReducedPowerPowerLevel%Rated100*806540ReducedPowerMCPR0eratinLimit1.301.311.341.37*104%powerusedinanalysisasdesignbases.
ANF-87-126 Revision I 7.2.3 LHGR Limits LHGR Limits Figures 3.3 and 3.4 of Reference 9.1 7.3 Surveillance Re uirements 7.3.1 Scram Insertion Time Surveillance Thermal limits established in Section 5.0 are based on minimum acceptable scram insertion performance as defined in the Technical Specifications.
ANF-87-126 RevisionI7.2.3LHGRLimitsLHGRLimitsFigures3.3and3.4ofReference 9.17.3Surveillance Reuirements 7.3.1ScramInsertion TimeSurveillance Thermallimitsestablished inSection5.0arebasedonminimumacceptable scraminsertion performance asdefinedintheTechnical Specifications.
No additional surveillance for scram insertion is required for validation of thermal limits..3.2 Stabilit Surveillance
Noadditional surveillance forscraminsertion isrequiredforvalidation ofthermallimits..3.2StabilitSurveillance
~~Power/Flow Map Figure 4.4 The Unit 2 Cycle 2 Technical Specifications require APRM/LPRM surveillance to the left of the 45%Constant Flow line and above the 80%Rod Block line.Based on core hydrodynamic stability analyses, operation at power/flow combinations above and to the left of the line connecting the 66%Power/45%Flow and 69%Power/47%Flow points but below the APRM Rod Block line needs to be added to the APRM/LPRM surveillance requirement (see Section 4.2.4).
~~Power/Flow MapFigure4.4TheUnit2Cycle2Technical Specifications requireAPRM/LPRM surveillance totheleftofthe45%ConstantFlowlineandabovethe80%RodBlockline.Basedoncorehydrodynamic stability
15 ANF-87-126 Revision 1 8.0 METHODOLOGY REFERENCES See XN-NF-80-19(P)(A), Volume 4, Revision 1 for complete bibliography.  
: analyses, operation atpower/flow combinations aboveandtotheleftofthelineconnecting the66%Power/45%
\q,"~r 0~)6 44-I 16 ANF-87-126 Revision 1 9.0 ADDITIONAL REFERENCES 9.1"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"~X---F,R.X,Addll 1 FXCF l,lhhl d, Washington, September 4, 1986.9.2"Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal 111<<N hd1 d P Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, January, 1987.9.3"Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power Research Institute, Palo Alto, California, Hay 1, 1984.9.4 Letter, G.N.Ward (ANF)to G.C.Lainas (NRC),"Additional Information on Rod Bow," serial no.GNW:021:87, dated March 11, 1987.9.5"Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"~h-p--,h 11 X,AddN1F1C l,ltlhl d, Washington, November 16, 1981.~~~9.6"Susquehanna Unit 2 Cycle 3 Plant Transient Analysis," ANF-87-125, Rev.2, Advanced Nuclear Fuels Corporation, Richland, Washington, November 1987.9.7 9~8"XCOBRA-T:
Flowand69%Power/47%
A Computer Code for BWR Transient Thermal-Hydraulic Core A 1 i,"X~, 1>>d 2, Advanced Nuclear Fuels Corporation, Richland, Washington, February 1987."Generic LOCA Break Spectrum Analysis BWR 3 8 4 with Hodified Low Pressure Coolant Injection Logic Using the EXEH Evaluation Model," XN-NF-~84-117 P, Advanced Nuclear Fuels Corporation, Richland, Washington, December 1984.9.9"Susquehanna LOCA-ECCS Analysis HAPLHGR Results for ENC 9x9 Fuel," XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.9.10"Principal Reload Fuel Design Parameters, Fuel Design, Susquehanna Unit 2 Reload XN-2," XN-NF-1058, Advanced Nuclear Fuels Corporation, Richland, Washington, March 1987, Formerly Exxon Nuclear Company.
FlowpointsbutbelowtheAPRMRodBlocklineneedstobeaddedtotheAPRM/LPRM surveillance requirement (seeSection4.2.4).
Advanced Nuclear Fuels 9x9 0~)O~I General Electric 8x8R~o p co o C p 0 O I Q~o L I p 0 q>vCi I.>o oo I K 100.00.105.00 TIO.OO 115.M 120.00 125.00 QO.M Q5.M 140.00 Assembly Flow Rate, KLB/HR 0 CO%5.00 150.00 Figure 3.1 Susquehanna Unit 2 Cycle 3 Hydraulic Oemand Curve Power vs.Flow 80 70 60 50 00 C)CL So 20 10 0 0 0.2 0.0 0.6 0.8 1 1.2 RRDIFIL POHER PERKING Figure 3.2 Susquehanna 2 Cycle 3 Oesign Basis Radial Power 19 ANF-87-126 Revision 1*~: 0.88: 0.91: 0.96: 1.04: 1.02: 1.04: 0.96: 1.00: 0.96:*~**~: 0.91: 0.93: 0.98: 1.07: 0.91: 1.07: 0.97: 1.04: 1.01*~**~0.96: 0.98: 0.90: 1.04: 1.03: 1.04: 1.04: 0.99: 0.96:**~1.04: 1.07: 1.04: 1.00: 0.99: 1,00: 1.05: 0.94: 1.04**~*: 1.02:*~0.91: 1.03 0.99: 0.00: 0.98: 1.05: 1.07: 1.04*~*~1.04: 1.07: 1.04: 1.00: 0.98: 0.00: 1.03: 0.94: 1.05: 0.96: 0'7: 1'4: 1.05: 1.05: 1.03: 1.06: 1.00: 0,97 1.00: 1.04: 0.99: 0,94: 1.07: 0.94: 1.00: 0.94 1.01 0.96: 1.01: 0.96: 1.04: 1.04: 1.05: 0.97: 1.01 0.97 Figure 3.3 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-2 9X9 Fuel 20 ANF-87-1 Revisio*~*~0.91: 0.92: 0.95: 1.01: 1.01: 1.01: 0.96: 0.98: 0.95*~*~*~0.92: 0.94: 0.98: 0.97;1.05: 0.95: 0.99: 0.95: 0.98*~*~*~*0.95: 0.98: 0,93: 1.06: 1.05: 1.06: 1.05: 0.97: 0.96*~**1.01: 0.97: 1,06: 1.03: 1,03: 1.04: 1.07: 1.06: 1.02*~*~1.01: F 05: 1.05: 1.03: 0.00: 1.01: 1.07: 1.06: 1.011 01'95 1.06: 1.04: 1.01: 0 F 00: 1.04: 0.96: 1.02 0.96: 0.99 1.05: 1.07: 1.07: 1.04: 1.06: 1.00: 0.96 0.98: 0.95: 0.97: 1.06: 1.06: 0.96 1.00: 0.95: 0.98 0.95: 0.98: 0.96: 1.02: 1.01: 1.02: 0.96: 0.98: 0.96 Figure 3.4 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-1 9X9 Fuel 21 ANF-87-126 Revision 1**~*~1.03: 1.00: 1.00: 1.00: F 00: 1.00: 1.01: 1.03*~*: 1.00: 0.98: 1.00*~*1.02: 1.02: 1.03: 1.00: 1.01*~*: 1.00*~**~1.00: 1.01: 1.01: 1.01: 0.90: 1.03: 1.00:*: 1.00: 1.02*~*1.01: 0.89: 0.00: 1.01: 1.02: 1.00*~*: 1.00*~1.02 1.01: 0.00: 0.89: 1.01: 0.99: 1.00*~*: 1.00*~*1.03: 0.90 1.01: 1.01: 0.98: 1.00: 1.00 1.01: 1.00: 1.03: 1.02 0.99: 1.00: 0.98: 1.00 1.03: 1.01: 1.00: 1.00: 1.00 1.00: 1.00: 1.03 Figure 3.5 Design Basis Local Power Distribution General Electric (Central)SXSR Fuel 22 ANF-87-'evisio*~1.00: 1,00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*~0*~1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*~**~*100*~**~1.00;1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*: 1.00: 1.00**~1.00: 1.00: 0.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00: 0.00: 1.00: 1.00: 1.00 1.00**100*~*1.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00 Figure 3.6 Design Basis Local Power Distribution General Electric (Peripheral)
15ANF-87-126 Revision18.0METHODOLOGY REFERENCES SeeXN-NF-80-19(P)(A),
SXSR Fuel 23 ANF-87-126 Revision 1 TABLE 4.1 NEUTRONIC DESIGN VALUES Fuel Pellet Reference 9.10 Fuel Rod Reference 9.10 Fuel Assembl Reference 9.10 Core Data Number of fuel assemblies Rated thermal power, HW Rated core flow, Hlbm/hr Core inlet subcooling, Btu/ibm Hoderator temperature, F Channel thickness, inch Fuel assembly pitch, inch Wide water gap thickness, inch Narrow water gap thickness, inch'64 3293 100 24.0 548.8.080 6.00 0.562 0.562 Control Rod Data Absorber material Total blade span, inch Total blade support span, inch Blade thickness, inch Blade face,-to-face internal dimension, inch Absorber rods per blade Absorber rod outside diameter, inch Absorber rod inside diameter, inch Absorber density,%of theoretical B4C 9.75 1.58 0.260 0.200 76 0.188 0.138 70.0 24 ANF-87-126 Revision**: LL: L: HL: M: N*H: HL: HL*~4 HL M: MH: N*HH N*'HL*~*: HL*M".H*H: H O': HH H: HL**: H: HH: H: H: H*\H: H M H: N*: H: H: W: HH: H: HH*'A'~.*N: MH: H H NH W: MH H*HL: H*: MH: H H MH: MH HL HL: H: H: M~MH: H*: M ML ML L: HL: HL: H M: HL: HL-: L LL RODS (1)L RODS (5)HL RODS (16)H RODS (20)MH RODS (13)H RODS (15)H*RODS (9)W RODS (2)1.45 W/0 U235 1.95 W/0 U235 2.55 W/0 U235 3.27 W/0 U235 4.23 W/0 U235 4.66 W/0 U235 3.27 W/0 U235+4.00 W/0 GD203 INERT WATER ROD Figure 4.1 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-9G4 Xi4-2 Fuel Lattice 25*************
Volume4,Revision1forcompletebibliography.  
