The ABWR PlantGeneral Description

GE Hitachi Nuclear Energy

ABWRPlant General Description

7.1.2007

DISCLAIMER OF RESPONSIBILITY

This document was prepared by the GE Hitachi Nuclear Energy (GEH) only for the purpose of providing general information about its Advanced Boiling Water Reactor (ABWR). No other use, direct or indirect, of the document or the information it contains is authorized; and with respect to any unauthorized use, neither GEH nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy, or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information. Furnishing this document does not convey any license, express or implied, to use any patented information or any information of GEH disclosed herein, or any rights to publish or make copies of the document without prior written permission of GEH.

Contents

Contents v

Acronyms ..................................................................................................................v

Introduction 1-1Chapter 1 Introducton 1-1

Nuclear Energy for the New Mllennum ............................................................... 1-1Forty Years n the Makng ..................................................................................... 1-1ABWR Development and Desgn Objectves ........................................................ 1-4ABWR Projects Worldwde ................................................................................... 1-4Nuclear Plant Projects n the New Mllennum ...................................................... 1-6So, how does GEs ABWR measure up? .............................................................. 1-7

Chapter 2 Plant Overvew 2-1Summary of the ABWR Key Features .................................................................. 2-1

Nuclear Island 3-1Chapter 3 Nuclear Boler Systems 3-1

Overvew ............................................................................................................... 3-1Reactor Vessel and Internals ............................................................................... 3-1Recrculaton System ............................................................................................ 3-7Control Rod Drve System .................................................................................. 3-12Man Steam System............................................................................................ 3-15Feedwater System (Nuclear Island) ................................................................... 3-19

Chapter 4 Safety Systems 4-1Overvew ............................................................................................................... 4-1Emergency Core Coolng Systems ....................................................................... 4-3

Hgh Pressure Core Flooder............................................................................ 4-3Reactor Core Isolaton Coolng ....................................................................... 4-4Automatc Depressurzaton System ............................................................... 4-5Resdual Heat Removal .................................................................................. 4-5

Standby Gas Treatment System ........................................................................... 4-7Atmospherc Control System ................................................................................ 4-7Flammablty Control System ................................................................................ 4-8Standby Lqud Control System ............................................................................ 4-9

Contents

Emergency Desel Generator System ................................................................ 4-10

Chapter 5 Auxlary Systems 5-1Reactor Auxlary Systems ................................................................................... 5-1Reactor Water Cleanup System ........................................................................... 5-1Fuel Pool Coolng and Cleanup and Suppresson Pool Cleanup System ............ 5-3Reactor Buldng Coolng Water System/Reactor Buldng Servce Water System 5-4Drywell Coolng System ........................................................................................ 5-5Radwaste .............................................................................................................. 5-5

Lqud Radwaste Management System ........................................................... 5-6Offgas System ................................................................................................. 5-8Sold Radwaste Management System ........................................................... 5-9

Chapter 6 Fuel Desgn 6-1Introducton and Summary.................................................................................... 6-1Core Configuration................................................................................................ 6-2Fuel Assembly Descrpton ................................................................................... 6-2Control Rod Descrpton........................................................................................ 6-5Core Orificing ........................................................................................................ 6-7Other Reactor Core Components ......................................................................... 6-7Core Nuclear Desgn ............................................................................................ 6-8Neutron Montorng System ................................................................................ 6-10

Chapter 7 Instrumentaton and Control 7-1Overvew ............................................................................................................... 7-1Dgtal Protecton System Applcatons ................................................................. 7-2

Reactor Protecton System.............................................................................. 7-3Leak Detecton and Isolaton System .............................................................. 7-4

Fault-Tolerant Process Control Systems .............................................................. 7-4Automatc Power Regulator System................................................................ 7-5Feedwater Control System .............................................................................. 7-5Steam Bypass and Pressure Control System ................................................. 7-6Recrculaton Flow Control System ................................................................. 7-6Turbne Control System ................................................................................... 7-7Power Generaton Control System .................................................................. 7-7Rod Control and Informaton System .............................................................. 7-7Process Radaton Montorng System ............................................................ 7-8Area Radaton Montorng System .................................................................. 7-8Contanment Atmospherc Montorng System ................................................ 7-8Process Computer........................................................................................... 7-9Remote Shutdown System .............................................................................. 7-9

Man Control Room ............................................................................................... 7-9Plant Automaton................................................................................................. 7-12Operaton ............................................................................................................ 7-13

Chapter 8 Plant Layout and Arrangement 8-1Plant Layout .......................................................................................................... 8-1

Contents

Reactor Buldng ................................................................................................... 8-3Prmary Contanment System ............................................................................... 8-7Renforced Concrete Contanment Vessel Descrpton ......................................... 8-9Fre Protecton .................................................................................................... 8-10Flood Protecton.................................................................................................. 8-10Other Buldngs ....................................................................................................8-11

Balance of Plant 9-1Chapter 9 Major Balance of Plant Features 9-1

Steam and Power Converson System ................................................................. 9-1Other Turbne Auxlary Systems .......................................................................... 9-5Staton Electrcal Power ........................................................................................ 9-5

Evaluations 10-1Chapter 10 Safety Evaluatons 10-1

Lcensng Framework.......................................................................................... 10-1Safety Desgn Approach ..................................................................................... 10-1Desgn Bass Transent and Accdent Performance ............................................ 10-3Severe Accdent Mtgaton ................................................................................. 10-3

Contanment Overpressure Protecton System ............................................. 10-5Summary ............................................................................................................ 10-5

Chapter 11 Plant Operatons 11-1Basc BWR Operaton ..........................................................................................11-1Operatng Map .....................................................................................................11-1Plant Startup and Shutdown ................................................................................11-2Automatc Load- Followng Capablty..................................................................11-3Automated Response to Desgn Bass Accdents ................................................11-3Flexblty n Fuel Cycle Length ............................................................................11-3Technical Specifications.......................................................................................11-4Emergency Plant Operaton.................................................................................11-4Summary .............................................................................................................11-4

Appendices A-1Appendx A Key Desgn Characterstcs A-1

Appendx B Frequently Asked Questons B-1What has the ABWR done to reduce the potental for Intergranular Stress Corroson Crackng?..............................................................................................................B-1Can the reactor vessel and attached ppng really last 60 years? ........................B-3What has ABWR done to address worker radaton exposure? ............................B-4

IndexI-1

Acronyms

v

AcronymsABWR AdvancedBoilingWaterReactorACS AtmosphericControlSystemADS AutomaticDepressurizationSystemAIWA AC-IndependentWaterAdditionSystemALARA AsLowAsReasonablyAchievableALWR AdvancedLightWaterReactorAPR AutomaticPowerRegulatorSystemAPRM AveragePowerRangeMonitorARD Anti-ReverseRotationDeviceARI AlternateRodInsertionARM AreaRadiationMonitoringASD AdjustableSpeedDriveASME AmericanSocietyofMechanicalEngineersAST AlternateSourceTermATIP AutomaticTraversingIn-CoreProbeATLM AutomaticThermalLimitMonitorATP AuthorizationtoProceedATWS AnticipatedTransientsWithoutScram

B&V BlackandVeatchBAF BottomofActiveFuelBOP BalanceofPlantBWR BoilingWaterReactor

C&I ControlandInstrumentationCAM ContainmentAtmosphericMontioringSystemCB ControlBuildingCCC ControlCellCoreCCFP ContingentContainmentFailureProbabilityCDF CoreDamageFrequencyCFS CondensateandFeedwaterSystemCO CommercialOperationCOE CoostofElectricityCP ConstructionPermitCPR CriticalPowerRatioCRD ControlRodDriveCRDHS ControlRodDriveHydraulicSystemCRT CathodeRayTubeCSP CondensateStoragePoolCST CondensateStorageTank

CTG CombustionTurbineGeneratorCWS CirculatingWaterSystem

DAT DesignAcceptanceDBA DesignBasisAccidentDCPS DCPowerSupplyDCV DrywellConnectingVentDG DieselGeneratorDMC DigitalMeasurementControllerDPS DiverseProtectionSystemDRM DryRadwasteManagementSystemDW DrywellDWC DrywellCooling

ECCS EmergencyCoreCoolingSystemECP ElectrochemicalPotentialECW EmergencyChilledWaterEDG EmergencyDieselGeneratorEHC Electro-hydraulicControl(TurbineControl

System)EMI Electro-MagneticInterferenceEMS EssentialMultiplexingSystemEPD ElectricalPowerDistributionEPRI ElectricPowerResearchInstituteESF EssentialSafeguardsFeatureEPRI ElectricPowerResearchInstitute

FCS FlammabilityControlSystemFDA FinalDesignApprovalFFTR FinalFeedwaterTemperatureReductionFIV Flow-InducedVibrationFMCRD FineMotionControlRodDriveFOAKE First-of-a-KindEngineeringFP FireProtectionFPCU FuelPoolCoolingandCleanupFSAR FinalSafetyAnalysisReportFSC FirstStructuralConcreteFTDC FaultTolerantDigitalControllerFW FeedwaterFWP FeedwaterPumpFWC FeedwaterControlSystem

GE GeneralElectricCompany

Acronyms

v

GETAB GeneralElectricThermalAnalysisBasisGPM Gallonsperminute

HCU HydraulicControlUnitHCW High-ConductivityWasteHEPA HighEfficiencyParticulateAirHFF HollowFiberFIlterHIC HighIntegrityContainerHPCF HighPressureCoreFlooderHPCP HighPressureCondensatePumpHPIN HighPressureNitrogenGasSupplyHVAC Heating,VentilationandAir-ConditioningHWC HydrogenWaterChemistryI&C InstrumentationandControlIASCC Irradiation-AssistedStressCorrosion

CrackingIGSCC IntergranularStressCorrosionCrackingILRT IntegratedLeakRateTestIMS InformationManagementSystemIRM IntermediateRangeMonitorISI In-ServiceInspectionITAAC Inspection,Test,Analysisand AcceptanceCriteria

LCW LowConductivityWasteLD LowerDrywellLDI LeakDetectionandIsolationSystemLHGR LinearHeatGenerationRateLLRT LocalLeakRateTestLOCA Loss-of-CoolantAccidentLOFW LossofFeedwaterLOOP LossofOffsitePower LOPP LossofPreferredPowerLPCI Low-PressureCoolantInjectionLPCP Low-PressureCondensatePumpLPCRD LockingPistonControlRodDriveLPFL LowPressureFlooderLPRM LocalPowerRangeMonitorLRM LiquidRadwasteManagementSystemLTP LowerTiePlate

MCC MainControlConsole/MotorControlCenter

MCES MainCondenserEvacuationSystemMCPR MinimumCriticalPowerRatioMCR MainControlRoomM-G Motor-GeneratorMITI MinistryofInternationalTradeand

Industry(Japan)

MLHGR MaximumLinearHeatGenerationRateMMI Man-MachineInterfaceMOV Motor-OperatedValveMRBM Multi-ChannelRodBlockMonitoring

SystemMS MainSteamSystemMSIV MainSteamIsolationValveMSR MoistureSeparatorReheaterMUW MakeupWaterSystemMUX MultiplexerMWB MakeupWaterBuilding

NBS NuclearBoilerSystemNCW NormalChilledWaterNDT NilDuctilityTemperatureNEMS Non-EssentialMultiplexingSystemNMS NeutronMonitoringSystemNRC NuclearRegulatoryCommissionNRHX Non-RegenerativeHeatExchangerNSS NuclearSteamSupply

O&M OperationandMaintenanceOG OffgasSystemOPRM OscillationPowerRangeMonitor

PCI PelletCladInteractionPCS PlantComputerSystemPCT PeakFuelCladTemperaturePCV PrimaryContainmentVolumePG PowerGeneration(loads)PGCS PowerGenerationControlSystemPIP PlantInvestmentProtection(loads)PIP PositionIndicatorProbePLR PartLengthFuelRodPRA ProbabilisticRiskAssessmentPRM ProcessRadiationMonitoringSystemPRNM PowerRangeNeutronMonitorSystemPWR PressurizedWaterReactor

