Regulatory Guide 1.183

Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors

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Issue date:	07/31/2000
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RegulatoryguidesareissuedtodescribeandmakeavailabletothepublicsuchinformationasmethodsacceptabletotheNRCstaffforimplementingspecificpartsoftheNRC'sregulations,techniquesusedbythestaffinevaluatingspecificproblemsorpostulatedaccidents,anddataneededbytheNRCstaffinitsreviewofapplicationsforpermitsandlicenses.Regulatoryguidesarenotsubstitutesforregulations,andcompliancewiththemisnotrequired.MethodsandsolutionsdifferentfromthosesetoutintheguideswillbeacceptableiftheyprovideabasisforthefindingsrequisitetotheissuanceorcontinuanceofapermitorlicensebytheCommission.Thisguidewasissuedafterconsiderationofcommentsreceivedfromthepublic.Commentsandsuggestionsforimprovementsintheseguidesareencouragedatalltimes,andguideswillberevised,asappropriate,toaccommodatecommentsandtoreflectnewinformationorexperience.WrittencommentsmaybesubmittedtotheRulesandDirectivesBranch,ADM,U.S.NuclearRegulatoryCommission,Washington,DC20555-0001.Regulatoryguidesareissuedintenbroaddivisions:1,PowerReactors;2,ResearchandTestReactors;3,FuelsandMaterialsFacilities;4,EnvironmentalandSiting;5,MaterialsandPlantProtection;6,Products;7,Transportation;8,OccupationalHealth;9,AntitrustandFinancialReview;and10,General.Singlecopiesofregulatoryguides(whichmaybereproduced)maybeobtainedfreeofchargebywritingtheDistributionServicesSection,U.S.NuclearRegulatoryCommission,Washington,DC20555-0001,orbyfaxto(301)415-2289,orbyemailtoDISTRIBUTION@NRC.GOV.Electroniccopiesofthisguide areavailableontheinternetatNRC'shomepageat<WWW.NRC.GOV>intheReferenceLibraryunderRegulatoryGuidesandthroughtheElectronicReadingRoom,asAccessionNumberML003716792,alongwithotherrecentlyissuedguides,atthesamewebsite.U.S.NUCLEARREGULATORYCOMMISSIONJuly2000REGULATORYGUIDEOFFICEOFNUCLEARREGULATORYRESEARCHREGULATORYGUIDE1.183(DraftwasissuedasDG-1081)ALTERNATIVERADIOLOGICALSOURCETERMSFOREVALUATINGDESIGNBASISACCIDENTSATNUCLEARPOWERREACTORS iiAVAILABILITYINFORMATIONSinglecopiesofregulatoryguides,bothactiveanddraft,anddraftNUREGdocumentsmaybeobtainedfreeofchargebywritingtheReproductionandDistributionServicesSection,OCIO, USNRC,Washington,DC20555-0001,orbyemailto<DISTRIBUTION@NRC.GOV>,orbyfaxto(301)415-2289.ActiveguidesmayalsobepurchasedfromtheNationalTechnicalInformation Serviceonastandingorderbasis.DetailsonthisservicemaybeobtainedbywritingNTIS,5285 PortRoyalRoad,Springfield,VA22161.ManyNRCdocumentsareavailableelectronicallyinourReferenceLibraryonourwebsite,<WWW.NRC.GOV>,andthroughourElectronicReadingRoom(ADAMS,orPARS,documentsystem)atthesamesite.CopiesofactiveanddraftguidesandmanyotherNRC documentsareavailableforinspectionorcopyingforafeefromtheNRCPublicDocumentRoom at2120LStreetNW.,Washington,DC;thePDR'smailingaddressisMailStopLL-6, Washington,DC20555;telephone(202)634-3273or(800)397-4209;fax(202)634-3343;emailis