\q,"~r0~)644-I 16ANF-87-126 Revision19.0ADDITIONAL REFERENCES 9.1"GenericMechanical DesignforExxonNuclearJetPumpBWRReloadFuel,"~X---F,R.X,Addll 1FXCFl,lhhld,Washington, September 4,1986.9.2"ExxonNuclearMethodology forBoilingWaterReactors, THERMEX:Thermal111<<Nhd1dPRevision2,AdvancedNuclearFuelsCorporation,
: Richland, Washington, January,1987.9.3"Demonstration of9x9Assemblies forBWRs,"EPRINP-3468,ElectricPowerResearchInstitute, PaloAlto,California, Hay1,1984.9.4Letter,G.N.Ward(ANF)toG.C.Lainas(NRC),"Additional Information onRodBow,"serialno.GNW:021:87, datedMarch11,1987.9.5"ExxonNuclearPlantTransient Methodology forBoilingWaterReactors,"
~h-p--,h11X,AddN1F1C l,ltlhld,Washington, November16,1981.~~~9.6"Susquehanna Unit2Cycle3PlantTransient Analysis,"
ANF-87-125,Rev.2,AdvancedNuclearFuelsCorporation,
: Richland, Washington, November1987.9.79~8"XCOBRA-T:
AComputerCodeforBWRTransient Thermal-Hydraulic CoreA1i,"X~,1>>d2,AdvancedNuclearFuelsCorporation,
: Richland, Washington, February1987."GenericLOCABreakSpectrumAnalysisBWR384withHodifiedLowPressureCoolantInjection LogicUsingtheEXEHEvaluation Model,"XN-NF-~84-117P,AdvancedNuclearFuelsCorporation,
: Richland, Washington, December1984.9.9"Susquehanna LOCA-ECCS AnalysisHAPLHGRResultsforENC9x9Fuel,"XN-NF-86-65,AdvancedNuclearFuelsCorporation,
: Richland, Washington, May1986.9.10"Principal ReloadFuelDesignParameters, FuelDesign,Susquehanna Unit2ReloadXN-2,"XN-NF-1058, AdvancedNuclearFuelsCorporation,
: Richland, Washington, March1987,FormerlyExxonNuclearCompany.
AdvancedNuclearFuels9x90~)O~IGeneralElectric8x8R~opcooCp0OIQ~oLIp0q>vCiI.>oooIK100.00.105.00TIO.OO115.M120.00125.00QO.MQ5.M140.00AssemblyFlowRate,KLB/HR0CO%5.00150.00Figure3.1Susquehanna Unit2Cycle3Hydraulic OemandCurvePowervs.Flow 8070605000C)CLSo2010000.20.00.60.811.2RRDIFILPOHERPERKINGFigure3.2Susquehanna 2Cycle3OesignBasisRadialPower 19ANF-87-126 Revision1*~:0.88:0.91:0.96:1.04:1.02:1.04:0.96:1.00:0.96:*~**~:0.91:0.93:0.98:1.07:0.91:1.07:0.97:1.04:1.01*~**~0.96:0.98:0.90:1.04:1.03:1.04:1.04:0.99:0.96:**~1.04:1.07:1.04:1.00:0.99:1,00:1.05:0.94:1.04**~*:1.02:*~0.91:1.030.99:0.00:0.98:1.05:1.07:1.04*~*~1.04:1.07:1.04:1.00:0.98:0.00:1.03:0.94:1.05:0.96:0'7:1'4:1.05:1.05:1.03:1.06:1.00:0,971.00:1.04:0.99:0,94:1.07:0.94:1.00:0.941.010.96:1.01:0.96:1.04:1.04:1.05:0.97:1.010.97Figure3.3DesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-29X9Fuel 20ANF-87-1Revisio*~*~0.91:0.92:0.95:1.01:1.01:1.01:0.96:0.98:0.95*~*~*~0.92:0.94:0.98:0.97;1.05:0.95:0.99:0.95:0.98*~*~*~*0.95:0.98:0,93:1.06:1.05:1.06:1.05:0.97:0.96*~**1.01:0.97:1,06:1.03:1,03:1.04:1.07:1.06:1.02*~*~1.01:F05:1.05:1.03:0.00:1.01:1.07:1.06:1.01101'951.06:1.04:1.01:0F00:1.04:0.96:1.020.96:0.991.05:1.07:1.07:1.04:1.06:1.00:0.960.98:0.95:0.97:1.06:1.06:0.961.00:0.95:0.980.95:0.98:0.96:1.02:1.01:1.02:0.96:0.98:0.96Figure3.4DesignBasisLocalPowerDistribution AdvancedNuclearFuelsXN-19X9Fuel 21ANF-87-126 Revision1**~*~1.03:1.00:1.00:1.00:F00:1.00:1.01:1.03*~*:1.00:0.98:1.00*~*1.02:1.02:1.03:1.00:1.01*~*:1.00*~**~1.00:1.01:1.01:1.01:0.90:1.03:1.00:*:1.00:1.02*~*1.01:0.89:0.00:1.01:1.02:1.00*~*:1.00*~1.021.01:0.00:0.89:1.01:0.99:1.00*~*:1.00*~*1.03:0.901.01:1.01:0.98:1.00:1.001.01:1.00:1.03:1.020.99:1.00:0.98:1.001.03:1.01:1.00:1.00:1.001.00:1.00:1.03Figure3.5DesignBasisLocalPowerDistribution GeneralElectric(Central)
SXSRFuel 22ANF-87-'evisio*~1.00:1,00:1.00:1.00:1.00:1.00:1.00:1.00:*~0*~1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:*~**~*100*~**~1.00;1.00:1.00:1.00:1.00:1.00:1.00:*:1.00:1.00**~1.00:1.00:0.00:1.00:1.001.00:1.00:1.00:1.00:0.00:1.00:1.00:1.001.00**100*~*1.00:1.00:1.001.00:1.00:1.00:1.001.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00:1.00Figure3.6DesignBasisLocalPowerDistribution GeneralElectric(Peripheral)
SXSRFuel 23ANF-87-126 Revision1TABLE4.1NEUTRONIC DESIGNVALUESFuelPelletReference 9.10FuelRodReference 9.10FuelAssemblReference 9.10CoreDataNumberoffuelassemblies Ratedthermalpower,HWRatedcoreflow,Hlbm/hrCoreinletsubcooling, Btu/ibmHoderator temperature, FChannelthickness, inchFuelassemblypitch,inchWidewatergapthickness, inchNarrowwatergapthickness, inch'64329310024.0548.8.0806.000.5620.562ControlRodDataAbsorbermaterialTotalbladespan,inchTotalbladesupportspan,inchBladethickness, inchBladeface,-to-face internaldimension, inchAbsorberrodsperbladeAbsorberrodoutsidediameter, inchAbsorberrodinsidediameter, inchAbsorberdensity,%oftheoretical B4C9.751.580.2600.200760.1880.13870.0 24ANF-87-126 Revision**:LL:L:HL:M:N*H:HL:HL*~4HLM:MH:N*HHN*'HL*~*:HL*M".H*H:HO':HHH:HL**:H:HH:H:H:H*\H:HMH:N*:H:H:W:HH:H:HH*'A'~.*N:MH:HHNHW:MHH*HL:H*:MH:HHMH:MHHLHL:H:H:M~MH:H*:MMLMLL:HL:HL:HM:HL:HL-:LLLRODS(1)LRODS(5)HLRODS(16)HRODS(20)MHRODS(13)HRODS(15)H*RODS(9)WRODS(2)1.45W/0U2351.95W/0U2352.55W/0U2353.27W/0U2354.23W/0U2354.66W/0U2353.27W/0U235+4.00W/0GD203INERTWATERRODFigure4.1Susquehanna Unit2Cycle3Enrichment Distribution fortheANF92-344L-9G4 Xi4-2FuelLattice 25*************
9:*0**********
9:*0**********
ANF-87-126 Revision1*\*~LLoL:MLMMLNLL*~*~*~N*MLNH:M*:MH:M*ML'*.ML.N*M**H:HMHM:NL*~*o*~MHH:HH-H:HN*:M~JN*~HW:"MHH:MHt*'*~*~M:NH:HHMHW:NH:M*:MML:N~:MH:H:H:NH:NHM:MLML:M:N:M*:MH:M*:MMLML:ML:NL:MM:MLMLLLRODS(1)LRODS(5)MLRODS(16)MRODS(19)MHRODS(13)HRODS(15)M*RODS(10)WRODS(2)1.45W/0U2351.95W/0U2352.55W/0U2353.27W/0U2354,23W/0U2354.66W/0U2353.27'W/0U235+5.00W/0GD203INERTWATERRODigure4.2Susquehanna Unit2Cycle3Enrichment Distribution fortheANF92-344L-IOG5 XN-2FuelLattice 26ANF-87-12 RevisioA2C1A2C1A2C1A2C1DOC1A2CiEOC1A2C1DOC1DOC1A2C1DOC1A2C1FOC1C1A2A2C1DOA2DOCiDOA200C1~DOC1EOC1A2C100A2DOCi0000A200C1EOC1C1A2A2C1DOC1A2C1DOC1DOA2DOC1EOC1A2C1A2C1DOO'IC1A2DOA2EOEOCiC1A2A2C100C100A2C1C100C1EOC1EOC1A2C100A2DOCiDOC1EOC1EOC100C1A2DOC100A200A200C1EOC1EOA282C1A2DOA2EOC1EOC1C1C1A2A2A2Ci00DOEOC1EOC1A2C1EO80CiEOC1DOA2A2EOC1EOCiEOC1EOC182A2C1C1C1CiC1C1C1A2A2A2A2A2A2A2A2XY=FuelTypeXBurnedYCycles~FuelTeHo.ofBuouieoDescritionA8C'E196832414096GEBX8TypeIII2.19w/oU~235GEBX8TypeII1.76M/oU.235XN.1ENC92.3318.7G4 XN.2ANF92.333B.904 XN.2ANF92.3338.