RAT ReserveAuxiliaryTransformerRB ReactorBuildingRBC RodBrakeControllerRCCV ReinforcedConcreteContainmentVesselRCW ReactorBuildingCoolingWaterSystemRCIS RodControlandInformationSystemRCIC ReactorCoreIsolationSystemRCPB ReactorCoolantPressureBoundaryRFC RecirculationFlowControlSystemRHR ResidualHeatRemovalRHX RegenerativeHeatExchanger

Acronyms

v

RIP ReactorInternalPumpRMC RecirculationMotorCoolingRMISS RecirculationMotorInflatableShaft

SealRMP RecirculationMotorPurgeRMU RemoteMultiplexerUnitRPS ReactorProtectionSystemRPV ReactorPressureVesselRRCS RedundantReactivityControlSystem RSS RemoteShutdownSystemRSW ReactorBuildingServiceWaterSystemRTNDT ReferenceNilDuctilityTemperatureRWCU ReactorWaterCleanupSystemRWM RodWorthMinimizerS&PC SteamandPowerConversionSystemSA SevereAccidentSAR SafetyAnalysisReportSBO StationBlackoutSBPC SteamBypassandPressureControl

SystemSCC StressCorrosionCrackingSCRRI SelectControlRodRun-inSDV ScramDischargeVolumeSGTS StandbyGasTreatmentSystemSJAE SteamJetAirEjectorSHE StandardHydrogenElectrodeSLCS StandbyLiquidControlSystemSOE SequenceofEventsSP SuppressionPoolSPCU SuppressionPoolCleanupSystemSPDS SafetyParameterDisplaySystemSRM SourceRangeMonitorSRNM StartupRangeNeutronMonitorSRV Safety/ReliefValve SSAR StandardSafetyAnalysisReportSSE SafeShutdownEarthquake

SSLC SafetySystemLogicandControlSW ServiceWater

TAF TopofActiveFuelTCW TurbineBuildingCoolingWaterSystemTBS TurbineBypassSystemTBV TurbineBypassValveTCCWS TurbineComponentCoolingWater

SystemTCS TurbineControlSystemTCV TurbineControlValveTCS TurbineControlSystemTEPCO TokyoElectricPowerCompanyTGSS TurbineGlandSteamSystemTIP TraversingIn-CoreProbeTIU TechnicianInterfaceUnitTMSS TurbineMainSteamSystem TPC TaiwanPowerCompanyTRA TransientRecordingandAnalysisTSC TechnicalSupportCenterTSW TurbineBuildingServiceWaterSystem

UAT UnitAuxilliaryTransformerUD UpperDrywellUHS UltimateHeatSinkUPS UninterruptablePowerSupplyURD UtilityRequirementsDocumentUTP UpperTiePlate

V&V VerificationandValidationVAC Volts-AlternatingCurrentVB VacuumBreakerVDC Volts-DirectCurrent

WDP WideDisplayPanelWRNM WideRangeNeutronMonitoringSysteWW Wetwell

Chapter 1 Introduction

1-1

Nuclear Energy for the New Millennium

Nuclear energy plays amajor role in meet-ingtheworldsenergyneeds.Attheendof2005,therewere443nuclearpowerplantsoperatingin32countries.Theseplantsaccountfor17%oftheworldselectricity.Theindustryremainsdynamic,asevidencedbythefactthatseveralnewplantsentercommercialoperationeveryyearandthereare,typi-cally,30ormoreinvariousstagesofconstructionatanygiventime.

Generating electricity with nuclear energypermits economic and social development tobesustainable;thatis,notlimitedbyencroachingenvi-ronmentalconcerns.Anon-nuclear,baseloadpowerplantgenerateselectricitybyburning fossil fuelsdayinanddayoutandreleasingtheby-productstotheenvironment.Anuclearplant,ontheotherhand,generateslargeamountsofelectricitywithvirtuallynoimpactontheenvironment.Inquantitativeterms,iftheworldsnuclearplantswerereplacedwithcoal-firedplants,globalCO2emissionswouldincreaseby8%everyyear.Thiswouldamountto1600milliontonsperyearatatimewhentheworldistryingtoreduceemissionsby4200million tonsperyear.Similarly,iftheworldsgrowingappetitefornewelectricityismetwithoutnuclearenergyplayingakeyrole,CO2emissionswouldquicklyrisetolevelsthatcurtaileconomicgrowth.

TheAdvancedBoilingWaterReactor(ABWR)advancednuclearplantwillplayanimportantroleinmeetingtheconflictingneedsofdevelopedanddevelopingeconomiesformassiveamountsofnewelectricity and theneedworldwide to limitCO2

emissions.FourABWRshavebeenconstructedinJapanandarereliablygeneratinglargeamountsoflowcostelectricity.TaiwanisconstructingtwomoreABWRswhichwillentercommercialoperationin2009and2010.Othercountrieshavesimilarstrate-giestodeployadvancednuclearplants,andthesuc-cessfuldeploymentofABWRsinJapanandTaiwan,coupledwithinternationalagreementstolimitCO2emissions,willonlyreinforcetheseplans.

TheABWRrepresentsanentirelynewapproach

tothewaynuclearplantprojectsareundertaken.TheABWRwaslicensedanddesignedindetailbeforeconstructioneverbegan.Onceconstructiondidbe-gin,itproceededsmoothlyfromstarttofinishinjustfouryears.Capitalcostsamountedto$1600/kWatwhichlevelnuclearisverycompetitivewithotherformsofpowergeneration.

Thesuccessfuldesign,licensing,constructionandoperationof theABWRnuclearpowerplantushersinaneweraofsafe,economicandenviron-mentally friendlynuclear electricity.TheABWRis thefirstof anewgenerationofnuclearplantsequippedwithadvancedtechnologiesandfeaturesthatraiseplantsafetytonewlevelsthatsignificantlyimprovetheeconomiccompetitivenessofthisformofgeneration.

Forty Years in the MakingThe BoilingWater Reactor (BWR) nuclear

plant,likethePressurizedWaterReactor(PWR),hasitsoriginsinthetechnologydevelopedinthe1950sfortheU.S.Navysnuclearsubmarineprogram.ThefirstBWRnuclearplanttobebuiltwasthe5MWe

1ChapterIntroduction

Chapter 1 Introduction

1-2

Vallecitosplant(1957)locatednearSanJose,Cali-fornia.TheVallecitosplantconfirmedtheabilityoftheBWRconcepttosuccessfullyandsafelyproduceelectricity for agrid.Thefirst large-scaleBWR,Dresden1(1960),thenfollowed.TheBWRdesignhassubsequentlyundergoneaseriesofevolutionarychangeswithonepurposeinmindsimplify.

TheBWRdesignhasbeensimplified in twokeyareasthe reactor systemsand the contain-mentdesign.Table1-1chroniclesthedevelopmentoftheBWR.

Dresden1was,interestinglyenough,notatrueBWR.Thedesignwasbasedupondualsteamcycle,notthedirectsteamcyclethatcharacterizesBWRs.Steamwasgeneratedinthereactorbutthenflowed

toanelevatedsteamdrumandasecondarysteamgeneratorbeforemaking itsway to the turbine.ThefirststepdownthepathofsimplicitythatledultimatelytotheABWRwastheeliminationoftheexternalsteamdrumbyintroducingtwotechnicalinnovationstheinternalsteamseparatoranddryer(KRB,1962).Thispracticeofsimplifyingthedesignwithtechnicalinnovationswastoberepeatedoverandover.

ThefirstlargedirectcycleBWRs(OysterCreek)appearedinthemid-1960sandwerecharacterizedbytheeliminationofthesteamgeneratorsandtheuseoffiveexternalrecirculationloops.Later,reactorsystemswerefurthersimplifiedbytheintroductionof internal jet pumps.Thesepumps sufficientlyboostedrecirculationflowsothatonlytwoexternal

Product Frst Commercal Representatve Plant/ Lne Operaton Date Characterstcs

BWR/1 1960 Dresden 1 Intal commercal-sze BWR BWR/2 1969 Oyster Creek Plants purchased solely on economcs Large drect cycle BWR/3 1971 Dresden 2 Frst jet pump applcaton Improved ECCS: spray and flood capability

BWR/4 1972 Vermont Yankee Increased power densty (20%) BWR/5 1977 Toka 2 Improved ECCS Valve flow control

BWR/6 1978 Cofrentes Compact control room Sold-state nuclear system protecton system

ABWR 1996 Kashwazak-Karwa 6 Reactor nternal pumps Fne-moton control rod drves Advanced control room, dgtal sold-state mcroprocessors Fber optc data transmsson / multplexng Increased number of fuel bundles Ttanum condenser Improved ECCS: high/low pressure flooders

Table 1-1. Evolution of the GE BWR

Chapter 1 Introduction

1-3

recirculationloopswereneeded.ThischangefirstappearedintheDresden-2BWR/3plant.

TheuseofreactorinternalpumpsintheABWRdesignhas taken thisprocessof simplification toits logical conclusion.Byusingpumps attacheddirectlytothevesselitself,thejetpumpsandtheexternalrecirculationsystems,withalltheirpumps,valves,piping,andsnubbers,havebeeneliminatedaltogether.ThisdesignfeatureisthesourceofmanyoftheABWRssafetyandoperationaladvantages.Figure1-1 illustrates theevolutionof the reactorsystemdesign.

recirculation loops, that allows the containment(and,byextension,thereactorbuilding)tobemorecompact.

TheMarkIcontainmentwasthefirstofthenewcontainmentdesigns.ThetorususedtohousealargewaterinventoryintheMarkIgivesthisdesignitscharacteristiclightbulbconfiguration.TheconicalMarkIIdesignhasaless-complicatedarrangement,basedonsteel-linedreinforcedconcrete.Akeyfea-tureisthelargecontainmentdrywellthatprovidesmore roomfor the steamandECCSpiping.TheMarkIIIcontainmentdesign,usedworldwidewithBWR/6sandsomeBWR/5s,representedamajorimprovement in simplicity. Its steel containmentstructureisarightcircularcylinderthatiseasytoconstruct,andprovidesreadyaccesstoequipmentandamplespaceformaintenanceactivities.OtherfeaturesoftheMarkIIIincludehorizontalventstoreduceoverall loss-of-coolant accident (LOCA)dynamicloadsandafree-standingall-steelstructuretoensureleak-tightness.

TheABWRcontainmentissignificantlysmallerthantheMarkIIIcontainmentbecausetheelimina-tionoftherecirculationloopstranslatesintoasig-nificantlymorecompactcontainmentand reactorbuilding.ThestructureitselfismadeofreinforcedconcretewithasteellinerfromwhichitderivesitsnameRCCV,orreinforcedconcretecontainmentvessel.Figure1-2 illustrates theevolutionof theBWRcontainment from the earliest versions totodaysABWRRCCVdesign.Where the reactorbuildingisalsoshown,thecontainmentisoutlined

Figure 1-1 Evolution of the Reactor System Design

Figure 1-2. Evolution of BWR Containment

ThefirstBWRcontainmentswere sphericaldrystructures,similartothosestillusedtodayinPWRdesigns.TheBWR,however,quicklymovedtothepressuresuppressioncontainmentdesignforitsmanyadvantages.Amongtheseare:

HighheatcapacityLowerdesignpressureSuperiorabilitytoaccommodaterapiddepres-surizationUniqueabilitytofilterandretainfissionprod-uctsProvisionofalargesourceofreadilyavailablemakeupwaterinthecaseofaccidentsSimplified,compactdesign

Itisthereductionincontainmentdesignpres-sures,togetherwiththeeliminationoftheexternal

Dresden 1

ABWR

KRB

Dresden 2Oyster Creek

DRY

MARK I

MARK II

MARK III

ABWR

Chapter 1 Introduction

1-4

inred.