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iiiTABLEOFCONTENTS

A. INTRODUCTION

........................................................1

B. DISCUSSION

...........................................................2 C.REGULATORYPOSITION................................................4 1.IMPLEMENTATIONOFAST..............................................41.1GenericConsiderations..............................................4 1.2ScopeofImplementation.............................................6 1.3ScopeofRequiredAnalyses..........................................7 1.4RiskImplications..................................................10 1.5SubmittalRequirements.............................................101.6FSARRequirements...............................................112.ATTRIBUTESOFANACCEPTABLEAST..................................11 3.ACCIDENTSOURCETERM.............................................123.1FissionProductInventory...........................................123.2ReleaseFractions..................................................13 3.3TimingofReleasePhases...........................................143.4RadionuclideComposition..........................................153.5ChemicalForm....................................................15 3.6FuelDamageinNon-LOCADBAs....................................16

METHODOLOGY

................................164.1OffsiteDoseConsequences..........................................16 4.2ControlRoomDoseConsequences....................................17 4.3OtherDoseConsequences...........................................19 4.4AcceptanceCriteria................................................195.ANALYSISASSUMPTIONSAND

METHODOLOGY

.........................205.1GeneralConsiderations.............................................20 5.2Accident-SpecificAssumptions.......................................22 5.3MeteorologyAssumptions...........................................226.ASSUMPTIONSFOREVALUATINGTHERADIATIONDOSESFOREQUIPMENTQUALIFICATION.......................................................23

D. IMPLEMENTATION

....................................................23 REFERENCES.................................................................24 ivAPPENDICESA.AssumptionsforEvaluatingtheRadiologicalConsequencesofaLWRLoss-of-CoolantAccident..................................................A-1B.AssumptionsforEvaluatingtheRadiologicalConsequencesofaFuelFuelHandlingAccident....................................................B-1C.AssumptionsforEvaluatingtheRadiologicalConsequencesofaBWRRodDropAccident........................................................C-1D.AssumptionsforEvaluatingtheRadiologicalConsequencesofaBWRMainSteamLineBreakAccident.................................................D-1E.AssumptionsforEvaluatingtheRadiologicalConsequencesofaPWRMainSteamLineBreakAccident.................................................E-1F.AssumptionsforEvaluatingtheRadiologicalConsequencesofaPWRMainSteamGeneratorTubeRuptureAccident.......................................F-1G.AssumptionsforEvaluatingtheRadiologicalConsequencesofaPWRLockedRotorAccident...........................................................G-1H.AssumptionsforEvaluatingtheRadiologicalConsequencesofaPWRRodEjectionAccident.........................................................H-1I.AssumptionsforEvaluatingRadiationDosesforEquipmentQualification............I-1 J.AnalysisDecisionChart....................................................J-1 K.Acronyms...............................................................K-1 1Applicantsforaconstructionpermit,adesigncertification,oracombinedlicensethatdonotreferenceastandarddesigncertificationwhoappliedafterJanuary10,1997,arerequiredbyregulationtomeetradiologicalcriteriaprovidedin10CFR 50.34.2Asdefinedin10CFR50.2,designbasesmeansinformationthatidentifiesthespecificfunctionstobeperformedbyastructure,system,orcomponentofafacilityandthespecificvaluesorrangesofvalueschosenforcontrollingparametersasreference boundsfordesign.Thesevaluesmaybe(1)restraintsderivedfromgenerallyaccepted"stateoftheart"practicesforachievingfunctionalgoalsor(2)requirementsderivedfromanalysis(basedoncalculationorexperimentsorboth)oftheeffectsofa postulatedaccidentforwhichastructure,system,orcomponentmustmeetitsfunctionalgoals.TheNRCconsiderstheaccidentsourcetermtobeanintegralpartofthedesignbasisbecauseitsetsforthspecificvalues(orarangeofvalues)forcontrolling parametersthatconstitutereferenceboundsfordesign.1.183-1

A. INTRODUCTION

Thisguideprovidesguidancetolicenseesofoperatingpowerreactorsonacceptableapplicationsofalternativesourceterms;thescope,nature,anddocumentationofassociated analysesandevaluations;considerationofimpactsonanalyzedrisk;andcontentofsubmittals.