ANF-87-126 Revision 1*\*~LL o L: ML M ML NL L*~*~*~N*ML NH: M*: MH: M*ML'*.ML.N*M**H: H MH M: NL*~*o*~MH H: H H-H: H N*: M~J N*~H W: "MH H: MH t*'*~*~M: NH: H H MH W: NH: M*: M ML: N~: MH: H: H: NH: NH M: ML ML: M: N: M*: MH: M*: M ML ML: ML: NL: M M: ML ML LL RODS (1)L RODS (5)ML RODS (16)M RODS (19)MH RODS (13)H RODS (15)M*RODS (10)W RODS (2)1.45 W/0 U235 1.95 W/0 U235 2.55 W/0 U235 3.27 W/0 U235 4,23 W/0 U235 4.66 W/0 U235 3.27'W/0 U235+5.00 W/0 GD203 INERT WATER ROD igure 4.2 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-IOG5 XN-2 Fuel Lattice 26 ANF-87-12 Revi sio A2 C1 A2 C1 A2 C1 A2 C1 DO C1 A2 Ci EO C1 A2 C1 DO C1 DO C1 A2 C1 DO C1 A2 C1 FO C1 C1 A2 A2 C1 DO A2 DO Ci DO A2 00 C1~DO C1 EO C1 A2 C1 00 A2 DO Ci 00 00 A2 00 C1 EO C1 C1 A2 A2 C1 DO C1 A2 C1 DO C1 DO A2 DO C1 EO C1 A2 C1 A2 C1 DO O'I C1 A2 DO A2 EO EO Ci C1 A2 A2 C1 00 C1 00 A2 C1 C1 00 C1 EO C1 EO C1 A2 C1 00 A2 DO Ci DO C1 EO C1 EO C1 00 C1 A2 DO C1 00 A2 00 A2 00 C1 EO C1 EO A2 82 C1 A2 DO A2 EO C1 EO C1 C1 C1 A2 A2 A2 Ci 00 DO EO C1 EO C1 A2 C1 EO 80 Ci EO C1 DO A2 A2 EO C1 EO Ci EO C1 EO C1 82 A2 C1 C1 C1 Ci C1 C1 C1 A2 A2 A2 A2 A2 A2 A2 A2 XY=Fuel Type X Burned Y Cycles~Fuel T e Ho.of Buouieo Descri tion A 8 C'E 196 8 324 140 96 GE BX8 Type III 2.19 w/o U~235 GE BX8 Type II 1.76 M/o U.235 XN.1 ENC92.3318.7G4 XN.2 ANF92.333B.904 XN.2 ANF92.3338.
10G5Figure4.3Susquehanna Unit2Cycle3Reference CoreI.eading 27ANF-87-126 Revision1120110~~~~~~~~~~~~~~~~~~~~~~~~h~~~~~~~~~~~~~~~~~~~1009080N70~~~~~~APPMRODBLOCK:eAPRNSCRAMLIN)~~~~~~4~~~~~~~~~~~~~~rr~/r~~r~/e''r/~/e/~~~~~~~~~~~~~~~~)~66/45)err~(~~/e100/vXeR00.'IN~4~~~~~~~~~$~~~ROOBiOCK~MONITOR80WE50~~ee~I~I~el'45K80KeCOREPLOWR00LINE403020e~~~~e~e~~~~~~~~~4'P~~~~~~~e~~~~~'I~~~~~~~~~~~~~~~~~~~~~4~~~~~~~4~~~~~~4~~~e10NTCIRC~~~M'-PUMP)NFLOW;00102030405060708090COREFLOW,%RATED100Figure4.4Susquehanna Unit.2Cycle3-CorePowervs.CoreFlow 28ANF-87-12''Revision595551261014182226303438424650545812--00--1220--26--26--205955514743----20--202000--12--08--12--0020474339--12--08--08-,-00--0808--123935----264444263531--00--04--00--00--0027----26444423--12--08--08--00*--'819----20--202004--002608--12203127191500--12--08--12--0020--26--26--2012--00--122610141822263034384246505458CycleExposureControlRodDensity0.0HHD/HTU23.3%ControlRodBeingWithdrawn
10G5 Figure 4.3 Susquehanna Unit 2 Cycle 3 Reference Core I.eading 27 ANF-87-126 Revision 1 120 110~~~~~~~~~~~~~~~~~~~~~~~~h~~~~~~~~~~~~~~~~~~~100 90 80 N 70~~~~~~APPM ROD BL OCK: e APRN SCRAM LIN)~~~~~~4~~~~~~~~~~~~~~r r~/r~~r~/e''r/~/e/~~~~~~~~~~~~~~~~)~66/45)e r r~(~~/e 100/v Xe R00.'IN~4~~~~~~~~~$~~~ROO BiOCK~MONITOR 80 W E 50~~e e~I~I~el'45K 80K e CORE PLOW R00 LINE 40 30 20 e~~~~e~e~~~~~~~~~4'P~~~~~~~e~~~~~'I~~~~~~~~~~~~~~~~~~~~~4~~~~~~~4~~~~~~4~~~e 10 NT CIRC~~~M'-PUMP)N FL OW;0 0 10 20 30 40 50 60 70 80 90 CORE FLOW,%RATED 100 Figure 4.4 Susquehanna Unit.2 Cycle 3-Core Power vs.Core Flow 28 ANF-87-12''Revision 59 55 51 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 12--00--12 20--26--26--20 59 55 51 47 43----20--20 20 00--12--08--12--00 20 47 43 39--12--08--08-,-00--08 08--12 39 35----26 44 44 26 35 31--00--04--00--00--00 27----26 44 44 23--12--08--08--00*--'8 19----20--20 20 04--00 26 08--12 20 31 2719 15 00--12--08--12--00 20--26--26--20 12--00--12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 Cycle Exposure Control Rod Density 0.0 HHD/HTU 23.3%Control Rod Being Withdrawn=00*Rod Fully Inserted=, 00 Rod Fully Withdrawn=--Figure 5.1 Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern 1.60 1.60 1.40 f4 1.30 O Note: The MCPR operating limit shall be the maximum of this curve, the full flow MCPR operating limit or the poorer dependent MCPR operating limit.A 1.80 O A 1.10 40 50 60 70 80 90 100 TOTAL CORE RECIRCULATION FLOW (%RATED)figure 5.2 Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit Ah+C" A-1 ANF-87-126 Revision 1 APPENDIX A U SINGLE LOOP OPERATION This Appendix provides limits and justification of those limits for Single Loop Operation (SLO).A.l ANTICIPATED OPERATIONAL OCCURRENCES Reference A.1 The NSSS supplier has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended V period of time.These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed.when oth recirculation systems are in oper ation.The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power.ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performance.
=00*RodFullyInserted=,00RodFullyWithdrawn
The ANF methodology has given results which are consistent with those of previous analyses for normal two-loop operation.
=--Figure5.1Susquehanna Unit2Cycle3ControlRodWithdrawal ErrorAnalysisLimitingInitialControlRodPattern 1.601.601.40f41.30ONote:TheMCPRoperating limitshallbethemaximumofthiscurve,thefullflowMCPRoperating limitorthepoorerdependent MCPRoperating limit.A1.80OA1.10405060708090100TOTALCORERECIRCULATION FLOW(%RATED)figure5.2Susquehanna Unit2Cycle3FlowMCPROperating Limit Ah+C" A-1ANF-87-126 Revision1APPENDIXAUSINGLELOOPOPERATION ThisAppendixprovideslimitsandjustification ofthoselimitsforSingleLoopOperation (SLO).A.lANTICIPATED OPERATIONAL OCCURRENCES Reference A.1TheNSSSsupplierhasprovidedanalyseswhichdemonstrate thesafetyofplantoperation withasinglerecirculation loopoutofserviceforanextendedVperiodoftime.Theseanalysesrestricttheoveralloperation oftheplanttolowerbundlepowerlevelsandlowernodalpowerlevelsthanareallowed.whenothrecirculation systemsareinoperation.Thephysicalinterdependence betweencorepowerandrecirculation flowrateinherently limitsthecoretolessthanratedpower.ANFfuelwasdesignedtobecompatible withtheco-residentfuelinthermalhydraulic, nuclear,andmechanical designperformance.