Thereare93BWRs,includingfourABWRs,currentlyoperatingworldwide.Manyareamongthebestoperatingplantsintheworld,performinginthebestofclasscategory.NumerouscountriesrelyheavilyuponBWRplantstomeettheirneedsfor electricity. Japan, for example,has32BWRplants,representingnearlytwo-thirdsofitsinstallednuclearcapacity.TheTokyoElectricPowerCom-pany(TEPCO)owns17nuclearplants,allofwhichareBWRs.TEPCOsKashiwazaki-Kariwanuclearstation,whichconsistsofseven(7)largeBWRs,isthelargestpowergenerationfacilityintheworld,licensedfor8,200MWe.Similarly,BWRplantspre-dominateinTaiwanandseveralEuropeancountries.IntheUnitedStates,thereare37operatingBWRs.

Todate,theABWRplantistheonlyadvancednuclearplantinoperationorunderconstruction.

ABWR Development and Design Objectives

Developmentof theABWR tookplacedur-ingthe1980sunderthesponsorshipoftheTokyoElectric Power Company (TEPCO).The statedpurposeofthedevelopmenteffortwastodesignaBWRplantthatincludedacarefulblendof(1)thebest featuresofworldwideoperatingBWRs, (2)availablenewtechnologies,and(3)newmodularconstruction techniques. Safety improvementswere,asalways,thetoppriority.Anticipatingtheeconomicchallengesthatlayahead,specialatten-tionwaspaidtosystematicallyreducingthecapitalcostandincorporatingfeaturesintotheplantdesignthatwouldmakemaintenancesignificantlyeasierandmoreefficient.

DevelopmentoftheABWRoccurredinaseriesofsteps.Phase1wasaconceptualdesignstudythatdetermined the feasibilityof theABWRconcept.Phase2,inwhichmostofthedevelopmentworktookplace,includedmoredetailedengineeringandthetestingofnewtechnologiesanddesignfeatures.ThepurposeofPhase3was toput thefinishing

toucheson thedesignand systematically reducecapitalcosts,whichprovedtobeahighlysuccessfuland,inhindsight,fortuitousendeavor.Thedevelop-mentphasescametoanendin1988whenTEPCOannouncedthatthenextKashiwazaki-KariwaunitstobeconstructedwouldbeABWRs.

WiththeselectionoftheABWRfortheK-6&7project,thedetailed,orproject,engineeringbegan.Licensingactivitieswith the Japanese regulatoryagency,MITI(MinistryofInternationalTradeandIndustry),alsostartedatthistimeand,interestingly,wereconductedinparallelforsometimewiththereviewof theABWRin theU.S.by theNuclearRegulatory Commission (NRC). MITI and theNRC,infact,heldseveralmeetingstodiscusstheirrespectivereviews.

By1991, thedetaileddesignwasessentiallycompleteandMITIconcludeditslicensingreview.AnEstablishmentPermit, or license,was issuedinMay1991.ExcavationbeganlaterthatyearonSeptember17,bringingadecadeofdevelopmentworktoasuccessfulconclusion.

Developmentofanadvancednuclearplantisamajorendeavor.ThedevelopmentoftheABWRspannedadecadeandcost anestimated$500M.Such an enterprise can only be undertaken incooperationwithmanyotherorganizations.TheABWRwasdevelopedbyGEincooperationwithits technical associatesHitachiLtd. andToshibaCorp.The sponsorship andguidanceofTEPCOwas instrumental.TheABWRdevelopment alsoreceivedfinancialsupportfromtheotherJapaneseutilitiesthatoperateBWRs,aswellasfromsixteenU.S.utilities.

ABWR Projects World-wideOperating ABWRs in Japan

FourABWRunitsinJapanarenowconstructedandfullyoperational.TwooftheseunitsarelocatedatTEPCOsKashiwazaki-Kariwa site100milesnorthofTokyoontheSeaofJapan.Theworldsfirst

Chapter 1 Introduction

1-5

advancednuclearplant,Unit6,begancommercialoperationonNovember7,1996.Unit7,thesecondABWR,followedshortlythereafterwithcommercialoperationcommencingonJuly2,1997.

BothABWRunitswereconstructedinworldrecord times.Fromfirst concrete to fuel load, ittookjust36.5monthstoconstructUnit6and38.3monthsforUnit7,theformerbeing10monthslessthanthebesttimeachievedforanyofthepreviousBWRsconstructedinJapan.Inaddition,bothunitswerebuiltonbudget,whichisanimpressiverecordofperformance, since thesewerefirst-of-a-kindunits.

TwomoreABWRsarenowoperationalinJapan-Hamaoka-5,whichbegancommercialoperationinJanuary,2005;andShika-2,whichwasconnectedtothegridinJuly,2005,andachievedcommercialoperationinMarch,2006.

BothTEPCOunitshavecompletedmanycyclesofoperation.Byallmeasures,theseABWRshavelivedup to their promise.Other than regulatorymandatedoutages,bothplanthaveoperatedessen-tiallyatfullpowerforeachfuelcycle.Thethermalefficiencyoftheplantis35%,slightlyhigherthanpreviousdesigns.SeeFigure1-3foraphotooftheKashiwazakiUnits6&7.

as themost exhaustive, andperhaps exhausting,revieweverundertakenbytheU.S.NuclearRegula-toryCommission.TheeffortsoftheNRCandGEcametofruitiononMay2,1997whenthenChairoftheNRC,Ms.ShirleyJackson,approvedandsignedtheABWRDesignCertificationintolaw.ThiswasrightlyhailedbytheU.S.industryasasignificantaccomplishment,onethathasbeenenvisionedforalongtimepre-approvalofastandarddesignofanadvancednuclearplant.SeeFigure1-4forarepro-ductionoftheABWRDesignCertification.

Figure 1-3. Kashiwazaki Units 6 & 7

Figure 1-4. ABWR Design Certification

The ABWR in the United StatesThelicensingoftheABWRhasbeendescribed

The successes continued when theABWRFirst-of-a-KindEngineering(FOAKE)programwascompletedinSeptember1996tothepraiseandsatis-factionoftheutilitysponsors.FOAKEisanequallysignificantaccomplishmentbecauseitrepresentsamajorsteptowardtheU.S.industrysothergoaltohavea(pre-licensed)designthatis90%engineeredpriortothestartofconstruction.AttheconclusionoftheFOAKEprogram,approximately65%oftheengineeringoftheU.S.versionoftheABWRwascomplete.TheremainingengineeringisbeingdoneaspartoftheLungmenproject,describedbelow.

The ABWR in TaiwanTwomoreABWRsarebeingconstructedforthe

TaiwanPowerCompany(TPC)atTPCsLungmensite,locatedonthePacificOceanabout40milesnortheastofTaipei.

CommercialoperationofLungmenUnit1isexpectedtobegininJuly2009.ThescheduleforUnit2,includingthestartofcommercialoperation,isoneyearlater.

Chapter 1 Introduction

1-6

Nuclear Plant Projects in the New Millennium

Theway inwhichABWRnuclearplantsaredesigned,licensedandconstructedisvastlydifferentthanwasthecase10or20 yearsago.

Design and Licensing TheABWRnuclearplant is licensedandde-

signedinitsentiretypriortothestartofconstruction.Today,longbeforefirstconcreteispoured,allsafetyandengineeringissuesareidentifiedandresolved.Thisprecludesconstructiondelaysduetore-engi-neering,aproblemwhichplaguedsomanyprojectsinthepastandcontributedsignificantlytothehigh(andinsomecasesmind-numbing)capitalcosts.

TheABWRhasbeendesignedtohigherlevelsofsafety,includingbeingdesignedtopreventandmitigate theconsequencesof aSevereAccident.Licensing documents approved by the USNRCindicatethatevenintheeventofasevereaccident,therewouldbenoreleaseofradioactivematerialtothepublic.

Todaysnuclearplantsareextensivelyandex-haustivelyreviewedbymultipleregulatorybodies.Infact,theABWRhasbeenreviewedandapprovedin threecountries (Japan,U.S.andTaiwan).ThisensuresthatthelicensingoftheABWRwillproceedonasmoothandtimelybasisinothercountriesthatchoosetodeployanABWR.

TheABWR design has been captured elec-tronicallyusingthelateststate-of-the-artinforma-tionmanagementtechnologycalledPOWRTRAK.Thebenefitsappearnotonlyinconstruction,whereithasbeenshownoverandoverwithfossilplantsthatuseofthisengineeringtoolreducesconstruc-tiontimeandcost,butalsoduringtheoperationandmaintenanceoftheplant.POWRTRAKisbotha3Dmodeldesigntoolandanextensivedatabaseforplantequipmentandmaterials.

Theapproachdescribedabove isbeing fullyutilizedfortheLungmenproject.ThedesignandlicensingoftheseABWRsareproceedingsmoothlyasexpected.

Construction of Nuclear Plants in the 2000sNuclearplantstodayareconstructedmuchdif-

ferentlythaninthepast.Themostnotabledifferenceistheschedule.TheABWRcanbebuiltinonlyfouryears,fromfirstconcretetothestartofcommercialoperation.Designsimplificationsandtheuseofnewconstructiontechnologiesandtechniquesmakethispossible.

Today,theplantownerissparedtheconcernforscheduledelaysandcostoverruns.Supplierscommittoafixedscheduleandprice,largelybecausethedesignhasbeenpre-licensedandpre-engineered.

Ofcourse,thereisnosubstituteforexperience.TheLungmenABWRsarebeingsuppliedbyateamofU.S.andJapanesesuppliers,ledbyGE,thatwerealsoinvolvedinthesupplyoftheJapaneseABWRs.Thisteamandthesupportingnetworkofequipmentsub-suppliersisaccustomedtoworkingonaninter-nationalstageandcanreadilytransplantitsexperi-enceandknow-howtoanewhostcountry.Thisisthebasisforthelearningcurveeffect,whichreducescapitalcostswitheachnewunit.

ABWR is Accumulating Operating ExperienceTheABWRsinJapanhavenowaccumulated

manyyearsofoperatingexperienceafterhavingcompleted a highly successful construction ef-fort.This representsawealthof informationandknow-howthatisofbenefittosubsequentABWRprojects.

Business Risks of the ABWRItisimportanttoscrutinizeaproposednew

nuclear project in terms of the business risks itcarries.Aplantdesignand itssuppliercontributesignificantlytotheownersabilitytomanagethoserisks.Therisksthatneedtobeconsideredarethoseassociatedwith:themarketforelectricity,thelicens-ingandconstructionoftheplantitself,theoperationofthenewfacility,thetechnicalaspectsoftheplantdesignandfinancingtheproject.

Inlightofthesemanyrisks,whatcharacteristicsshouldapotentialownerlookforinadesignandinasupplier?

Licensingrisk-can theplantbe licensedona

Chapter 1 Introduction

1-7

reasonableandpredictablescheduleorwillthisbecomealengthyeffortthatseriouslyeffectsthedateofcommercialoperation?Engineering risk-is theplant fullydesignedbeforeconstructionstartsorwilltherebenewsurprisesthatresultincostlydesignchangesandconstruction slowdownswhen it is de-signedduringthecourseoftheproject?Technology risk-will the plant perform asexpectedorwillsomeunknowntechnicalprob-lemkeeptheplantshutdownandthusunabletomeetitsrevenueprojections?Costrisk-will theplantcostmorethanbud-geted,threateningitsabilitytocompeteintheopenmarket?Schedule risk-will theplant startgeneratingelectricity-andrevenue-asplannedorwilltheschedulebecomeprotracted leading to costincreases?Financingrisk-willthenewplanthavepredict-ablerevenuesandcostsoristhereuncertaintyandlackofconfidencebylendersandinvestorsforthenewproject?

So, how does GEs ABWR measure up? Licensing Risk

TheABWRhasbeendesignedtothehigheststandardsofsafetyandhasalreadyreceivedaDe-signCertification.Moreover,theABWRhasbeenlicensedinJapan,wheretwoABWRshaveoper-atedsafelyandsuccessfullyfornearlytenyears,andinTaiwan.ThissuggeststousthattheprocessforreviewingaCOLbaseduponanABWRprojectshouldreasonablytakeoneyearonly.