Thisguideestablishesanacceptablealternativesourceterm(AST)andidentifiesthesignificant attributesofotherASTsthatmaybefoundacceptablebytheNRCstaff.Thisguidealsoidentifies acceptableradiologicalanalysisassumptionsforuseinconjunctionwiththeacceptedAST.In10CFRPart50,"DomesticLicensingofProductionandUtilizationFacilities,"Section50.34,"ContentsofApplications;TechnicalInformation,"requiresthateachapplicantfora constructionpermitoroperatinglicenseprovideananalysisandevaluationofthedesignand performanceofstructures,systems,andcomponentsofthefacilitywiththeobjectiveofassessing therisktopublichealthandsafetyresultingfromoperationofthefacility.Applicantsarealso requiredby10CFR50.34toprovideananalysisoftheproposedsite.In10CFRPart100,

"ReactorSiteCriteria,"Section100.11,1"DeterminationofExclusionArea,LowPopulationZone,andPopulationCenterDistance,"providescriteriaforevaluatingtheradiologicalaspectsofthe proposedsite.Afootnoteto10CFR100.11statesthatthefissionproductreleaseassumedin theseevaluationsshouldbebaseduponamajoraccidentinvolvingsubstantialmeltdownofthe corewithsubsequentreleaseofappreciablequantitiesoffissionproducts.TechnicalInformationDocument(TID)14844,"CalculationofDistanceFactorsforPowerandTestReactorSites"(Ref.1),iscitedin10CFRPart100asasourceoffurtherguidanceon theseanalyses.Althoughinitiallyusedonlyforsitingevaluations,theTID-14844sourcetermhas beenusedinotherdesignbasisapplications,suchasenvironmentalqualificationofequipment under10CFR50.49,"EnvironmentalQualificationofElectricEquipmentImportanttoSafetyfor NuclearPowerPlants,"andinsomerequirementsrelatedtoThreeMileIsland(TMI)asstatedin NUREG-0737,"ClarificationofTMIActionPlanRequirements"(Ref.2).Theanalysesand evaluationsrequiredby10CFR50.34foranoperatinglicensearedocumentedinthefacilityfinal safetyanalysisreport(FSAR).Fundamentalassumptionsthataredesigninputs,includingthe sourceterm,aretobeincludedintheFSARandbecomepartofthefacilitydesignbasis.2SincethepublicationofTID-14844,significantadvanceshavebeenmadeinunderstandingthetiming,magnitude,andchemicalformoffissionproductreleasesfromseverenuclearpower plantaccidents.AholderofanoperatinglicenseissuedpriortoJanuary10,1997,oraholderofa renewedlicenseunder10CFRPart54whoseinitialoperatinglicensewasissuedpriortoJanuary 1.183-210,1997,isallowedby10CFR50.67,"AccidentSourceTerm,"tovoluntarilyrevisetheaccidentsourcetermusedindesignbasisradiologicalconsequenceanalyses.Ingeneral,informationprovidedbyregulatoryguidesisreflectedinNUREG-0800,theStandardReviewPlan(SRP)(Ref3).TheNRCstaffusestheSRPtoreviewapplicationsto constructandoperatenuclearpowerplants.ThisregulatoryguideappliestoChapter15.0.1ofthe SRP.Theinformationcollectionscontainedinthisregulatoryguidearecoveredbytherequirementsof10CFRPart50,whichwereapprovedbytheOfficeofManagementandBudget (OMB),approvalnumber3150-0011.TheNRCmaynotconductorsponsor,andapersonisnot requiredtorespondto,acollectionofinformationunlessitdisplaysacurrentlyvalidOMBcontrol number.