Many analyses performed by the NSSS supplier for single loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.For single loop operation, the NSSS vendor found that an increase of 0.01 in the HCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation.
TheANFmethodology hasgivenresultswhichareconsistent withthoseofpreviousanalysesfornormaltwo-loopoperation.
ANF has evaluated the effects of the increased flow measurement uncertainties on the safety limit HCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also applicable to ANF fuel during single loop operation.
Manyanalysesperformed bytheNSSSsupplierforsingleloopoperation arealsoapplicable tosingleloopoperation withfuelandanalysesprovidedbyANF.Forsingleloopoperation, theNSSSvendorfoundthatanincreaseof0.01intheHCPRsafetylimitwasneededtoaccountfortheincreased flowmeasurement uncertainties andincreased tipuncertainties associated withsinglepumpoperation.
Thus, increasing the safety limit HCPR by 0.01 for single loop operation (1.07)with ANF fuel is sufficiently onservative to also bound the increased flow measurement uncertainties for single loop operation.
ANFhasevaluated theeffectsoftheincreased flowmeasurement uncertainties onthesafetylimitHCPRandfoundthattheNSSSvendordetermined increaseintheallowedsafetylimitMCPRisalsoapplicable toANFfuelduringsingleloopoperation.
A-2 ANF-87-Revisio The limiting MCPR operating limit for single loop operation is conservatively set using the limiting pump seizure accident delta CPR plus the single loop operation HCPR safety limit.This limit together with the.HCPRf curve for two loop operation plus.Ol and the MCPRp curve for two loop operation plus.Ol conservatively bound all transients.
Thus,increasing thesafetylimitHCPRby0.01forsingleloopoperation (1.07)withANFfuelissufficiently onservative toalsoboundtheincreased flowmeasurement uncertainties forsingleloopoperation.
The Technical Specifications require APRH/LPRH surveillance to the left of the 45%Constant Flow line and above the 80%Rod Block line.Based on core hydrodynamic stability analyses for Cycle 3, operation at power/flow combinations above and to the left of the line connecting the 66%Power/45%Flow and 69%Power/47%Flow points needs to be added to the APRM/LPRM surveillance requirements.
A-2ANF-87-RevisioThelimitingMCPRoperating limitforsingleloopoperation isconservatively setusingthelimitingpumpseizureaccidentdeltaCPRplusthesingleloopoperation HCPRsafetylimit.Thislimittogetherwiththe.HCPRfcurvefortwoloopoperation plus.OlandtheMCPRpcurvefortwoloopoperation plus.Olconservatively boundalltransients.
Figure 4.4 shows the core power versus core flow established for Cycle 3.
TheTechnical Specifications requireAPRH/LPRH surveillance totheleftofthe45%ConstantFlowlineandabovethe80%RodBlockline.Basedoncorehydrodynamic stability analysesforCycle3,operation atpower/flow combinations aboveandtotheleftofthelineconnecting the66%Power/45%
A-3 ANF-87-126 Revision 1 A.2 POSTULATED ACCIDENTS Reference A.2 ANF performed LOCA analyses for single loop conditions and has determined that the MAPLHGR limit curve (Section 7.2)for two-loop operation is also applicable to single loop operation for ANF 9x9 fuels.
Flowand69%Power/47%
A-4 ANF-87-Revisio REFERENCES A.1"Susquehanna Unit 2 Cycle 2 Single Loop Operation Analysis," XN-NF-86-146, Advanced Nuclear Fuels Corporation; Richland, WA 99352, November 1986.A.2"Susquehanna LOCA Analysis for Single Loop Operation," XN-NF-86-125, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.
FlowpointsneedstobeaddedtotheAPRM/LPRM surveillance requirements.
B-1 ANF-87-126 Revision I APPENDIX B SEISMIC-LOCA EVALUATION The structural response of Advanced Nuclear Fuels Corporation's (ANF's)9x9 fuel is similar to the structural response of the GE BxBR fuel it replaces in the Susquehanna Unit 2 core.Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.The physical and structural properties of the 9x9 and the Bx8 fuel types which are important to the dynamic response of the fuel are summarized in Table B.l.he close agreement between the important parameters for the ANF 9x9 and GE x8R fuel types indicates that the structural response would be very similar for both fuel types.Similarity in the natural frequencies of the two fuel types mentioned above is further assured by the stiffness of the fuel assembly channel box.Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel.ANF calculations show that approximately 97%of the stiffness of a fuel assembly is attributable to the stiffness of the channel box.For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable.
Figure4.4showsthecorepowerversuscoreflowestablished forCycle3.
Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.
A-3ANF-87-126 Revision1A.2POSTULATED ACCIDENTS Reference A.2ANFperformed LOCAanalysesforsingleloopconditions andhasdetermined thattheMAPLHGRlimitcurve(Section7.2)fortwo-loopoperation isalsoapplicable tosingleloopoperation forANF9x9fuels.
B-2 ANF-87-Revisio TABLE B.1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8X8 AND 9X9 FUEL ASSEMBLIES
A-4ANF-87-RevisioREFERENCES A.1"Susquehanna Unit2Cycle2SingleLoopOperation Analysis,"
~Pro ert Assembly Weight, lbs Number of Spacers Overall Assembly Length, in Assembly Frequencies, cps Mode 1 2 3 5 6 7 ANF 9x9 580 171.29 1.9 3.7 6.5 10.4 15.5 21.9 29.1 Fuel T es GE 8x8R 600 171.40*GE proprietary ANF-87-126 Revision 1 Issue Date: 11/25/87 SUS(UEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Distribution:
XN-NF 146,AdvancedNuclearFuelsCorporation;
D.D.R.L.S.R.K.H.S.T.J.L.D.G.C.J.H.A.Adkisson J.Braun E.Collingham J.Federico F.Gaines G.Grummer D.Hartley J.Hibbard E.Jensen H.Keheley N.Morgan A.Nielsen F.Richey L.Ritter J.Volmer AD White E.Williamson H.G.Shaw/PP8L (20)Document Control (5)}}
: Richland, WA99352,November1986.A.2"Susquehanna LOCAAnalysisforSingleLoopOperation,"
XN-NF-86-125, AdvancedNuclearFuelsCorporation,
: Richland, WA99352,November1986.
B-1ANF-87-126 RevisionIAPPENDIXBSEISMIC-LOCAEVALUATION Thestructural responseofAdvancedNuclearFuelsCorporation's (ANF's)9x9fuelissimilartothestructural responseoftheGEBxBRfuelitreplacesintheSusquehanna Unit2core.Therefore, theseismic-LOCA structural responseevaluation performed insupportoftheinitialcoreremainsapplicable andcontinues toprovideassurance thatcontrolbladeinsertion willnotbeinhibited following theoccurrence ofthedesignbasisseismic-LOCA event.Thephysicalandstructural properties ofthe9x9andtheBx8fueltypeswhichareimportant tothedynamicresponseofthefuelaresummarized inTableB.l.hecloseagreement betweentheimportant parameters fortheANF9x9andGEx8Rfueltypesindicates thatthestructural responsewouldbeverysimilarforbothfueltypes.Similarity inthenaturalfrequencies ofthetwofueltypesmentioned aboveisfurtherassuredbythestiffness ofthefuelassemblychannelbox.Bothfueltypesusethesamefuelassemblychannelbox,andthechannelboxdominates theoveralldynamicresponseoftheincorefuel.ANFcalculations showthatapproximately 97%ofthestiffness ofafuelassemblyisattributable tothestiffness ofthechannelbox.Forthisreason,thedynamicstructural responseofthereloadcoreisessentially thatoftheinitialcore,andtheoriginalseismic-LOCA analysisremainsapplicable.
Deformation ofthechanneltothepointthatcontrolbladeinsertion isinhibited isnotpredicted tooccur.
B-2ANF-87-RevisioTABLEB.1COMPARISON OFPHYSICALANDSTRUCTURAL CHARACTERISTICS FOR8X8AND9X9FUELASSEMBLIES
~ProertAssemblyWeight,lbsNumberofSpacersOverallAssemblyLength,inAssemblyFrequencies, cpsMode123567ANF9x9580171.291.93.76.510.415.521.929.1FuelTesGE8x8R600171.40*GEproprietary ANF-87-126 Revision1IssueDate:11/25/87SUS(UEHANNA UNIT2CYCLE3RELOADANALYSISDesignandSafetyAnalysesDistribution:
D.D.R.L.S.R.K.H.S.T.J.L.D.G.C.J.H.A.AdkissonJ.BraunE.Collingham J.FedericoF.GainesG.GrummerD.HartleyJ.HibbardE.JensenH.KeheleyN.MorganA.NielsenF.RicheyL.RitterJ.VolmerADWhiteE.Williamson H.G.Shaw/PP8L (20)DocumentControl(5)}}

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Rev 1 to Susquehanna Unit 2 Cycle 3 Reload Analysis Design & Safety Analyses.