Engineering RiskTheABWRisafullydesignedplantcomplete

withequipmentandmanufacturingdrawings.Ma-terials,quantitiesandcostsarepreciselyknown.Thismeanstherewillbenomajorsurprisesduringconstructionthatwouldcreatecostlydelaysandre-designs.ThisisalsothebasisuponwhichGEisabletoofferafirmpriceforitsscopeofsupply,eliminatingthatelementofrisk.

Technology RiskTheABWRistheonlyadvancednuclearplant

beingofferedthathasactuallybeenconstructed.Infact,twoABWRunitsinJapanhaveacombinedtenyearsofoperationalexperience.Itsowner,theTokyoElectricPowerCompany,haspublishedinformationthatshowstheseplantshavemetorexceededallofitsdesignandperformancegoalsandno technicalproblemsordesignflawshavesurfaced.Thisisdirectlyattributabletothe$500M,fiveyear test anddevelopmentprogram jointlyundertakenbyTEPCO,theotherJapaneseutilitiesthatownBWRs,GE,Hitachi andToshiba.TheTEPCOunitsareenjoyinghighavailabilityandcapacityfactors.

Cost and Schedule RiskGEhaseliminated this risk fornewowners

byoffering theABWRwithboth afixedpriceandconstructionschedule,thesamebasisfortheJapaneseandTaiwanABWRprojects.Thisreflectsthe confidence that comeswith the experienceofhavingdeliveredfourABWRunitsonastrictbudgetandschedule.

Financing RiskObtainingfinancingforanewnuclearproject

issometimesperceivedasasignificantobstacletonewconstruction.

ThisperceptionisbaseduponbadexperiencesinthepastwiththeU.S.beingacaseinpoint.TheexperienceofWallStreetduring the late1970sand early 1980s was that it took 15 years andseveralbilliondollarstocompleteanuclearplant.Suchperceptionstendtolingereventhoughtodaynuclearplantsare routinelybuiltonbudgetandschedule.

Lendersandinvestors,however,arepracticalpeople.Theywillfinanceaproject, includinganuclearproject,iftheycanbeassuredofareturnontheirinvestment.Investorslookforasolidprojectpro forma sheetandthismeansthatthepotentialownermustbeabletodemonstratepersuasivelythat revenues andproject costs arepredictable.Awell-managedutility that hasover time suc-cessfullyoperatednuclearplants(avoidingsafetylapses,achievinghighcapacityfactors,controllingcosts)assuresthefinancialcommunitythattargets

Chapter 1 Introduction

1-8

forrevenuesandon-goingoperationalcostswillbeachieved.

Likewise, a well-managed engineering andconstructionteamthathasexperiencebuildingad-vancednuclearplantsassuresinvestorsthatcapitalcostsprojectionswillnotbeexceededandthattheplantwillbefreeoftechnicalproblems.TheGEled

teambringsthiskindofexperienceandmanagementtoaproject.Thisactuallyhastheeffectoffurtherreducing the owners capital costs for the verygoodreasonthatstrongprojectfundamentalsleadtomorefavorablefinancing.Areductionofeven0.25%intheinterestrateorafavorablechangeindebt-to-equityrequirementscansavetensofmil-lionsofdollars.

Chapter 2 Plant Overview

2-1

2ChapterPlantOverviewThekeydesignobjectivesfortheABWRwere

establishedduringthedevelopmentprogram.Thekeygoals,allofwhichwereachieved,areasfol-lows:

Designlifeof60years.Plantavailabilityfactorof87%orgreater.Lessthanoneunplannedscramperyear.18to24-monthrefuelinginterval.Operatingpersonnel radiationexposure limit

Chapter 2 Plant Overview

2-2

Feature ABWR BWR/6

Recrculaton Vessel-mounted reactor nternal Two external loop Recrc system pumps wth jet pumps nsde RPV

Control Rod Drves Fne-moton CRDs Lockng pston CRDs

ECCS 3-dvson ECCS 2-dvson ECCS plus HPCS

Reactor Vessel Extensve use of forged rngs Welded plate

Prmary Contanment Advanced - compact, nerted Mark III - large, low pressure, not nerted

Secondary Contanment Reactor Buldng Sheld, fuel, auxlary & DG buldngs

Control & Instrumentation Digital, multiplexed, fiber optics, Analog, hardwired, single multple channel channel

Control Room Operator task-based System-based

Severe Accident Mitigation Inerting, drywell flooding, Not specifically addressed contanment ventng

Reactor Water Cleanup 2%, sealless pumps n cold leg 1%, pumps n hot leg

Offgas Passve offgas wth room- Actve offgas wth chlled temperature charcoal charcoal filters

Table 2-1. Comparison of Key ABWR Features to a BWR/6

penetrationsbelowthetopofthecoreelevation,andmakepossibleasmallerEmergencyCoreCoolingSystem(ECCS)networktomaintaincorecoverageduringpostulatedloss-of-coolantevents.

TheABWRECCSnetworkwasdesignedasafullthree-division*system,withbothahighandlowpressureinjectionpumpandheatremovalcapabilityineachdivision.Fordiversity,oneofthesystems,theReactorCore IsolationCooling (RCIC)Sys-tem,includesasteam-driven,highpressurepump.Transient response was improved by designing

threeavailablehigh-pressure injectionsystems inadditiontofeedwater.Theadoptionofthreeon-siteemergencydiesel-generatorstosupportcorecoolingandheatremoval,aswellastheadditionofanon-sitegasturbine-generator,reducesthepotentialforStationBlackout(SBO).ThebalancedECCSsystemhaslessrelianceontheAutomaticDepressurizationSystem(ADS)function,sinceasingle,motor-drivenhighpressurecoreflooder(HPCF)canmaintaincoresafetyforanypostulatedpipebreak.

Response toAnticipatedTransientsWithoutScram (ATWS) is improved by the adoption offine-motioncontrolroddrives(FMCRDs),whichallow reactor shutdown either by hydraulic orelectric insertion. In addition, theneed for rapidoperatoractiontomitigateanATWSisavoidedbyautomationofemergencyproceduressuchasfeed-

*Thetermdivisionmeansthatallsystemsandsupportsystems necessary to complete the safety functionare contained within the division, and that divisionisphysicallyseparatedfromotherdivisions toavoidanypropagatingfailures,suchasthreatsduetofireorflood.

Chapter 2 Plant Overview

2-3

Figure 2-2 ABWR Major Systems

CondensateStorage

BoratedWater

SuppressionPools (SP)

FuelPool

StandbyLiquid

ControlSystem

AC IndependentWater Addition

HeatExchanger

RuptureDisc

RuptureDisc

ContainmentOverpressure

ProtectionSystem(COPS)

FuelPool

Cooling/CleanupSystem

HydraulicControl Unit

FMCRD

High PressureTurbine

Low Pressure Turbines

CirculatingWater

Generator

CRD Pump

To MainCondenser Gland Steam

CondenserCBP

FeedwaterPump

HighPressureFeedwaterHeaters

MoistureSeparatorReheaterSRV

HP HeaterDrain Tank

LowPressureFeedwaterHeaters

CondensatePump

CondensateFilter

CondensateDemin

Steam JetAir Ejector

ReactorPressure

Vessel

(RPV)

QuencherQuencher

F/D

Hx

SPCU

HPCF C

(Hx)

Vent toStack

Vent toStack

OffgasSystem

Main Condensers

Reactor WaterCleanup System

F/DNon-Regen Hx

RegenerativeHx

SP

RIPRIP

SP

HPCF B

RHR A

RHR B

RCIC

Hx

Hx

RHR C

2-4

Chapter 2 Plant Overview

Chapter 2 Plant Overview

2-5

waterrunbackandStandbyLiquidControlSystem(SLCS)injection.

CalculatedcoredamagefrequencyisreducedbymorethanafactoroftenrelativetotheBWR/6design. Furthermore, theABWRalso improvedthe capability tomitigate severe accidents, eventhoughsucheventsareextremelyunlikely.Throughnitrogeninerting,containmentintegritythreatsfromhydrogengenerationwere eliminated.Sufficientspreadingareainthelowerdrywell,togetherwithadrywellflooding system,assurescoolabilityofpostulatedcoredebris.Manualconnectionsmakeitpossibletouseonsiteoroffsitewatersystemstomaintaincorecooling.Finally,toreducepotentialoffsiteconsequences,apassive,hard-pipedwetwellvent,controlledbyrupturedisks,isdesignedtopre-

ventcatastrophiccontainmentfailureandprovidemaximumfissionproduct scrubbing.The resultofthisdesigneffortisthatintheeventofasevereaccident,thewholebodydoseconsequenceatthecalculatedsiteboundaryislessthan25Rem.Theprobabilityofsuchanoccurrenceiscalculatedattheverylowlevelof10-9/year.MoreinformationonthissubjectcanbefoundinChapter10.

Improvements to Operation and Maintenance

Withthegoalofsimplifyingtheutilitysburdenofoperationandmaintenance (O&M) tasks, thedesignofeveryABWRelectricalandmechanicalsystem,aswellasthelayoutofequipmentintheplant,isfocusedonimprovedO&M.

Thereactorvesselismadeofforgedringsratherthanweldedplates.Thiseliminates30%oftheweldsfrom thecorebeltline region, forwhichperiodicin-serviceinspectionisrequired.SincetherearetenRIPsonfourpowerbuses,theABWRsrecirculationsystemisquiterobust.Pumpspeediscontrolledbysolid-stateadjustablespeeddrives,eliminatingtherequirementforflowcontrolvalvesandlow-speedmotor-generator sets.Thewetmotordesignalsoeliminatesrotatingseals.

TheFMCRDspermit a numberof simplifi-cations.First, scramdischargepipingand scramdischargevolumes(SDVs)wereeliminated,sincethehydraulic scramwater isdischarged into thereactorvessel.By supporting thedrivesdirectlyfrom thecoreplate, shootout steel locatedbelowthereactorvessel tomitigate therodejectionac-cidentwas eliminated.Thenumberofhydrauliccontrolunits (HCUs)was reducedbyconnectingtwodrivestoeachHCU.Thenumberofrodspergangwasincreasedupto26rods,greatlyimprov-ing reactor startup times.Finally, since therearenoorganicseals,onlytwoorthreedriveswillbeinspectedperoutage,ratherthanthe30specifiedinmostcurrentplants.

ItwaspossibletosignificantlydownsizeECCSequipmentasa resultof eliminating largevesselnozzlesbelowthetopofthecore.Capacityrequire-mentsaresizedbasedonoperatingrequirementstransient responseand shutdowncoolingratherthanontheneedforlargerefloodcapability.Inside

Figure 2-3. ABWR Reactor Pressure Vessel and Internals

Chapter 2 Plant Overview

2-6

Figure 2-4. ABWR Reactor Building and Containment

thereactorvessel,coresprayspargerswereelimi-nated,sincenopostulatedLOCAwouldleadtocoreuncovery.Fortransientresponse,theinitiationwaterlevelsforRCICandHPCFwereseparatedsothatthereisreduceddutyontheequipmentrelativetoearlierBWRs.Therearethreecompleteshutdowncoolingloops,includingdedicatedvesselnozzles.ComplexoperatingmodesoftheResidualHeatRe-moval(RHR)Systems,suchassteamcondensing,wereeliminated.Finally,heatremoval,inadditiontocoreinjection,wasautomatedsothattheoperatornolongerneedstochoosewhichmodetoperformduringtransientsandaccidents.

Lessons learned from operating experiencewereappliedtotheselectionofABWRmaterials.Stainlesssteelmaterialswhichqualifiedasresistanttointergranularstresscorrosioncracking(IGSCC)wereused.Inareasofhighneutronflux,materialswerealsospeciallyselectedforresistancetoirra-diation-assistedstresscorrosioncracking(IASCC).HydrogenWaterChemistry(HWC)isrecommendedfornormaloperationtofurthermitigateanypotentialforstresscorrosioncracking.