B. DISCUSSION

Anaccidentsourcetermisintendedtoberepresentativeofamajoraccidentinvolvingsignificantcoredamageandistypicallypostulatedtooccurinconjunctionwithalargeloss-of-coolant accident(LOCA).AlthoughtheLOCAistypicallythemaximumcredibleaccident,NRCstaff experienceinreviewinglicenseapplicationshasindicatedtheneedtoconsiderotheraccident sequencesoflesserconsequencebuthigherprobabilityofoccurrence.Thedesignbasisaccidents (DBAs)werenotintendedtobeactualeventsequences,butrather,wereintendedtobesurrogatesto enabledeterministicevaluationoftheresponseofafacility'sengineeredsafetyfeatures.These accidentanalysesareintentionallyconservativeinordertocompensateforknownuncertaintiesin accidentprogression,fissionproducttransport,andatmosphericdispersion.Althoughprobabilistic riskassessments(PRAs)canprovideusefulinsightsintosystemperformanceandsuggestchangesin howthedesireddepthisachieved,defenseindepthcontinuestobeaneffectivewaytoaccountfor uncertaintiesinequipmentandhumanperformance.TheNRC'spolicystatementontheuseofPRA methods(Ref.4)callsfortheuseofPRAtechnologyinallregulatorymattersinamannerthat complementstheNRC'sdeterministicapproachandsupportsthetraditionaldefense-in-depth philosophy.SincethepublicationofTID-14844(Ref.1),significantadvanceshavebeenmadeinunderstandingthetiming,magnitude,andchemicalformoffissionproductreleasesfromsevere nuclearpowerplantaccidents.In1995,theNRCpublishedNUREG-1465,"AccidentSourceTerms forLight-WaterNuclearPowerPlants"(Ref.5).NUREG-1465usedthisresearchtoprovide estimatesoftheaccidentsourcetermthatweremorephysicallybasedandthatcouldbeappliedtothe designoffuturelight-waterpowerreactors.NUREG-1465presentsarepresentativeaccidentsource termforaboiling-waterreactor(BWR)andforapressurized-waterreactor(PWR).Thesesource termsarecharacterizedbythecompositionandmagnitudeoftheradioactivematerial,thechemical andphysicalpropertiesofthematerial,andthetimingofthereleasetothecontainment.TheNRC staffconsideredtheapplicabilityoftherevisedsourcetermstooperatingreactorsanddeterminedthat thecurrentanalyticalapproachbasedontheTID-14844sourcetermwouldcontinuetobeadequateto protectpublichealthandsafety.Operatingreactorslicensedunderthatapproachwouldnotbe requiredtore-analyzeaccidentsusingtherevisedsourceterms.TheNRCstaffalsodeterminedthat somelicenseesmightwishtouseanASTinanalysestosupportcost-beneficiallicensingaction TheNUREG-1465sourcetermshaveoftenbeenreferredtoasthe"revisedsourceterms."Inrecognitionthattheremaybeadditionalsourcetermsidentifiedinthefuture,10CFR50.67addresses"alternativesourceterms."Thisregulatoryguide endorsesasourcetermderivedfromNUREG-1465andprovidesguidanceontheacceptableattributesofotheralternativesourceterms.1.183-3TheNRCstaff,therefore,initiatedseveralactionstoprovidearegulatorybasisforoperatingreactorstouseanAST3indesignbasisanalyses.Theseinitiativesresultedinthedevelopmentandissuanceof10CFR50.67andthisregulatoryguide.TheNRC'straditionalmethodsforcalculatingtheradiologicalconsequencesofdesignbasisaccidentsaredescribedinaseriesofregulatoryguidesandSRPchapters.Thatguidancewas developedtobeconsistentwiththeTID-14844sourcetermandthewholebodyandthyroiddose guidelinesstatedin10CFR100.11.Manyofthoseanalysisassumptionsandmethodsare inconsistentwiththeASTsandwiththetotaleffectivedoseequivalent(TEDE)criteriaprovidedin10 CFR50.67.ThisguideprovidesassumptionsandmethodsthatareacceptabletotheNRCstafffor performingdesignbasisradiologicalanalysesusinganAST.Thisguidancesupersedescorresponding radiologicalanalysisassumptionsprovidedinotherregulatoryguidesandSRPchapterswhenusedin conjunctionwithanapprovedASTandtheTEDEcriteriaprovidedin10CFR50.67.Theaffected guideswillnotbewithdrawnastheirguidancestillapplieswhenanASTisnotused.Specifically, theaffectedregulatoryguidesare:RegulatoryGuide1.3,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaLossofCoolantAccidentforBoilingWaterReactors"(Ref.6)RegulatoryGuide1.4,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaLossofCoolantAccidentforPressurizedWaterReactors"(Ref.7)RegulatoryGuide1.5,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaSteamLineBreakAccidentforBoilingWaterReactors"(Ref.8)RegulatoryGuide1.25,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaFuelHandlingAccidentintheFuelHandlingandStorageFacilityforBoilingandPressurized WaterReactors"(Ref.9)RegulatoryGuide1.77,"AssumptionsUsedforEvaluatingaControlRodEjectionAccidentforPressurizedWaterReactors"(Ref.10)TheguidanceinRegulatoryGuide1.89,"EnvironmentalQualificationofCertainElectricEquipmentImportanttoSafetyforNuclearPowerPlant."