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Issue date: 11/30/1987
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ANF-87-1 26 REVIStON 1 AD~MHCSDo HUCIt.EAR FUSM CORPORATION SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS DESIGN AND SAFETY ANALYSES.NOVEMBER 198787i2310i58 87i223 POR ADOCK 0500058]~AN AFFII.IATE OF KRAF TWERK UNION Q~KRU ADVANCED NUCLEAR FUELS CORPORATION ANF-87-126 Revision 1 Issue Date: 11/25/87 SUSQUEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Prepared By: J.A.White BWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Services AII AFFILIATE OF KRAFTWERK UNION Qxsvu CUSTOMER DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY Advanced Nuclear Fuels Corporation's warranties and representations con-ceming the subject matter of this document are those set forth in the Agreement between Advanced Nuclear Fuels Corporation and the Customer pursuant to which this document is issued.Accordingly, except as otherwise expressly pro-vided In such Agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained In this document, or that the use of any information, apparatus, method or process disclosed ln this document will not infringe privately owned rights: or assumes any liabilities with respect to the use of any information, ap-paratus, method or process disclosed in this document.The information contained herein is for the sole use of Customer.In order to avoid Impairment of rights of Advanced Nuclear Fuels Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term)of such information until so authorized in writing by Advanced Nuclear Fuels Corporation or until after six (6)months following termination or expiration of the aforesaid Agreement and any extension thereof, unless otherwise expressly provided in the Agreement.

No rights or licenses In or to any patents are implied by the furnishing of this docu-ment.XN NF F00.765 (1 ANF-87-126 Revision 1 TABLE OF CONTENTS Section 1.0 2.0 Pacae INTRODUCTION.

~..............,....,................................

1 FUEL MECHANICAL DESIGN ANALYSIS...................................

2 3.0 3.2 3.2.1 3.2.3 3.2.5 THERMAL HYDRAULIC DESIGN ANALYSIS..............

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

3 Hydraul i c Characteri zati on........................................

3 Hydraul i c Compatibility...........................................

3 Fuel Centerline Temperature.......................................

3 B ypass Flowe~~~~~~~~~~~~~~~~~~~~~~~~\~~~~~~~~~~~~~~~~~~~~~~~~~~~~3 3.3 3.3.1 3.3.2.3.3 4.0 MCPR Fuel Cladding Integrity Safety Limit...........

Coolant Thermodynamic Conditions Design Basis Radial Power Distribution.

Design Basis Local Power Distribution.

NUCLEAR DESIGN ANALYSIS..

~~~~~~~~~0~~~~3 3~~~~~~~~~~~~~4~~~~~~~\~~~~~~~~~~~~~~~5 4.1 4.2 4.2.1 4.2.2 4.2.4 5.0 Fuel Bundle Nuclear Design Analysis.......

Core Nuclear Design Analysis Core Configuration.......

Core Reactivity Characteristics...,....,..

Core Hydrodynamic Stability.....

~.ANTICIPATED OPERATIONAL OCCURRENCES......,~~~~~~~~~~~~~~~~~~~~~~~~5~~~~~~~5 5 6~~~~~~~~~~~~~~~~~~~~~~~~7 5.1 5.2 5.3 5.4 5.5 5.6 5.7 6.0 6.1 F 1.1 itlons~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~7~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~8~~~~~~~~~~~~~~~~~~~~~~~~9~~~~~~~~~~~~~~~~~~~~~~~~9 10 10 Loss-Of-Coolant Accident....,,......

Break Location Spectrum........

10 Analysis Of Plant Transients At Rated Cond Analyses For Reduced Flow Operation.......

Analyses For Reduced Power Operation......

ASME Overpressurization Analysis..........

Control Rod Withdrawal Error (CRWE)Fuel Loading Error........

Determination Of Thermal Margins..........

POSTULATED ACCIDENTS...

ANF-87-1 Revision TABLE OF CONTENTS (Continued)

Section 6.1.2 6.1.3 6.2 7.0 7.1 T.1.1 7.1.2 7.2 7.2.1 7.2.2 7.2.3 73 7.3.1 7.3.2 8.0 Limiting Safety System Settings......

HCPR Fuel Cladding Integrity Safety L Steam Dome Pressure Safety Limit Limiting Conditions For Operation.

Average Planar Linear Heat Generation Minimum Critical Power Ratio~~~~~~~~~~~~~~~~~~~~~~~~~~~~~imit Rate imits.................

L'HGR Llml ts~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~o~~~~~~~Surveillance Requirements......

.....Scram Insertion Time Surveillance....

Stability Surveillance..........

METHODOLOGY REFERENCES..........

9.0 ADDITIONAL

REFERENCES.......

reak Size Spectrum...............................................

B APLHGR Analyses.............................'.....................

H Control Rod Drop Accident........,...

TECHNICAL SPECIFICATIONS........

Pacae 10 10 11 12 12 12 12 12 12 1 1 14 14 14 15 16 APPENDICES A.SINGLE LOOP OPERATION.............

A-1 B.SEISMIC-LOCA EVALUATION....,.................,.........,..........

B-1 ANF-87-126 Revision 1 LIST OF TABLES Table Pacae 4.1 Neutronic Design Values...........................................

23 B.1 Comparison Of Physical And Structural Characteristics For 8x8 And 9x9 Fuel Assemblies.........................

.....B-2 LIST OF FIGURES Ficiur e 3.1 Susquehanna Unit 2 Power vs.Flow....Cycle 3 Hydraulic Demand Curve Pacae 17 3.2 3.3 3.5 Susquehanna Unit 2 Design Basis Local Design Basis Local Design Basis Local Fuel.Cycle 3 Design Basis Radial Power..............

18 Power Distribution

-ANF XN-2 9x9 Fuel.........

19 Power Distribution

-ANF XN-1 9x9 Fuel.........

20 Power Distribution

-GE 8x8R (Central)21 3.6 4.1 4.2 4.3 4.4 5.1 5.2~~......22 24~~~~~~~~25~~~~~~~~~26 27 28 29 Design Basis Local Power Distribution

-GE (Peripheral) 8x8R Fuel..........

~~0~~~~Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-9G4 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Enrichment Distribution For ANF92-344L-10G5 XN-2 Fuel Lattice.Susquehanna Unit 2 Cycle 3 Reference Core Loading Plan...Susquehanna Unit 2 Cycle 3-Core Power vs.Core Flow......

Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern..Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit.......

~fj ANF-87-126 Revision 1

1.0 INTRODUCTION

This report provides the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF)*in support of the Cycle 3 reload for Susquehanna Unit 2, which is scheduled to commence operation in the spring of 1988.This report is intended to be used in conjunction with ANF topical report~XN-Np--191 A, 91 4, R 11 1, Nppti 1 1 1 1 N Company Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list.However, LHGR mechanical design limits (Reference 9.1)and plant transient simulation model developments (Reference 9.141 b 1 dbyANF b 4 t NRN P 1 F~F-Volume 4, Revi'sion 1.Both References 9.1 and 9.2 have been approved by the NRC for use in referencing in license applications.

Section numbers in this 9 t 1 9 dtd tt b 1 X-N--fNJ, olume 4, Revision 1.The Susquehanna Unit 2 Cycle 3 core will comprise a total of 764 fuel assemblies, including 236 unirradiated ANF XN-2 9x9 assemblies, 324 irradiated ANF XN-1 9x9 assemblies, 112 irradiated General Electric 8x8R fuel assemblies (central region), and 92 irradiated GE 8x8R assemblies in the peripheral region.The reference core configuration is described in Section 4.2.The design and safety analyses reported in this document were based on the design and operational assumptions in effect for Susquehanna Unit 2 during the previous operating cycle.Additional information and the results of design studies covering the development of 9x9 fuel assemblies for BWR reloads are contained in Reference 9.3.

f ANF-87-126 Revision 1 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable ANF Fuel Design Report: Reference 9.1 To assure that the expected power history for the fuels to be irradiated during Cycle 3 of Susquehanna Unit 2 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits (Figure 3.3 of Reference 9.1)have been specified.

In addition, an LHGR transient operating'imit for Anticipated Operating Occurrences (Figure 3.4 of Reference 9.1)has been specified for ANF 9x9 fuel.Additional information on rod bow, as requested in the NRC's safety evaluation report for Reference 9.1, has been transmitted in Reference 9.4.

ANF-87-126 Revision 1 3.0 THERMAL HYDRAULIC DESIGN ANALYSIS 3.2 H draul i c Char aeter i zat i on 3.2.1 H draulic Com atibilit Component hydraulic resistances for the constituent fuel types in the Susquehanna Unit 2 Cycle 3 core have been determined in single phase flow tests of full scale assemblies.