Theuseofmaterialproducingradioactivecobaltwasminimized.Thecondenserusestitaniumtubingatseawatersitesandstainlesssteeltubingforcool-ingtowersites.Theuseofstainlesssteelinapplica-tionsthatcurrentlyusecarbonsteelwasexpanded.DepletedZincOxide is recommended to furthercontrolradiationbuildup.Thesematerialschoicesreduceplant-wideradiationlevelsandradwasteandwillaccommodatemorestringentwaterchemistryrequirements.

Alsocontributingtogoodreactorwaterchem-istryistheincreaseoftheReactorWaterCleanupSystem(RWCU)capacitytotwopercent.AmorecompletesummaryofmaterialsandwaterchemistryconsiderationsisgiveninAppendixB.

TheOffgasSystemwassimplified,reflectinglessons learned from operating experience.Thecharcoalbedsaremaintainedatambienttemperaturerather than refrigerated.Thedesiccantdrierwaseliminated.

TheABWRReactorBuilding(includingcon-tainment)wasconfigured to simplifyand reducetheO&Mburden.Figure2-4illustratessomeofthe

keydesignfeaturesoftheABWRcontainment.Thecontainmentitselfisareinforcedconcretecontain-mentvessel(RCCV).

Within the containment itself, no equipment

SECONDARYCONTAINMENT

BOUNDARY

DRYWELLHEAD

DRYWELLCONNECTING

VENT

MSIV SRVPRIMARY

CONTAINMENTVESSEL

UPPERDRYWELL

DIAPHRAGMFLOOR

CLEANZONE

SUPPRESS.CHAMBERAIRSPACE

(WETWELL)

VACUUMBREAKER

LOWERDRYWELL

SUPPRESS.POOL

THERMALACTUATED VALVE BASEMATHORIZ. VENT

PRIMARYCONTAINMENT

BOUNDARY

SPILLOVERVENT

requires servicing during plant operation.ThecontainmentissignificantlysmallerthanthatoftheprecedingBWR/6.However,primarilyduetotheeliminationof the external recirculation system,thereisactuallymoreroomtoconductmaintenanceoperations.To simplifymaintenanceand surveil-lanceduring scheduledoutages,permanently in-stalledmonorailsandplatformspermit360access,andboththeupperandlowerdrywellshaveseparatepersonnelandequipmenthatches.TosimplifyRIPandFMCRDmaintenance, a rotatingplatform ispermanently installed in the lowerdrywell, andsemi-automatedequipmentwasspeciallydesignedtoremoveandinstallthatequipment.Thewetwellareaiscompactandisolatedfromtherestofcontain-ment,thusminimizingthechanceforsuppressionpoolcontaminationwithforeignmaterial.

AnewReactorBuildingdesignsurroundsthecontainmentand incorporates the same functionsastheBWR/6auxiliary,fuelanddiesel-generatorbuildings. Itsvolume (includingcontainment) isabout30%lessthanthatoftheBWR/6andrequiressubstantiallylowerconstructionquantities.Itslayout

Chapter 2 Plant Overview

2-7

isintegratedwiththecontainment,providing360accesswithservicingareaslocatedascloseaspracti-caltotheequipmentrequiringregularservice.Cleanandcontaminatedzonesarewelldefinedandkeptseparatebylimitedcontrolledaccess.Thefuelpoolissizedtostoreatleasttenyearsofspentfuelplusafullcore.Therefore,theBWR/6-typefueltransfersystemhasbeeneliminated.

Controls and instrumentationwereenhancedthroughincorporationofdigitaltechnologieswithautomated, self-diagnostic features.The use ofmultiplexingandfiberopticcablehaseliminated1.3millionfeetofcabling.Withinthesafetysystems,theadoptionofatwo-out-of-fourtriplogicandthefiberopticdatalinkshavesignificantlyreducedthenumberof requirednuclearboiler safety systemrelated transmitters. In addition, a three-channelcontrollerarchitecturewasadoptedfortheprimaryprocesscontrolsystemstoprovidesystemfailuretoleranceandon-linerepaircapability.

AnumberofimprovementsweremadetotheNeutronMonitoringSystem(NMS).Fixedwide-rangeneutrondetectorshave replaced retractablesourceand intermediate rangemonitors. Inaddi-tion,anautomatic,period-basedprotectionsystemreplaced themanual range switchesusedduringstartup.

Theman-machine interfacewas significantlyimproved and simplified for theABWR usingadvanced technologies such as large,flat-paneldisplays,touch-screenCRTsandfunction-orientedkeyboards.Thenumberofalarmtileswasreducedbyalmostafactoroften.Manyoperatingprocessesandproceduresareautomated,withthecontrolroomoperatorperformingaconfirmatoryfunction.Figure2-5illustratesthemaincontrolroom.

The plant features discussed above, whilesimplifying theoperatorsburden,haveanancil-

larybenefitof increased failure toleranceand/orreducederrorrates.StudiesshowthatlessthanoneunplannedscramperyearwillbeexperiencedwiththeABWR. Increased system redundancieswillalsopermiton-linemaintenance.Thus,bothforcedoutagesandplannedmaintenanceoutageswillbesignificantlyreduced.

Minimization of Radiation Exposure and RadwasteTheABWRcombinesadvancedfacilitydesign

featuresandadministrativeproceduresdesignedtokeeptheoccupationalradiationexposuretoperson-nelaslowasreasonablyachievable(ALARA).Dur-ingthedesignphase,layout,shielding,ventilationandmonitoringinstrumentdesignswereintegratedwithtraffic,securityandaccesscontrol.Operatingplant resultswerecontinuously integratedduringthedesignphase.Cleanandcontrolledaccessareasareseparated.

Reductionintheplantpersonnelradiationex-posurewasachievedby(1)minimizingthenecessityforandamountofpersonneltimespentinradiationareasand(2)minimizingradiationlevelsinroutinelyoccupiedplantareasinthevicinityofplantequip-mentexpectedtorequirepersonnelattention.

Changesinthematerialshaveasignificanteffectonthequantityofradwastegeneratedthroughradio-activecorrosionproducts.Inaddition,theconden-satetreatmentsystemwasimprovedtoincludebothpre-filtrationanddeepbeddemineralizerswithoutregenerationwhichreduceliquidandsolidradwasteinput.RadwastereductionintheABWRcanalsobefacilitatedthroughtheuseofadvancedincinerationandsuper-compactiontechnologies.

Reduced Capital CostDesignsimplificationsandquantitiesreductions

asdiscussedabove, togetherwith an increase inplantelectricaloutput,combinetomakeasignificantimprovementinplantcapitalcost.

Chapter 2 Plant Overview

2-8

Figure 2-5. ABWR (Lungmen) Main Control Room Panels

Chapter 3 Nuclear Boiler Systems

3-1

3ChapterNuclearBoilerSystemsOverview

TheNuclearBoilerSystems (NBS)producesteamfromthenuclearfissionprocess,anddirectthissteamtothemainturbine.TheNBSiscomprisedof the reactorvessel,which serves as ahousingforthenuclearfuelandassociatedcomponent,therecirculationsystem,thecontrolroddrivesystem,themain steam systemand the reactor buildingportionofthefeedwatersystem.Othersupportingsystems are described in Chapter 5,AuxiliarySystems.

Reactor Vessel and Internals

Thereactorvesselhousesthereactorcorethatistheheatsourceforsteamgeneration.Thevesselcontains thisheat, produces the steamwithin itsboundaries,andservesasoneofthefissionproductbarriers during normal operation.TheABWRreactorassemblyisshowninFigure3-1.Forthissizereactor,thediameteroftheABWRRPVisincreasedbut the height is decreased compared to earlierproductlines.Theincreaseddiameterhasresultedinincreasedwallthickness.TheRPVisapproximately21minheightand7.4mindiameter.

The most significant differences betweentheABWRRPVandearlierproduct linesareasfollows:

InwardvesselflangedesignSteamnozzlewithflowrestrictorDoublefeedwaternozzlethermalsleeve

ConicalvesselsupportskirtRelativelyflatbottomheadEliminationofnozzlesbelowthecoreReactorinternalpumppenetrationsUseof forged shell rings at andbelowcoreelevation.

The RPV design is based on proven BWRtechnology.A noteworthy feature is the lack ofanylargenozzlesbelowtheelevationofthetopofthecore.ThisRPVnozzleconfigurationprecludesanylargepiperupturesatorbelowtheelevationofthecore.ItisakeyfactorintheabilityofABWRsafety systems tokeep the core completely andcontinuouslyflooded for the entire spectrumofdesignbasisloss-of-coolantaccidents(LOCAs).

Thevesselcontainsthecoresupportstructurethatextendstothetopofthecore.Thepresenceofalargevolumeofsteamandwaterresultsintwovery important and beneficial characteristics. Itprovidesa largereserveofwaterabove thecore,whichtranslatesdirectlyintoamuchlongerperiodof timebeingavailablebeforecoreuncoverycanoccurasaresultoffeedflowinterruptionoraLOCA.Consequently,thisgivesanextendedperiodoftimeduringwhichautomaticsystemsorplantoperatorscan reestablish reactor inventory control usinganynormal,non-safety-relatedsystemcapableofinjectingwater into the reactor.Timely initiationofthesesystemsisdesignedtoprecludeinitiationof the emergency safety equipment.This easilycontrolledresponsetolossofnormalfeedwaterisasignificantoperationalbenefit.Inaddition,thelargerRPVvolume leads to a reduction in theABWRpressurization rate thatwouldoccuraftera rapidisolationofthereactorfromthenormalheatsink.

Chapter 3 Nuclear Boiler Systems

3-2

The following sections provide furtherdescriptionsof theuniquefeaturesof theABWRRPVandinternals.

Vessel Flange and Closure Head (1) 1To minimize the number of main closure

bolts, theABWRRPVhasan inside typeflange.This is different from the earlier product lines,whichhadoutsidetypevesselflanges.Theinsidetype vesselflange allows a hemisphericalmainclosure with radius less than the vessel radius.Also, this contributes tominimize theweightofthemainclosure.Thevesselclosuresealconsists

1.NumbersrefertoFigure3-1

oftwoconcentricO-ringswhichperformwithoutdetectable leakage at all operating conditions,includinghydrostatictesting.

Vent and Head Spray Assembly (2)The reactorwater cleanup returnflow to the

reactorvessel,viafeedwaterlines,canbedivertedpartly to a spray nozzle in the reactor head inpreparation for refueling cooldown.The spraymaintainssaturatedconditionsinthereactorvesselheadvolumebycondensingsteambeinggeneratedbythehotreactorvesselwallsandinternals.Theheadspraysubsystemisdesignedtorapidlycooldown the reactor vessel head flange region forrefuelingandtoallowinstallationofsteamlineplugs

Figure 3-1. ABWR Reactor Assembly

9

2

1

3

5

6

7

4

8

10

11

12

13

14

15

16

17

18

19

2021

2223

24

25

26

27

28

29

32

3130

1 - Vessel flange and closure head 2 - Vent and head spray assembly 3 - Steam outlet flow restrictor 4 - RPV stabilizer 5 - Feedwater nozzle 6 - Forged shell rings 7 - Vessel support skirt 8 - Vessel bottom head 9 - RIP penetrations10 - Thermal insulation11 - Core shroud12 - Core plate13 - Top guide14 - Fuel supports15 - Control rod drive housings16 - Control rod guide tubes17 - In-core housing18 - In-core guide tubes and stabilizers19 - Feedwater sparger20 - High pressure core flooder (HPCF) sparger21 - HPCF coupling22 - Low pressure flooder (LPFL)23 - Shutdown cooling outlet24 - Shroud head and steam separator assembly25 - Steam dryer assembly26 - Reactor internal pumps (RIP)27 - RIP motor casing28 - Core and RIP differential pressure line29 - Fine motion control rod drives30 - Fuel assemblies31 - Control rods32 - Local power range monitor

Chapter 3 Nuclear Boiler Systems

3-3

Figure 3-2. ABWR Reactor Pressure Vessel Feedwater Nozzle

beforevesselfloodupforrefueling.