(Ref.11),regardingtheradiologicalsource termusedinthedeterminationofintegrateddosesforenvironmentalqualificationpurposesis supersededbythecorrespondingguidanceinthisregulatoryguideforthosefacilitiesthatare proposingto,orhavealready,implementedanAST.AllotherguidanceinRegulatoryGuide1.89 remainseffective.Thisguideprimarilyaddressesdesignbasisaccidents,suchasthoseaddressedinChapter15oftypicalfinalsafetyanalysisreports(FSARs).Thisguidedoesnotaddressallareasofpotentially significantrisk.Althoughthisguideaddressesfuelhandlingaccidents,othereventsthatcouldoccur duringshutdownoperationsarenotcurrentlyaddressed.TheNRCstaffhasseveralongoing 1.183-4initiativesinvolvingrisksofshutdownoperations,extendedburnupfuels,andrisk-informingcurrentregulations.TheinformationinthisguidemayberevisedinthefutureasNRCstaffevaluationsare completedandregulatorydecisionsontheseissuesaremade.C.REGULATORYPOSITION1.IMPLEMENTATIONOFAST 1.1GenericConsiderationsAsusedinthisguide,anASTisanaccidentsourcetermthatisdifferentfromtheaccidentsourcetermusedintheoriginaldesignandlicensingofthefacilityandthathasbeenapprovedforuse under10CFR50.67.ThisguideidentifiesanASTthatisacceptabletotheNRCstaffandidentifies significantcharacteristicsofotherASTsthatmaybefoundacceptable.WhiletheNRCstaff recognizesseveralpotentialusesofanAST,itisnotpossibletoforeseeallpossibleuses.TheNRC staffwillallowlicenseestopursuetechnicallyjustifiableusesoftheASTsinthemostflexible mannercompatiblewithmaintainingaclear,logical,andconsistentdesignbasis.TheNRCstaffwill approvetheselicenseamendmentrequestsifthefacility,asmodified,willcontinuetoprovide sufficientsafetymarginswithadequatedefenseindepthtoaddressunanticipatedeventsandto compensateforuncertaintiesinaccidentprogressionandanalysisassumptionsandparameterinputs.1.1.1SafetyMarginsTheproposedusesofanASTandtheassociatedproposedfacilitymodificationsandchangestoproceduresshouldbeevaluatedtodeterminewhethertheproposedchangesareconsistentwiththe principlethatsufficientsafetymarginsaremaintained,includingamargintoaccountforanalysis uncertainties.Thesafetymarginsareproductsofspecificvaluesandlimitscontainedinthetechnical specifications(whichcannotbechangedwithoutNRCapproval)andothervalues,suchasassumed accidentortransientinitialconditionsorassumedsafetysystemresponsetimes.Changes,orthenet effectsofmultiplechanges,thatresultinareductioninsafetymarginsmayrequirepriorNRC approval.OncetheinitialASTimplementationhasbeenapprovedbythestaffandhasbecomepart ofthefacilitydesignbasis,thelicenseemayuse10CFR50.59anditssupportingguidancein assessingsafetymarginsrelatedtosubsequentfacilitymodificationsandchangestoprocedures.1.1.2DefenseinDepthTheproposedusesofanASTandtheassociatedproposedfacilitymodificationsandchangestoproceduresshouldbeevaluatedtodeterminewhethertheproposedchangesareconsistentwiththe principlethatadequatedefenseindepthismaintainedtocompensateforuncertaintiesinaccident progressionandanalysisdata.Consistencywiththedefense-in-depthphilosophyismaintainedif systemredundancy,independence,anddiversityarepreservedcommensuratewiththeexpected frequency,consequencesofchallengestothesystem,anduncertainties.Inallcases,compliancewith theGeneralDesignCriteriainAppendixAto10CFRPart50isessential.Modificationsproposed forthefacilitygenerallyshouldnotcreateaneedforcompensatoryprogrammaticactivities,suchas relianceonmanualoperatoractions.Proposedmodificationsthatseektodowngradeorremoverequiredengineeredsafeguardsequipmentshouldbeevaluatedtobesurethatthemodificationdoesnotinvalidateassumptionsmade infacilityPRAsanddoesnotadverselyimpactthefacility'ssevereaccidentmanagementprogra ThisplanningbasisisalsoaddressedinNUREG-0654,"CriteriaforPreparationandEvaluationofRadiologicalEmergencyResponsePlansandPreparednessinSupportofNuclearPowerPlants"(Ref.13).1.183-51.1.3IntegrityofFacilityDesignBasisThedesignbasisaccidentsourcetermisafundamentalassumptionuponwhichasignificantportionofthefacilitydesignisbased.Additionally,manyaspectsoffacilityoperationderivefrom thedesignanalysesthatincorporatedtheearlieraccidentsourceterm.Althoughacompletere- assessmentofallfacilityradiologicalanalyseswouldbedesirable,theNRCstaffdeterminedthat recalculationofalldesignanalyseswouldgenerallynotbenecessary.RegulatoryPosition1.3ofthis guideprovidesguidanceonwhichanalysesneedupdatingaspartoftheASTimplementation submittalandwhichmayneedupdatinginthefutureasadditionalmodificationsareperformed.Thisapproachwouldcreatetwotiersofanalyses,thosebasedontheprevioussourcetermandthosebasedonanAST.Theradiologicalacceptancecriteriawouldalsobedifferentwithsome analysesbasedonwholebodyandthyroidcriteriaandsomebasedonTEDEcriteria.Full implementationoftheASTrevisestheplantlicensingbasistospecifytheASTinplaceofthe previousaccidentsourcetermandestablishestheTEDEdoseasthenewacceptancecriteria.