Figure 3.1 shows the hydraulic demand curves for ANF 9x9 fuel and GE 8x8R fuel in the Susquehanna Unit 2 core.The similar hydraulic performance indicates compatibility for co-residence in'he Susquehanna Unit 2 core.Applicable Generic Report 3.2'Fuel Centerline Tem erature~~Reference 9.1.2.2~21 Calculated Bypass Flow Fraction at 104%Power/100%

Flow 10.1%3.3 MCPR Fuel Claddin Inte rit Safet Limit Safety Limit MCPR=1.06 3.3.1 Coolant Thermod namic Condition Rated Thermal Power Feedwater Flowrate (at SLMCPR)Core Pressure (at SLMCPR)Feedwater Temperature 3293 Mwt 16.1 Mlbm/hr 1042.9 psia 383'F ANF-87-1 Revision 3.3.2 Desi n Basis Radial Power Distribution See Figure 3.2 3.3.3 Desi n Basis Local Power Distribution See Figures 3.3 through 3.6 ANF-87-126 Revision 1 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fue Bund e Nuclea Desi n Anal sis Assembly Average Enrichment Radial Enrichment Distribution Axial Enrichment Distribution Burnable Poisons Note: Burnabl e poi sons are distributed uniformly over the enriched length of the designated rods.The natural urania axial blanket sections do not contain burnable absorber material.Non-Fueled Rods Neutronic Design Parameters 3.33%Figure 4.1 and 4.2 Uniform 3.44%with 6" natural uranium top blanket Figure 4.1 and 4.2 Figure 4.1 and 4.2 Table 4.1 4.2 Core Nuclear Desi n Anal sis';2.1 Core Confi oration Figure 4.3 Core Exposure at EOC2, HWd/HTU Core Exposure at BOC3, MWd/HTU Core Exposure at EOC3, HWd/MTU Maximum Cycle 3 Licensing Exposure Limit, HWd/MTU 18350.7 10911.2 21740.8 22076 ANF-87-12 Revision 4.2.2 ore Reactiv't Characteris ics BOC Cold K-effective, All Rods Out BOC Cold K-effective, Strongest Rod Out 1.11353 0.98524 Reactivity Defect (R-Value)0.00%rho 4.2.4 Standby Liquid Control System Reactivity, Cold Conditions, 660 ppm I Core H drod namic Stabilit 0.98348 Power/flow Map Figure 4.4 Power Flow State Points 64/42*69/47**66/45**Deca Ratio COTRA 0.82 0.75 0.75*Two pump minimum flow-APRN Rod Block intercept point.Extended operation at lower flow is not allowed by Technical Specifications.

    • Operation at less than 45%flow requires APRH/LPRN surveillance.

In addition, operation at power/flow, combinations above and to the left of the line connecting these two points requires APRH/LPRtl surveillance.

See Figbre 4.4.

ANF-87-126 Revision 1 5.0 ANTICIPATED OPERATIONAL OCCURRENCES Applicable Generic Transient Analysis Methodology Report References 9.5 5 9.7 5.1 Anal sis Of Plant Transients At Rated Conditions Reference 9.6 Limiting Transient(s):

Load Rejection Without Bypass (LRWB)Feedwater Controller Failure (FWCF)Loss of Feedwater Heating (LFWH)Event LRWB Power*100%FWCF 100%100%1 16.8%LFWH 100%100%121'%233%123%1179 1078%Rated%Rated Maximum Maximum Maximum Pressure Flow Heat Flux Power ,~aia I 100%116.2%267%1194 Del ta CPR**0.24 0.23 0.16 Model COTRANSA/XCOBRA-T COTRANSA/XCOBRA-T PTSBWR3/XCOBRA Single Loop Operation:

Appendix A 5.2 Anal ses For Reduced Flow 0 eration Reference 9.6 Limiting Transient(s):

Recirculation Flow Increase Transient (RFIT)*104%power used in analysis as design bases.**Delta-CPR results for most limiting fuel type.

ANF-87-1 Revision 5.3 Anal ses For Reduced Power 0 eration Reference 9.6 Limiting Transient(s):

Feedwater Controller Failure (FWCF)%Power Transient Delta CPR ANF 9x9 GE 8x8R 104 80 65 40 FWCF FWCF FWCF FWCF 0.23 0.25 0.28 0.31 0.20 0.23 0.26 0.28 5.4 ASME Over ressurization Anal sis Reference 9.6 Limiting Event Worst Single Failure Maximum Pressure Maximum Steam Dome Pressure Full MSIV Isolation Direct Sera 1297 psig 1281 psig 5.5 Control Rod Withdrawal Error CRWE Starting Control Rod Pattern for Analysis Figure 5.1 Rod Block Settin 105 106*107 108*100%Flow Distance Withdrawn~ft 4.0 4.5 5.0 5.0 Delta CPR 0.22 0.24 0.26 0.26*Rod Block Monitor settings recommended for Cycle 3 operation.

ANF-87-126 Revision 1 5.6 Fuel Loadin Error Maximum Delta CPR 0.16 5.7 Determination Of Thermal Har ins Summary of Thermal Margin Requirements Event LRWB FWCF LFWH CRWE Power 1P0%**1PP%**1PP%9c*100%Flow 100%100%100%100%Delta CPR*0.24 0.23 0.16 0.24 at 106%RBH 0.26 at 108%RBM MCPR Limit 1.30 1.29 1.22 1.30 1.32 HCPR Operating Limits at Rated Conditions MCPR 0 eratin Limit 1.30 at 106%RBM 1.32 at 108%RBM Reduced Flow MCPR Limits Figure 5.2 Power Dependent HCPR Operating Limit Results for Cycle 3: 100*+/100 80/100 65/100 40/100 Limiting Transient LRWB FWCF FWCF FWCF ANF 9x9 1.30 1.31 1.34 1.37 GE 8x8R 1.27 1.29 1.32 1.34 ,i*Delta CPR results for most limiting fuel type.**104%power used in analysis as design bases.

10 ANF-87-126 Revi si on 1 6.0 POSTULATED ACCIDENTS 6.1 Loss-Of-Coolant Accident Sei smi c-LOCA: Appendix B 6.1.1 Break Location S ectrum Reference 9.8 6.1.2 Break Size S ectrum Reference 9.8 6.1.3 MAPLHGR Anal ses ANF 9x9 Fuel Reference 9.9 Limiting Break: Double-ended guillotine pipe break Recirculation pump discharge line 0.4 Discharge Coefficient Bundle Average Exposure GWD MTU 0 5 10 15 20 25 30 35 40 MAPLHGR~kw ft 10.2 10.2 10.2 10.2 10.2 9.6 8.9 8.2 7.5 Peak Clad Temperature*

~F 2060 2069 2121 2140 2147 2016 1839 1752 1676 Peak Local MWR**~Percent 3.9 3.7 3.7 4.8 5.2 2.7 1.0 0.7 0.5*Peak clad temperatures for XN-1 and XN-2 fuel are bounded by these results.**Metal Water Reaction.

ll ANF-87-1 Revision 6.2 Control Rod Dro Accident Section 8.0 Dropped Control Rod Worth, mk Doppler Coefficient, 1/k dk/dT Effective Delayed Neutron Fraction Four-Bundle Local Peaking Factor maximum Deposited Fuel Rod Enthalpy, cal/gm Number of Rods Exceeding 170 cal/gm 13.5-10.6 x(10)6 0.0058 1.34 205 (250 12 ANF-87-126 Revision 1 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Safet Limit Pressure Safety Limit (as measured in steam dome)1325 psig Analysis shows that a steam dome pressure safety limit of 1358 psig is allowed but the 1325 psig value used in Cycle 2 is to be conservatively retained.7.2 Limitin Conditions For 0 er ation 7.2.1 Avera e Planar Linear Heat Generation Rate Limits Bundle Average Exposure GWD MT 0 5 10 15 20 25 30 35 40 MAPLHGR Limits kw ft ANF 9x9 Fuel 10.2 10.2 10.2 10.2 10.2 9.6 8.9 8.2 7.5 13 ANF-87-1 Revision 7.2.2 Minimum Critical Power Ratio MCPR Operating Limits at Rated Conditions:

MCPR 0 eratin Limit 1.30 at 106%RBM 1.32 at 108%RBM MCPR Operating Limits at Off-Rated Conditions:

At Reduced Flow Figure 5.2 Total Core Recirculation Flow%Rated 100 96 92 83 76 60 50 40 Reduced Flow MCPR 0 eratin Limit 1.12 1.14 1.16 1,20 1.23 1.31 1.44 1.61 At Reduced Power Power Level%Rated 100*80 65 40 Reduced Power MCPR 0 eratin Limit 1.30 1.31 1.34 1.37*104%power used in analysis as design bases.

ANF-87-126 Revision I 7.2.3 LHGR Limits LHGR Limits Figures 3.3 and 3.4 of Reference 9.1 7.3 Surveillance Re uirements 7.3.1 Scram Insertion Time Surveillance Thermal limits established in Section 5.0 are based on minimum acceptable scram insertion performance as defined in the Technical Specifications.

No additional surveillance for scram insertion is required for validation of thermal limits..3.2 Stabilit Surveillance

~~Power/Flow Map Figure 4.4 The Unit 2 Cycle 2 Technical Specifications require APRM/LPRM surveillance to the left of the 45%Constant Flow line and above the 80%Rod Block line.Based on core hydrodynamic stability analyses, operation at power/flow combinations above and to the left of the line connecting the 66%Power/45%Flow and 69%Power/47%Flow points but below the APRM Rod Block line needs to be added to the APRM/LPRM surveillance requirement (see Section 4.2.4).

15 ANF-87-126 Revision 1 8.0 METHODOLOGY REFERENCES See XN-NF-80-19(P)(A), Volume 4, Revision 1 for complete bibliography.