The head vent side of the assembly passessteamandnoncondensablegasesfromthereactorheadtothesteamlinesduringstartupandoperation.Duringshutdownandfillingforhydrotesting,steamandnoncondensablegasesmaybevented to thedrywellequipmentsumpwhile theconnection tothesteamlineisblocked.Whendrainingthevesselduringshutdown,airentersthevesselthroughthevent.

Steam Nozzle with Flow Restrictor (3)TheABWRRPVhasflowrestrictingventuri

located in the steam outlet nozzles. Besidesproviding an outlet for steam from the reactorpressurevessel,thesteamoutletnozzleswillprovidefor(1)steamlinebreakdetectionbymeasuringsteamflow to signala trip for themainsteam isolationvalves,(2)steamflowmeasurementforinputtothefeedwatercontrolsystem,and(3)aflow-chokingdevicetolimitblowdownandassociatedloadsontheRPVandinternalsintheeventofapostulatedmainsteamlinebreak.Calculationsshowthatthepressuredropinthenozzleiswithintherequirementsofthesteady-stateperformancespecification.

RPV Stabilizer (4)Stabilizers are located around theperiphery

of theRPV toward itsupperend.TheseprovidereactionpointstoresisthorizontalloadsandsuppressRPVmotionduetoearthquakesandpostulatedpiperuptureevents.

Feedwater Nozzle Thermal Sleeve (5)Thefeedwaternozzlesutilizedouble thermal

sleevesweldedtothenozzles.Thedoublethermalsleeveprotectsthevesselnozzleinnerblendradiusfromtheeffectsofhighfrequencythermalcycling.Aschematicof the feedwaternozzle is shown inFigure3-2.

Use of Forged Shell Rings (6)TheABWRRPVutilizeslowalloyforgedshell

rings,perASMESA-508,Class3,adjacenttoandbelow thecorebelt line region.TheflangesandlargenozzlesarealsoperASMESA-508,Class3.TheshellringsabovethecorebeltlineregionandthemainclosurearemadefromlowalloysteelplateperASMEA-533,TypeB,Class1.Therequired

Reference Nil Ductility, RTNDT, of the vesselmaterialis-20C.Figure3-3showsoneoftheRPVforgedshellringsduringfabrication.

Vessel Support Skirt (7)Thevesselsupportskirthasaconicalgeometry

andisattachedtothelowervesselcylindricalshellcourse.Thesupport skirt attachment (knuckle) isan integralpartof thevessel shell ring.Locatingthe conical support skirt on the lower shell ringprovides:

Needed space for the reactor internal pump(RIP)heatexchangers.

Figure 3-3. RPV Forged Steel Ring

Chapter 3 Nuclear Boiler Systems

3-4

AccessforISIofthebottomheadweld.

Reactor Vessel Bottom Head (8) Thebottomheadconsistsofasphericalbottom

cap, made from a single forging, extending toencompass the CRD penetrations and a conicaltransitionsectiontothetoroidalknucklebetweenthehead andvessel cylinder.Withbottomheadthicknessof approximately250mm, thebottomheadmeetstheASMEallowablesforthespecifieddesignloads.ThemainadvantageofusingasingleforgingforthebottomheadisthatiteliminatesallRPVweldswithintheCRDpattern,thusreducingfuturein-serviceinspection(ISI)requirements.

Reactor Internal Pump Penetrations and Weld (9)Themost significant differencebetween the

ABWRandearlierBWRproductlinesiseliminationof allmajorpipeconnectionsbelow thecorebyincorporating internal recirculationpumps in thereactor.TheRIPmotorcasingsareweldedtothevesselbottomheadbyadesignasshowninFigure3-4.

VerticalrestraintsareprovidedtopreventthemotorcasingorthemotorcoverfromblowingoutintheunlikelyeventofafailureoftheweldbetweentheRPVandthemotorcasingorafailureof the

motorcoverbolts.Inaddition,iftherestraintsshouldfail,thepumpimpellerisdesignedtobackseatonthestretchtubethatkeepsthepumpdiffuserinplaceandpreventsignificantleakagethroughthefailedpart.MoreinformationabouttheRIPcanbefoundunderRecirculationSystemlaterinthischapter.

Thermal Insulation (10)TheRPV insulation is reflectivemetal type,

constructedentirelyof series300 stainless steel.Theinsulationismadeupofacombinationoftwobasicshapes:flatpanelsandcylindricalpanels.Theinsulationforthebottomheadandlowershellcourseinside thevessel support is avertical cylindricalpanelapproximately75to100mmthick.Thereisalsoahorizontalpanelofthesamethicknesswhichconnectsacrossthebottomoftheverticalpanels.TheCRDhousings,in-corehousingsanddrainnozzlespenetratethispanel.

TheinsulationfortheRPVissupportedfromthereactorshieldwallsurroundingthevessel,andnotfromthevesselshell.Insulationfortheupperheadandflangeissupportedbyasteelframeindependentofthevessel.

Atoperatingconditions, the approximate airtemperaturesoutsidethevesselandinsulationare57Cabove the tophead and38Ceverywhereelse.

Core Shroud (11)The shroud is a stainless steel cylindrical

assembly thatprovidesapartition toseparate theupwardflowofcoolantthroughthecorefromthedownwardrecirculationflow.Thevolumeenclosedbytheshroudischaracterizedbyupperandlowerregions.Theupperportionoftheshroudsurroundstheactivefuelandformsthelongestsectionoftheshroud.Thissectionisboundedatthebottombythecoreplate.Thelowershroud,surroundingpartofthelowerplenum,isweldedtotheRPVshroudsupport.Theshroudprovideslateralsupportforthecorebysupportingthecoreplateandtopguide.

Core Plate (12)Thecoreplateconsistsofacircularplatewith

round openings.The core plate provides lateralsupportandguidanceforthecontrolrodguidetubes,in-corefluxmonitorguide tubes,peripheral fuel

Figure 3-4. Reactor Internal Pump Motor Casing Including Weld to RPV

Chapter 3 Nuclear Boiler Systems

3-5

supports,andstartupneutronsources.Thelasttwoitemsarealsosupportedverticallybythecoreplate.Theentireassemblyisboltedtoasupportledgeinthe shroud.Thecoreplatealso formsapartitionwithin theshroud,whichcauses the recirculationflowtopassintotheorificedfuelsupportandthroughthefuelassemblies.

Top Guide (13)Thetopguideconsistsofagridthatgiveslateral

supportofthetopofthefuelassemblies,acylindersupportingcoreflooderspargers,andatopflangeforattachingtheshroudhead.Eachopeningprovideslateralsupportandguidanceforfourfuelassembliesor,inthecaseofperipheralfuel,one,twoorthreefuelassemblies.Holesareprovidedinthebottomofthesupportintersectionstoanchorthein-corefluxmonitorsandstartupneutronsources.Thetopguideisboltedtothetopoftheshroud.

Fuel Supports (14)Thefuelsupportsareoftwobasictypes;namely,

peripheralfuelsupportsandorificedfuelsupports.The peripheral fuel supports are located at theouteredgeoftheactivecoreandarenotadjacenttocontrolrods.Eachperipheralfuelsupportsustainsonefuelassemblyandcontainsanorificedesignedtoassurepropercoolantflowtotheperipheralfuelassembly.Eachorificedfuelsupportsustainsfourfuelassembliesverticallyupwardandhorizontallyandisprovidedwithorificestoassurepropercoolantflowdistributiontoeachfuelbundle.Theorificedfuelsupportsitsonthetopofthecontrolrodguidetube,whichcarriestheweightofthefuelrodsdownto thebottomof theRPV.Thecontrol rodspassthroughcruciformopenings in the center of theorificedfuelsupport.

Control Rod Drive Housing (15)Thecontrolroddrivehousingprovidesextension

oftheRPVforinstallationofthecontrolroddrive,andtheattachmentoftheCRDline.Italsosupportstheweightofacontrolrod,controlroddrive,controlrodguidetube,orificedfuelsupportandfourfuelassemblies.

Control Rod Guide Tubes (16)Thecontrolrodguidetubesextendfromthetop

ofthecontrolroddrivehousingsupthroughholesinthecoreplate.Eachguidetubeisdesignedasthe

guideforthelowerendofacontrolrodandasthesupportforanorificedfuelsupport.Thislocatesthefour fuel assemblies surrounding thecontrol roddrivehousing,which,inturn,transmitstheweightoftheguidetube,fuelsupport,andfuelassembliestothereactorvesselbottomhead.Thecontrolrodguidetubealsocontainsholes,nearthetopofthecontrolrodguidetubeandbelowthecoreplate,forcoolantflowtotheorificedfuelsupports.Inaddition,theguidetubeprovidesaconnectiontotheFMCRDtorestrainahypotheticalejectionoftheFMCRD.

In-core Housing (17)The in-core housings provide extensions of

theRPVatthebottomheadfortheinstallationofvariousin-corefluxmonitoringsensorassemblieswhicharecomponentsoftheNeutronMonitoringSystem.Italsosupportstheweightofanin-corefluxmonitoringsensorassembly,in-coreguidetubeandpartofthein-coreguidetubestabilizerassembly.

In-Core Guide Tubes and Stabilizers (18)Thein-coreguidetubesextendfromthetopof

thein-corehousingtothetopofthecoreplate.Theyprovidethein-coreinstrumentationwithprotectionfromflowofwaterinthebottomheadplenum,andguidanceforinsertionandwithdrawalfromthecore.The in-coreguide tube stabilizersprovide lateralsupportandrigiditytothein-coreguidetubes.

Feedwater Spargers (19)Thefeedwaterspargersareattachedtobrackets

onthevesselwallanddelivermakeupwatertothereactorduringplantstartup,powergenerationandplantshutdownmodesofoperation.Nozzlesinthespargersprovideuniformdistributionoffeedwaterflowwithinthedowncomerflowpassage.

High Pressure Core Flooder Sparger Assembly (20)Thehighpressurecoreflooder(HPCF)spargers

insidethecylinderofatopguidearearrangedtoprovideemergencycoolantinjectionovertheupperendofthecore.Thespargershavethefunctionofastandbyliquidcontrolsolutioninjection.TheHPCFspargers are connected to theHPCFnozzlesbymeansofanHPCFcoupling(21).

Low Pressure Flooder Spargers (22)Thetwofloodingspargersthatareattachedto

thevesselwalldeliverflowat lowpressurefrom

Chapter 3 Nuclear Boiler Systems

3-6

the RHR System and distribute it in the upperplenumabovetheshroudheadofthereactor.Flowisdeliveredineitheroftwomodes:(1)forthefloodingofthereactorintheeventofanabnormaldropinwaterlevel,or(2)inthecirculationofcoolingwaterto remove residualandcoredecayheat from thereactorduringshutdown.

Shutdown Cooling Nozzles (23)Suction for the RHR System in shutdown

coolingmodeisprovidedbythreeshutdowncoolingnozzles.

Shroud Head and Steam Separator Assembly (24)The steam separator assembly consists of a

slightly domed base on top of which is weldedan array of standpipes with a three-stage steamseparatorlocatedatthetopofeachstandpipe.Thesteamseparator assembly restson the topflangeofthecoreshroudandformsthecoverofthecoredischarge plenum region.The seal between theseparatorassemblyandcoreshroudflangesismetal-to-metalcontactanddoesnotrequireagasketorotherreplacementsealingdevices.Theseparatorassemblyisboltedtothecoreshroudflange,bylongholddownboltswhich,foreaseofremoval,extendabovetheseparators.During installation, theseparatorbaseis alignedon the core shroudflangewithguiderodsandfinallypositionedwithlocatingpins.Theobjectiveofthelong-boltdesignistoprovidedirectaccesstotheboltsduringreactorrefuelingoperationswithminimum-depthunderwatertoolmanipulationduringtheremovalandinstallationoftheassemblies.Itisnotnecessarytoengagethreadsinmatinguptheshroudhead.Atee-boltengagesinthetopguideanditsnutistightenedtoonlynominaltorque.Finalloadingisestablishedthroughdifferentialexpansionoftheboltandcompressionsleeve.Thefixedaxialflowtypesteamseparatorshavenomovingpartsandaremadeofstainlesssteel.Ineachseparator,thesteam-watermixturerisingthroughthestandpipeimpingesonvaneswhichgivethemixtureaspintoestablishavortexwhereinthecentrifugalforcesseparatethewaterfromthesteamineachofthreestages.Steam leaves the separatorat the topandpassesintothewetsteamplenumbelowthedryer(Figure3-5).The separatedwater exits from thelowerendofeachstageoftheseparatorandentersthepool thatsurrounds thestandpipes to join thedowncomerannulusflow.