SelectiveimplementationoftheASTalsorevisestheplantlicensingbasisandmayestablishthe TEDEdoseasthenewacceptancecriteria.Selectiveimplementationdiffersfromfull implementationonlyinthescopeofthechange.Ineithercase,thefacilitydesignbasesshouldclearly indicatethatthesourcetermassumptionsandradiologicalcriteriaintheseaffectedanalyseshave beensupersededandthatfuturerevisionsoftheseanalyses,ifany,willusetheupdatedapproved assumptionsandcriteria.Radiologicalanalysesgenerallyshouldbebasedonassumptionsandinputsthatareconsistentwithcorrespondingdatausedinotherdesignbasissafetyanalyses,radiologicalandnonradiological, unlessthesedatawouldresultinnonconservativeresultsorotherwiseconflictwiththeguidancein thisguide.1.1.4EmergencyPreparednessApplicationsRequirementsforemergencypreparednessatnuclearpowerplantsaresetforthin10CFR50.47,"EmergencyPlans."AdditionalrequirementsaresetforthinAppendixE,"Emergency PlanningandPreparednessforProductionandUtilizationFacilities,"to10CFRPart50.The planningbasisformanyoftheserequirementswaspublishedinNUREG-0396,"PlanningBasisfor theDevelopmentofStateandLocalGovernmentRadiologicalEmergencyResponsePlansinSupport ofLightWaterNuclearPowerPlants"4(Ref.12).ThisjointeffortbytheEnvironmentalProtectionAgency(EPA)andtheNRCconsideredtheprincipalcharacteristics(suchasnuclidesreleasedand distances)likelytobeinvolvedforaspectrumofdesignbasisandsevere(coremelt)accidents.No singleaccidentscenarioisthebasisoftherequiredpreparedness.Theobjectiveoftheplanningisto providepublicprotectionthatwouldencompassawidespectrumofpossibleeventswithasufficient basisforextensionofresponseeffortsforunanticipatedevents.Theserequirementswereissuedafter alongperiodofinvolvementbynumerousstakeholders,includingtheFederalEmergency ManagementAgency,otherFederalagencies,localandStategovernments(andinsomecases,foreign governments),privatecitizens,utilities,andindustrygroups.AlthoughtheASTprovidedinthisguidewasbasedonalimitedspectrumofsevereaccidents,theparticularcharacteristicshavebeentailoredspecificallyforDBAanalysisuse.TheASTisnot 1.183-6representativeofthewidespectrumofpossibleeventsthatmakeuptheplanningbasisofemergencypreparedness.Therefore,theASTisinsufficientbyitselfasabasisforrequestingrelieffromtheemergencypreparednessrequirementsof10CFR50.47andAppendixEto10CFRPart50.Thisguidancedoesnot,however,precludetheappropriateuseoftheinsightsoftheASTinestablishingemergencyresponseproceduressuchasthoseassociatedwithemergencydose projections,protectivemeasures,andsevereaccidentmanagementguides.1.2ScopeofImplementationTheASTdescribedinthisguideischaracterizedbyradionuclidecompositionandmagnitude,chemicalandphysicalformoftheradionuclides,andthetimingofthereleaseoftheseradionuclides.