\q,"~r 0~)6 44-I 16 ANF-87-126 Revision 1 9.0 ADDITIONAL REFERENCES 9.1"Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel,"~X---F,R.X,Addll 1 FXCF l,lhhl d, Washington, September 4, 1986.9.2"Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal 111<<N hd1 d P Revision 2, Advanced Nuclear Fuels Corporation, Richland, Washington, January, 1987.9.3"Demonstration of 9x9 Assemblies for BWRs," EPRI NP-3468, Electric Power Research Institute, Palo Alto, California, Hay 1, 1984.9.4 Letter, G.N.Ward (ANF)to G.C.Lainas (NRC),"Additional Information on Rod Bow," serial no.GNW:021:87, dated March 11, 1987.9.5"Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors,"~h-p--,h 11 X,AddN1F1C l,ltlhl d, Washington, November 16, 1981.~~~9.6"Susquehanna Unit 2 Cycle 3 Plant Transient Analysis," ANF-87-125, Rev.2, Advanced Nuclear Fuels Corporation, Richland, Washington, November 1987.9.7 9~8"XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core A 1 i,"X~, 1>>d 2, Advanced Nuclear Fuels Corporation, Richland, Washington, February 1987."Generic LOCA Break Spectrum Analysis BWR 3 8 4 with Hodified Low Pressure Coolant Injection Logic Using the EXEH Evaluation Model," XN-NF-~84-117 P, Advanced Nuclear Fuels Corporation, Richland, Washington, December 1984.9.9"Susquehanna LOCA-ECCS Analysis HAPLHGR Results for ENC 9x9 Fuel," XN-NF-86-65, Advanced Nuclear Fuels Corporation, Richland, Washington, May 1986.9.10"Principal Reload Fuel Design Parameters, Fuel Design, Susquehanna Unit 2 Reload XN-2," XN-NF-1058, Advanced Nuclear Fuels Corporation, Richland, Washington, March 1987, Formerly Exxon Nuclear Company.

Advanced Nuclear Fuels 9x9 0~)O~I General Electric 8x8R~o p co o C p 0 O I Q~o L I p 0 q>vCi I.>o oo I K 100.00.105.00 TIO.OO 115.M 120.00 125.00 QO.M Q5.M 140.00 Assembly Flow Rate, KLB/HR 0 CO%5.00 150.00 Figure 3.1 Susquehanna Unit 2 Cycle 3 Hydraulic Oemand Curve Power vs.Flow 80 70 60 50 00 C)CL So 20 10 0 0 0.2 0.0 0.6 0.8 1 1.2 RRDIFIL POHER PERKING Figure 3.2 Susquehanna 2 Cycle 3 Oesign Basis Radial Power 19 ANF-87-126 Revision 1*~: 0.88: 0.91: 0.96: 1.04: 1.02: 1.04: 0.96: 1.00: 0.96:*~**~: 0.91: 0.93: 0.98: 1.07: 0.91: 1.07: 0.97: 1.04: 1.01*~**~0.96: 0.98: 0.90: 1.04: 1.03: 1.04: 1.04: 0.99: 0.96:**~1.04: 1.07: 1.04: 1.00: 0.99: 1,00: 1.05: 0.94: 1.04**~*: 1.02:*~0.91: 1.03 0.99: 0.00: 0.98: 1.05: 1.07: 1.04*~*~1.04: 1.07: 1.04: 1.00: 0.98: 0.00: 1.03: 0.94: 1.05: 0.96: 0'7: 1'4: 1.05: 1.05: 1.03: 1.06: 1.00: 0,97 1.00: 1.04: 0.99: 0,94: 1.07: 0.94: 1.00: 0.94 1.01 0.96: 1.01: 0.96: 1.04: 1.04: 1.05: 0.97: 1.01 0.97 Figure 3.3 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-2 9X9 Fuel 20 ANF-87-1 Revisio*~*~0.91: 0.92: 0.95: 1.01: 1.01: 1.01: 0.96: 0.98: 0.95*~*~*~0.92: 0.94: 0.98: 0.97;1.05: 0.95: 0.99: 0.95: 0.98*~*~*~*0.95: 0.98: 0,93: 1.06: 1.05: 1.06: 1.05: 0.97: 0.96*~**1.01: 0.97: 1,06: 1.03: 1,03: 1.04: 1.07: 1.06: 1.02*~*~1.01: F 05: 1.05: 1.03: 0.00: 1.01: 1.07: 1.06: 1.011 01'95 1.06: 1.04: 1.01: 0 F 00: 1.04: 0.96: 1.02 0.96: 0.99 1.05: 1.07: 1.07: 1.04: 1.06: 1.00: 0.96 0.98: 0.95: 0.97: 1.06: 1.06: 0.96 1.00: 0.95: 0.98 0.95: 0.98: 0.96: 1.02: 1.01: 1.02: 0.96: 0.98: 0.96 Figure 3.4 Design Basis Local Power Distribution Advanced Nuclear Fuels XN-1 9X9 Fuel 21 ANF-87-126 Revision 1**~*~1.03: 1.00: 1.00: 1.00: F 00: 1.00: 1.01: 1.03*~*: 1.00: 0.98: 1.00*~*1.02: 1.02: 1.03: 1.00: 1.01*~*: 1.00*~**~1.00: 1.01: 1.01: 1.01: 0.90: 1.03: 1.00:*: 1.00: 1.02*~*1.01: 0.89: 0.00: 1.01: 1.02: 1.00*~*: 1.00*~1.02 1.01: 0.00: 0.89: 1.01: 0.99: 1.00*~*: 1.00*~*1.03: 0.90 1.01: 1.01: 0.98: 1.00: 1.00 1.01: 1.00: 1.03: 1.02 0.99: 1.00: 0.98: 1.00 1.03: 1.01: 1.00: 1.00: 1.00 1.00: 1.00: 1.03 Figure 3.5 Design Basis Local Power Distribution General Electric (Central)SXSR Fuel 22 ANF-87-'evisio*~1.00: 1,00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*~0*~1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*~**~*100*~**~1.00;1.00: 1.00: 1.00: 1.00: 1.00: 1.00:*: 1.00: 1.00**~1.00: 1.00: 0.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00: 0.00: 1.00: 1.00: 1.00 1.00**100*~*1.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00: 1.00 Figure 3.6 Design Basis Local Power Distribution General Electric (Peripheral)

SXSR Fuel 23 ANF-87-126 Revision 1 TABLE 4.1 NEUTRONIC DESIGN VALUES Fuel Pellet Reference 9.10 Fuel Rod Reference 9.10 Fuel Assembl Reference 9.10 Core Data Number of fuel assemblies Rated thermal power, HW Rated core flow, Hlbm/hr Core inlet subcooling, Btu/ibm Hoderator temperature, F Channel thickness, inch Fuel assembly pitch, inch Wide water gap thickness, inch Narrow water gap thickness, inch'64 3293 100 24.0 548.8.080 6.00 0.562 0.562 Control Rod Data Absorber material Total blade span, inch Total blade support span, inch Blade thickness, inch Blade face,-to-face internal dimension, inch Absorber rods per blade Absorber rod outside diameter, inch Absorber rod inside diameter, inch Absorber density,%of theoretical B4C 9.75 1.58 0.260 0.200 76 0.188 0.138 70.0 24 ANF-87-126 Revision**: LL: L: HL: M: N*H: HL: HL*~4 HL M: MH: N*HH N*'HL*~*: HL*M".H*H: H O': HH H: HL**: H: HH: H: H: H*\H: H M H: N*: H: H: W: HH: H: HH*'A'~.*N: MH: H H NH W: MH H*HL: H*: MH: H H MH: MH HL HL: H: H: M~MH: H*: M ML ML L: HL: HL: H M: HL: HL-: L LL RODS (1)L RODS (5)HL RODS (16)H RODS (20)MH RODS (13)H RODS (15)H*RODS (9)W RODS (2)1.45 W/0 U235 1.95 W/0 U235 2.55 W/0 U235 3.27 W/0 U235 4.23 W/0 U235 4.66 W/0 U235 3.27 W/0 U235+4.00 W/0 GD203 INERT WATER ROD Figure 4.1 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-9G4 Xi4-2 Fuel Lattice 25*************

9:*0**********

ANF-87-126 Revision 1*\*~LL o L: ML M ML NL L*~*~*~N*ML NH: M*: MH: M*ML'*.ML.N*M**H: H MH M: NL*~*o*~MH H: H H-H: H N*: M~J N*~H W: "MH H: MH t*'*~*~M: NH: H H MH W: NH: M*: M ML: N~: MH: H: H: NH: NH M: ML ML: M: N: M*: MH: M*: M ML ML: ML: NL: M M: ML ML LL RODS (1)L RODS (5)ML RODS (16)M RODS (19)MH RODS (13)H RODS (15)M*RODS (10)W RODS (2)1.45 W/0 U235 1.95 W/0 U235 2.55 W/0 U235 3.27 W/0 U235 4,23 W/0 U235 4.66 W/0 U235 3.27'W/0 U235+5.00 W/0 GD203 INERT WATER ROD igure 4.2 Susquehanna Unit 2 Cycle 3 Enrichment Distribution for the ANF92-344L-IOG5 XN-2 Fuel Lattice 26 ANF-87-12 Revi sio A2 C1 A2 C1 A2 C1 A2 C1 DO C1 A2 Ci EO C1 A2 C1 DO C1 DO C1 A2 C1 DO C1 A2 C1 FO C1 C1 A2 A2 C1 DO A2 DO Ci DO A2 00 C1~DO C1 EO C1 A2 C1 00 A2 DO Ci 00 00 A2 00 C1 EO C1 C1 A2 A2 C1 DO C1 A2 C1 DO C1 DO A2 DO C1 EO C1 A2 C1 A2 C1 DO O'I C1 A2 DO A2 EO EO Ci C1 A2 A2 C1 00 C1 00 A2 C1 C1 00 C1 EO C1 EO C1 A2 C1 00 A2 DO Ci DO C1 EO C1 EO C1 00 C1 A2 DO C1 00 A2 00 A2 00 C1 EO C1 EO A2 82 C1 A2 DO A2 EO C1 EO C1 C1 C1 A2 A2 A2 Ci 00 DO EO C1 EO C1 A2 C1 EO 80 Ci EO C1 DO A2 A2 EO C1 EO Ci EO C1 EO C1 82 A2 C1 C1 C1 Ci C1 C1 C1 A2 A2 A2 A2 A2 A2 A2 A2 XY=Fuel Type X Burned Y Cycles~Fuel T e Ho.of Buouieo Descri tion A 8 C'E 196 8 324 140 96 GE BX8 Type III 2.19 w/o U~235 GE BX8 Type II 1.76 M/o U.235 XN.1 ENC92.3318.7G4 XN.2 ANF92.333B.904 XN.2 ANF92.3338.