Steam Dryer Assembly (25)

ThesteamdryerassemblyconsistsofmultiplebanksofdryerunitsmountedonacommonstructurewhichisremovablefromtheRPVasanintegralunit.Theassemblyincludesthedryerbanks,dryersupplyanddischargeducting,draincollectingtrough,drainducts,andaskirtwhichformsawatersealextendingbelowtheseparatorreferencezeroelevation.Steamfrom the separators flows upward and outwardthroughthedryingvanes(Figure3-6).Thesevanes

Figure 3-5. Schematic of Steam Flow through Separator

DRYER STEAM

RETURNINGWATER

WATER LEVELOPERATING RANGE

SKIRT

WETSTEAM STANDPIPECORE

DISCHARGEPLENUM

TURNINGVANES

Chapter 3 Nuclear Boiler Systems

3-7

areattachedtoatopandbottomsupportingmemberformingarigid,integralunit.Moistureisremovedandcarriedbyasystemoftroughsanddrainstothepoolsurroundingtheseparatorsandthenintotherecirculationdowncomerannulusbetweenthecoreshroudandreactorvesselwall.Upwardandradialmovementofthedryerassemblyundertheactionofblowdownandseismicloadsislimitedbysupportbracketsonthevesselshellandholddownbracketsinsidethemainclosure.Theassemblyisarrangedforremovalfromthevesselasanintegralunitonaroutinebasis.

Reactor Internal Pumps (RIP) (26,27)Refer to thenext section for informationon

ReactorInternalPumps(RIP).

Core and RIP Differential Pressure Lines (28)Theselinescomprisethecoreflowmeasurement

subsystemoftheRecirculationFlowControlSystem(RFCS) andprovide twomethodsofmeasuringtheABWRcoreflowrates.ThecoreDPlinesandinternal pumpDP lines enter the reactor vessel

separatelythroughreactorbottomheadpenetrations.FourpairsofthecoreDPlinesentertheheadinfourquadrantsthroughfourpenetrationsandterminateimmediatelyaboveandbelowthecoreplatetosensethepressureintheregionoutsidethebottomofthefuel assemblies andbelow the coreplateduringnormaloperation.Similarly,fourpairsoftheinternalpumpDPlinesterminateaboveandbelowthepumpdeckandareusedtosensethepressureacrossthepumpduringnormalpumpoperation.

Fine Motion Control Rod Drives (29)RefertothediscussionontheControlRodDrive

Systemlaterinthischapter.

Fuel Assemblies, Control Rods and Local Power Range Monitors (30-32)

RefertotheChapter6discussionforfuelandrelatedhardware.

Recirculation SystemThe function of the Reactor Recirculation

System(RCIR)isto:Provide forcedcirculationof reactor coolantfor energy transfer from fuel to the coolingfluidand,asaresult,generatealargeramountofsteam.Control the reactor power by changing therecirculationflow;theflowiscontrolledbytheuseofadjustablespeedpumps.

TheRCIRSystemprovidesforcedcirculationofreactorwaterthroughthecore,removingtheheatproducedby the fuel.The reactorwater ismadeupofwater removed from the two-phase reactorcoolant(coreflow)inthemoistureseparatorsandsteamdryersandtheincomingfeedwaterflow.TheRCIRSystemusesanarrangementoftenpumpstoprovidethemotiveforceforcoreflow.Thepumpsaremounted internally in the reactorvessel andarecalledreactorinternalpumps(RIPs).TheRIPsfunctioncollectively to force the reactor coolantthroughthelowerplenumofthereactorandupwardthrough openings in the fuel support castings,through the fuel bundles, steam separators, anddowntheannulustobemixedwithfeedwaterand

Figure 3-6. Schematic of Steam Flow Through Dryer

A A

DRYERLIFTINGBAR (4)

DRYERSKIRT

STEAMWATER

FLOW

DRAIN CHANNELS(AT EACH DRYER

SECTION)

STEAM WATER FLOWFROM SEPARATORS

VERTICAL GUTTER STRIPOR "MOISTURE HOOK"

STEAM DRYER ASSEMBLY

CROSS SECTION A-A

Chapter 3 Nuclear Boiler Systems

3-8

recirculatedthroughthecore.Figure3-7showstheRIPsandthepumpedflowpath.

Recirculationflowrateisvariableoverarangefromnaturalcirculationflowof20%toabovetheratedflowrequiredtoachieveratedcorepower.Infact,theRCIRdesigncanproduceratedcoreflowrateat100%reactorpowerwithnineofitstenpumpsoperating.Theflowcontrolrangeallowsautomaticregulationof reactorpoweroutputbetween~70to 100% without control rod movement. Coreflow(RCIRpumpingcapacity)isregulatedbytheRecirculationFlowControlSystem (RFC).TheRFC System provides conditioned control andlogicsignals,whichregulatetheRIPspeed,which,

inturn,regulatesthepumpflow.Becausethecoreflowaffectsreactorpowerandfuelthermalmargins,theRCIRSystemisalsousedtomitigatetheeffectsoftransient,upsetandemergencymodesofreactoroperation.

TherearethreeRCIRsubsystemswhichareusedinconjunctionwiththereactorinternalpump:

RecirculationMotorCoolingSubsystem(RMCSubsystem)RecirculationMotorPurgeSubsystem (RMPSubsystem)Recirculation Motor Inflatable Shaft SealSubsystem(RMISS)

Recirculation Motor Cooling Subsystem (RMC)EachRIPhasitsownexternalheatexchanger

(Figure3-8).EachRIPmotorcasingandtheRIPheatexchanger isconnectedvia stainless steelpiping.Theheatexchangerisatypicalshell-tubetypewithU-tubessupportedbybaffles.Thehotwatercomingfromthemotorentersfromtheupperendoftheheatexchangershellsideandleavesfromthelowerendoftheshellsideandreturnsbacktothemotor.TheconnectingpipingisweldedtotheRIPmotorcasingandalsototheheatexchangershelltopreventanyleakageduringtheplantoperation.

Recirculation Motor Purge Subsystem (RMP)Thepurgesystemprovidesasourceofclean

control roddrive (CRD)water thatflowsup theannulus between the stretch tube and the shaftandpreventstheintrusionofreactorwaterwithitsassociatedcontaminationintothemotor.ThepurgesystemnormallyoperatescontinuouslyevenwhentheRIPsaretrippedorthereactorisshutdownfortherefuelingand/ormaintenance.

Recirculation Motor Inflatable Shaft Seal Subsystem (RMISS)

Duringnormaloperation, thepurposeof thissystemistopreventanyleakageofreactorwater(escaping from the primary seal) during plantoutagesandtoassistinmaintenanceorinspectionofmotors.DuringthemaintenanceoftheRIPs,aportablepumpisusedtopressurizethesealusingwaterfromtheMakeupWaterSystem(MUW).Thesealismadeofelastomericmaterialandsealstightlybetweenthepumpshaftandthepumpmotorcasing.

Figure 3-7 Recirculation Flow Schematic

Chapter 3 Nuclear Boiler Systems

3-9

Thepumppressurizesthesealandmaintainsitatshutoffheadconditions.

Reactor Internal PumpsThevessel-mountedRIPs simplify theRPV

by eliminating all largenozzlesbelow the core,significantlyreducingpipingin-serviceinspection(ISI)andpersonnelexposure, andallowing foracompact containment design due to eliminationof the external recirculation systempiping.Useof theRIP featureallows for theoptimizationofthe Emergency Core Cooling System (ECCS)andassuresnocoreuncoveryforpostulatedpipebreaks.OneofthegoalsfortheABWRistoreducecalculatedcoredamagefrequencybyanorderofmagnitude relative toGEspreviouslydesignedBWRoperatingplants.Oneofthemostimportantdesign features contributing to thisgoalwas theadoptionofRIPs inplaceof externallypumpedreactorrecirculationsystem/pumps.

Theinternalpumpsareanimprovedversionofa

EuropeandesignedRIPthatisinoperationinmanyEuropeannuclearpowerplants.About9millionpumphoursofsuccessfuloperatingexperiencehasbeenaccumulated,withsomepumpshavingbeeninservicesincethemid-1970s.

Thegeneraldesigndetailsof theRIP,motor,andheatexchangerareasfollows:

Number of Pumps: 10Type of Pump: Vertcal shaft, sngle stage, mixed flowRated Flow: 7700 m3/hr/pumpRated Head: 40 mRated Pump Speed: 1500 rpmOverall Height (Impeller & Motor): 3 mOverall Weight: 5000 kgMotor Type: 3-Phase, Wet Inducton Motor

Figure 3-8. RCIR Subsystems

PUMPMOTORCASING

RMCSUBSYSTEM

RBCW

REACTOR VESSEL

RMP SUBSYSTEM

RMISS SUBSYSTEM

ARD

PUMPDECK

SHROUD

HEATEXCHANGER

RIP MOTORCOOLING PIPING

RPV

RIPs

A

B

C

D

EF

G

H

J

K

MOTORCOVER

Chapter 3 Nuclear Boiler Systems

3-10

Rated Output Power: 830 kWRated Voltage: ~ 3300 VHeat Exchanger Type: Shell and TubesHx Cooling Capacity: 1.15 kcal/hr

Reactor Internal Pump Component DescriptionThereare10RIPsarrangedcircumferentially

between the shroud and theRPVnear theRPVbottomhead.Figure3-9showsacrosssectionoftheRIPusedintheABWRandkeycomponentsaredescribedbelow.

Diffuser:TheRIPhastheimpelleranddiffuserinsidetheRPV.Thediffuserisinstalledinthepumpdeckandsealedbyapistonringarrangement.TheRIPdiffuserisremovable.ThediffuserisretainedontheRPVnozzlebythestretchtube.

StretchTube:Thestretch tube isessentiallyalonghollowboltwhichpassesthroughtheRPVnozzlepenetrationfromthediffuser to thetopoftheRIPmotorcasingwhereitisheldintensionbyalargenut.Thestretchtubeispreloadedbyuseofastud tensionersimilar to thatusedfor themainclosurestudsof theRPV.Thepumpshaftpassesthrough thecenterof the stretch tubeandmotorrotor.Thepumpshaftkeyfitsinaslotinthemotorrotortube.

ImpellerandPumpShaft:Theimpellerandthepumpshaftareconnectedbytheimpellerbolt.Thepumpshaftpassesthroughthestretchtube,rotorandisconnectedtothethrustbearingdiskatitslowerend.Themotorrotorkeywayslotfitswiththekeyonthepumpshaftandtransmitsmotortorque.

Radial Bearings: The motor upper radialbearingisbelowthesecondaryseal.Thisbearingdesign has been tested and proven to eliminatebearinginstabilityduetohalfspeedrotation.

Thelowerradialbearingislocatedbelowthemotorrotorandthestator.Thelowerradialbearingissimilartotheupperradialbearing.

ThrustBearing:The thrustbearing isofanoffsettiltingpadconfiguration.Therotatingportionof the thrustbearing is integralwith thecoolingwater auxiliary impeller,which circulateswaterthroughthemotorandbearingtoprovidecoolingandcleaningviathepurgesystem.

Anti-Reverse Rotation Device: Below thecoolingwaterauxiliaryimpelleristheAnti-ReverseRotation Device (ARD).This is a cam clutcharrangement thatprevents theRIP from rotatingin the reversedirectionwhenoneRIP is trippedwhile theothersare running(whichcanresult inbackflow through the trippedRIP).Thepurposeof theARD is toprevent reverse rotationof thepumpshaftandminimizethebackflowthroughanidle/trippedRIP.