10G5 Figure 4.3 Susquehanna Unit 2 Cycle 3 Reference Core I.eading 27 ANF-87-126 Revision 1 120 110~~~~~~~~~~~~~~~~~~~~~~~~h~~~~~~~~~~~~~~~~~~~100 90 80 N 70~~~~~~APPM ROD BL OCK: e APRN SCRAM LIN)~~~~~~4~~~~~~~~~~~~~~r r~/r~~r~/er/~/e/~~~~~~~~~~~~~~~~)~66/45)e r r~(~~/e 100/v Xe R00.'IN~4~~~~~~~~~$~~~ROO BiOCK~MONITOR 80 W E 50~~e e~I~I~el'45K 80K e CORE PLOW R00 LINE 40 30 20 e~~~~e~e~~~~~~~~~4'P~~~~~~~e~~~~~'I~~~~~~~~~~~~~~~~~~~~~4~~~~~~~4~~~~~~4~~~e 10 NT CIRC~~~M'-PUMP)N FL OW;0 0 10 20 30 40 50 60 70 80 90 CORE FLOW,%RATED 100 Figure 4.4 Susquehanna Unit.2 Cycle 3-Core Power vs.Core Flow 28 ANF-87-12Revision 59 55 51 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 12--00--12 20--26--26--20 59 55 51 47 43----20--20 20 00--12--08--12--00 20 47 43 39--12--08--08-,-00--08 08--12 39 35----26 44 44 26 35 31--00--04--00--00--00 27----26 44 44 23--12--08--08--00*--'8 19----20--20 20 04--00 26 08--12 20 31 2719 15 00--12--08--12--00 20--26--26--20 12--00--12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 Cycle Exposure Control Rod Density 0.0 HHD/HTU 23.3%Control Rod Being Withdrawn=00*Rod Fully Inserted=, 00 Rod Fully Withdrawn=--Figure 5.1 Susquehanna Unit 2 Cycle 3 Control Rod Withdrawal Error Analysis Limiting Initial Control Rod Pattern 1.60 1.60 1.40 f4 1.30 O Note: The MCPR operating limit shall be the maximum of this curve, the full flow MCPR operating limit or the poorer dependent MCPR operating limit.A 1.80 O A 1.10 40 50 60 70 80 90 100 TOTAL CORE RECIRCULATION FLOW (%RATED)figure 5.2 Susquehanna Unit 2 Cycle 3 Flow MCPR Operating Limit Ah+C" A-1 ANF-87-126 Revision 1 APPENDIX A U SINGLE LOOP OPERATION This Appendix provides limits and justification of those limits for Single Loop Operation (SLO).A.l ANTICIPATED OPERATIONAL OCCURRENCES Reference A.1 The NSSS supplier has provided analyses which demonstrate the safety of plant operation with a single recirculation loop out of service for an extended V period of time.These analyses restrict the overall operation of the plant to lower bundle power levels and lower nodal power levels than are allowed.when oth recirculation systems are in oper ation.The physical interdependence between core power and recirculation flow rate inherently limits the core to less than rated power.ANF fuel was designed to be compatible with the co-resident fuel in thermal hydraulic, nuclear, and mechanical design performance.

The ANF methodology has given results which are consistent with those of previous analyses for normal two-loop operation.

Many analyses performed by the NSSS supplier for single loop operation are also applicable to single loop operation with fuel and analyses provided by ANF.For single loop operation, the NSSS vendor found that an increase of 0.01 in the HCPR safety limit was needed to account for the increased flow measurement uncertainties and increased tip uncertainties associated with single pump operation.

ANF has evaluated the effects of the increased flow measurement uncertainties on the safety limit HCPR and found that the NSSS vendor determined increase in the allowed safety limit MCPR is also applicable to ANF fuel during single loop operation.

Thus, increasing the safety limit HCPR by 0.01 for single loop operation (1.07)with ANF fuel is sufficiently onservative to also bound the increased flow measurement uncertainties for single loop operation.

A-2 ANF-87-Revisio The limiting MCPR operating limit for single loop operation is conservatively set using the limiting pump seizure accident delta CPR plus the single loop operation HCPR safety limit.This limit together with the.HCPRf curve for two loop operation plus.Ol and the MCPRp curve for two loop operation plus.Ol conservatively bound all transients.

The Technical Specifications require APRH/LPRH surveillance to the left of the 45%Constant Flow line and above the 80%Rod Block line.Based on core hydrodynamic stability analyses for Cycle 3, operation at power/flow combinations above and to the left of the line connecting the 66%Power/45%Flow and 69%Power/47%Flow points needs to be added to the APRM/LPRM surveillance requirements.

Figure 4.4 shows the core power versus core flow established for Cycle 3.

A-3 ANF-87-126 Revision 1 A.2 POSTULATED ACCIDENTS Reference A.2 ANF performed LOCA analyses for single loop conditions and has determined that the MAPLHGR limit curve (Section 7.2)for two-loop operation is also applicable to single loop operation for ANF 9x9 fuels.

A-4 ANF-87-Revisio REFERENCES A.1"Susquehanna Unit 2 Cycle 2 Single Loop Operation Analysis," XN-NF-86-146, Advanced Nuclear Fuels Corporation; Richland, WA 99352, November 1986.A.2"Susquehanna LOCA Analysis for Single Loop Operation," XN-NF-86-125, Advanced Nuclear Fuels Corporation, Richland, WA 99352, November 1986.

B-1 ANF-87-126 Revision I APPENDIX B SEISMIC-LOCA EVALUATION The structural response of Advanced Nuclear Fuels Corporation's (ANF's)9x9 fuel is similar to the structural response of the GE BxBR fuel it replaces in the Susquehanna Unit 2 core.Therefore, the seismic-LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertion will not be inhibited following the occurrence of the design basis seismic-LOCA event.The physical and structural properties of the 9x9 and the Bx8 fuel types which are important to the dynamic response of the fuel are summarized in Table B.l.he close agreement between the important parameters for the ANF 9x9 and GE x8R fuel types indicates that the structural response would be very similar for both fuel types.Similarity in the natural frequencies of the two fuel types mentioned above is further assured by the stiffness of the fuel assembly channel box.Both fuel types use the same fuel assembly channel box, and the channel box dominates the overall dynamic response of the incore fuel.ANF calculations show that approximately 97%of the stiffness of a fuel assembly is attributable to the stiffness of the channel box.For this reason, the dynamic structural response of the reload core is essentially that of the initial core, and the original seismic-LOCA analysis remains applicable.

Deformation of the channel to the point that control blade insertion is inhibited is not predicted to occur.

B-2 ANF-87-Revisio TABLE B.1 COMPARISON OF PHYSICAL AND STRUCTURAL CHARACTERISTICS FOR 8X8 AND 9X9 FUEL ASSEMBLIES

~Pro ert Assembly Weight, lbs Number of Spacers Overall Assembly Length, in Assembly Frequencies, cps Mode 1 2 3 5 6 7 ANF 9x9 580 171.29 1.9 3.7 6.5 10.4 15.5 21.9 29.1 Fuel T es GE 8x8R 600 171.40*GE proprietary ANF-87-126 Revision 1 Issue Date: 11/25/87 SUS(UEHANNA UNIT 2 CYCLE 3 RELOAD ANALYSIS Design and Safety Analyses Distribution:

D.D.R.L.S.R.K.H.S.T.J.L.D.G.C.J.H.A.Adkisson J.Braun E.Collingham J.Federico F.Gaines G.Grummer D.Hartley J.Hibbard E.Jensen H.Keheley N.Morgan A.Nielsen F.Richey L.Ritter J.Volmer AD White E.Williamson H.G.Shaw/PP8L (20)Document Control (5)