Motor(StatorandRotor):TheRIPmotorisa3-phase,4-polewetinductionmotor.Thecoolingwaterflowsupward through thewindingsof thestatorandtherotor.Themotorstatorisattachedtothemotorcover.

Figure 3-9 Cross-Section of RIP

RPV

IMPELLER

PURGE WATERINLET

SECONDARY SEALPRESSURIZATIONWATER INLET

COOLING WATEROUTLET

MOTOR CASING

PUMP SHAFT

MOTOR ROTOR

COUPLING STUD

THRUST DISKAUXILIARY IMPELLLER

TERMINALBOX

ARD

CABLECONNECTOR

MOTOR COVER

AUXILIARY COVER SPEED SENSOR

COOLINGWATERINLET

THRUSTBEARINGPADS

LOWER JOURNALBEARING

STATOR

UPPER JOURNALBEARING

SECONDARYSEAL

STRETCH TUBENUT

STRETCH TUBE

DIFFUSERWEAR RING

PISTON RINGDIFFUSER

Chapter 3 Nuclear Boiler Systems

3-11

TerminalBox:Theelectricalterminalboxisboltedtothemotorcover.Themotorwindingcablepenetrationspassthroughthemotorcovercoolantpressureboundaryandareconnectedtothepowersupplyleadsatthislocation.EachmotorisdrivenbyitsownvariablefrequencypowersupplyknownastheAdjustableSpeedDrive(ASD).

SpeedandVibrationSensors:Thereare2-pumpspeedsensorsand2-motorcasingvibrationsensors on each RIP motor casing.There is anadditionalsensoroneachRIPtodetectrubbingofinternalpartsofthepump.

RIP OperationWhenever the RIP motor is started, it is

controlledtoreachitsminimumspeed.Similarly,onebyone,theother9RIPsarestartedandbroughttotheminimumspeedlevel.Fromthiscondition,thespeedofall10RIPscanbeincreasedindividuallywhenintheindividualspeedcontrolmode,orasagroupwhen in the automaticgangedmodeofoperation,withthegangedmodebeingthenormallypreferredmodeafterall10RIPshavebeenstarted.TheRFCSystemcontrols the speedof theRIPsasdescribed earlier in this section.Achange inRIPspeedconditionswillvarythecoreflowinthereactor,which,inturn,willchangethereactorpowerduringnormalpowerrangeoperation.

TheRFCoperationalmodesalso include thecoreflowcontrolmodeand the automatic load-followingmode.ThecoreflowmodecontrolsthespeedoftheRIPsselectedforgangspeedoperationtomaintainthesteady-statecoreflowequaltothecoreflowdemandsignal.Fortheautomaticload-followingmode,theRFCSystemcontrolsthespeedofthoseRIPsselectedforgangspeedoperationtoreducetheloaddemanderrorsignal(fromtheturbinecontrolsystem)tozero.

IndividualRIPspeedcontroloperationmodeand thegangedspeedmodeofoperationprovidesignificantflexibilityduringnormalplantoperation.If,foranyreason,oneRIPdevelopsaproblem,theneitherspeedcanbereducedtoeliminatetheproblemor thatRIPcanbe tripped, ifnecessary,withoutaffectingthecontinuedoperationofotherRIPs.

Duringnormalplantratedpoweroperation(in

eitherthecoreflowcontrolmodeortheautomaticload-followingmode),ifoneRIPisloweredinspeedortripped,thenthespeedoftheremaining9RIPsisincreasedbythecontrolsystemtomaintainthedemandedcoreflow;thus,steady-stateplantoutputpowerremainsunaffected.

RIP Power SupplyThe RIP motor is driven by a solid-state

variable-frequency power supply known as theAdjustable Speed Drive (ASD).TheASD is aprovenproductwithwideindustrialapplicationsaswellasexperienceintheEuropeannuclearplants.TheABWRapplicationuses~3000Vfortheoutputvoltage rating.TheASDpower supplyprovidesextremely lowmaintenance,high reliability, andprovidesexcellentRIPspeedmaneuverability.

EachRIPisdrivenbyitsdedicatedASD.SixRIPASDsreceivepowerfromconstantspeedMotor-Generator (M-G) sets and theother fourdirectlyfrom medium voltage buses.A representativesimplifiedpowerdistributionone-linediagram isshowninFigure3-10.EachM-Gsetprovidespowerto3associatedRIPASDs.TheotherfourRIPASDsaredividedintotwosetsreceivingpowerdirectlyfromtwoseparatemainbuses.

The assignment of the power distribution

MEDIUMVOLTAGEBUS NO. 1

MEDIUMVOLTAGEBUS NO. 4

MEDIUMVOLTAGEBUS NO. 3

MEDIUMVOLTAGEBUS NO.2

DIS-CONNECT

SWITCH

RECTIFIER

DC LINKGTO

INVERTEROUTPUTTRANS-

FORMERRIPB

RIPE

RIPH

RIPA

RIPF

RIPD

RIPJ

RIPC

RIPG

RIPK

M

G

M

G

Figure 3-10. RIP Power Supply Diagram

Chapter 3 Nuclear Boiler Systems

3-12

to individualRIPASDs is chosen tobalance theazimuthaldistributionwithinthevessel(e.g.,whenaM-Gsettripsoronemediumvoltagebusislost).TheM-Gsetshave inertialflywheels toprovidecontinuedoperationoftheassociatedRIPsduringeitherthemomentaryorcompletelossofincomingpower.Aftercompletelossofthemainbuspower,continuedoperationof theseRIPs for at least 3secondsisprovidedviatheM-Gsets.

Control Rod Drive SystemTheControlRodDrive(CRD)Systemcontrols

changesincorereactivityduringpoweroperationbymovementandpositioningoftheneutronabsorbingcontrolrodswithinthecoreinfineincrementsinresponse tocontrolsignalsfromtheRodControlandInformationSystem(RCIS).TheCRDSystemprovides rapid control rod insertion in responsetomanualorautomatic signals from theReactor

ProtectionSystem (RPS).Figure3-11 shows thebasicsystemconfigurationandscope.

WhenscramisinitiatedbytheRPS,theCRDSysteminsertsthenegativereactivitynecessarytoshutdownthereactor.Eachcontrolrodisnormallycontrolledbyanelectricmotorunit.Whenascramsignal is received, high-pressurewater stored innitrogenchargedaccumulators forces thecontrolrodsintothecore.Thus,thehydraulicscramactionisbackedupbyanelectricallyenergizedinsertionofthecontrolrods.

The CRD System consists of three majorelements:

Electro-hydraulicfinemotioncontrolroddrive(FMCRD)mechanismsHydrauliccontrolunit(HCU)assembliesControl Rod Drive Hydraulic System(CRDHS)

TheFMCRDsprovide electric-motor-drivenpositioningfornormalinsertionandwithdrawalof

Figure 3-11. CRD System Schematic

SEPARA-TIONSIGNAL

POSITIONSIGNAL

CRD

CRD MOTOR

SCRAM ANDPURGE

WATER TOREACTOR

POSITIONSIGNAL

CRD

CRD MOTOR

SCRAM AND

WATER TOREACTOR

SCRAMVALVE

TESTCONNECTION

TO OTHERHCUS

TO OTHERHCUS

SCRAMACCUMULATOR

SCRAMVALVE

SCRAMPILOTVALVE

EXHAUST

TO ALLOTHERSCRAMPILOT

VALVES

INSTRUMENTAIR SUPPLY

EXHAUST EXHAUSTAIR HEADER DUMP

VALVES

CSPCONDENSATE/FEEDWATER

CHARGING LINE

PURGEWATERLINE

HYDRAULIC CONTROL UNIT (HCU)(TYPICAL)

EXHAUST

PURGE

N2SUPPLY

PURGEFLOW

CONTROLVALVES

DRIVEWATERFILTERS

FILTER

FILTER

SUCTIONFILTERS

FILTER

FILTER

CRDPUMPS

Chapter 3 Nuclear Boiler Systems

3-13

thecontrolrodsandhydraulic-poweredrapidcontrolrod insertion for abnormaloperating conditions.Simultaneouswithscram,theFMCRDsalsoprovideelectric-motordrivenrun-inofcontrolrodsasapathtorodinsertionthatisdiversefromthehydraulic-powered.Thehydraulicpowerrequiredforscramisprovidedbyhighpressurewater stored in theindividualHCUs.AnHCUcanscramtwoFMCRDs.It alsoprovides theflowpath forpurgewater totheassociateddrivesduringnormaloperation.TheCRDHSsuppliespressurizedwater for chargingtheHCUscramaccumulatorsandpurging to theFMCRDs.

Fine Motion Control Rod DrivesTheABWRFMCRDsaredistinguishedfrom

thelockingpistonCRDs,whichareinoperationinallcurrentGEplants,inthatthecontrolbladesaremovedelectricallyduringnormaloperation.Thisfeature permits small power changes, improvedstartup time, and improvedpowermaneuvering.TheFMCRD,aswithcurrentdrives,isinsertedintothecorehydraulicallyduringemergencyshutdown.BecausetheFMCRDhastheadditionalelectricalmotor,itdrivesthecontrolbladeintothecoreeveniftheprimaryhydraulicsystemfailstodoso,thusprovidinganadditionallevelofprotectionagainstATWSevents.TheFMCRDdesignisanimprovedversionofsimilardrivesthathavebeeninoperationinEuropeanBWRssince1972.

Figure 3-12 shows a cross-section of theFMCRD as used in theABWR.The FMCRDconsistsoffourmajorsubassemblies:thedrive,thespoolpiece,thebrakeandthemotor/synchros.Thespoolpiece andmotormaybe removedwithoutdisturbing thedriveand this allowsmaintenancewithlowpersonnelexposure.

Thedrive consistsof theouter tube,hollowpiston,guidetube,buffer,labyrinthseal,ballcheckvalve,spindleadaptorandsplinedspindleadaptorbackseat.

Thecouplingisabayonetconfigurationwhich,when coupled with the mating coupling on thecontrolrodblade,precludesseparationofthebladeandthehollowpiston.

Thehollowpistonisalonghollowtubewitha

pistonheadatthelowerend.Thehollowpistonisdrivenintothereactorduringscrambythepressuredifferential that is produced by the high scramflow.Thelabyrinthseal,whichiscontainedinsidethebuffer,atthetopendoftheoutertuberestrictsthe flow from the drive to the reactor, therebymaximizing the pressure drop which enhancesscramperformance.Additionally,itallowsthepurgeflowduringnormaloperationtoprecludeentranceofreactorwaterandassociatedcrudintothedrive.The piston head contains latches that latch intonotches in thedriveguide tubeafter scram.ThescrambufferingactionisprovidedbyanassemblyofBellevillewashersinthebufferandissupplementedbyhydraulicdampingasthebufferassemblypartscometogether.

SPOOL PIECE

POSITION INDICATORPROBE (PIP)

FULL-IN MECHANISM

BAYONET COUPLING TYPEINTERNAL CRD BLOWOUTSUPPORT (TO CONTROL RODGUIDE TUBE BASE COUPLING)

BUFFER

SCRAM POSITIONSENSING MAGNET

BACK SEAT

SEAL HOUSING

MOTOR

SYNCHRO SIGNALGENERATOR

SEPARATIONPROBE

SEPARATIONSENSING SPRING

SEPARATIONSENSING MAGNET

BALL CHECKVALVE

SCRAM LINEINLET

GUIDE TUBE

BALL SCREW

LABYRINTH SEAL

OUTER TUBE

BAYONET COUPLING TYPECRD SPUD(TO CONTROL RODSOCKET COUPLING)

BALL NUT

MIDDLE FLANGE

HOLLOW PISTON

FMCRD HOUSING

LEAK-OFF PIPING

BRAKE

Figure 3-12. Fine Motion Control Rod Drive Cross-Section

Chapter 3 Nuclear Boiler Systems

3-14

The outer tube performs several functions,one of whic