Theaccidentsourcetermisafundamentalassumptionuponwhichalargeportionofthefacility designisbased.Additionally,manyaspectsoffacilityoperationderivefromthedesignanalysesthat incorporatedtheearlieraccidentsourceterm.AcompleteimplementationofanASTwouldupgrade allexistingradiologicalanalysesandwouldconsidertheimpactofallfivecharacteristicsoftheAST asdefinedin10CFR50.2.However,theNRCstaffhasdeterminedthattherecouldbe implementationsforwhichthislevelofre-analysismaynotbenecessary.Twocategoriesare defined:Fullandselectiveimplementations.1.2.1FullImplementationFullimplementationisamodificationofthefacilitydesignbasisthataddressesallcharacteristicsoftheAST,thatis,compositionandmagnitudeoftheradioactivematerial,its chemicalandphysicalform,andthetimingofitsrelease.Fullimplementationrevisestheplant licensingbasistospecifytheASTinplaceofthepreviousaccidentsourcetermandestablishesthe TEDEdoseasthenewacceptancecriteria.Thisappliesnotonlytotheanalysesperformedinthe application(whichmayonlyincludeasubsetoftheplantanalyses),butalsotoallfuturedesignbasis analyses.Ataminimumforfullimplementations,theDBALOCAmustbere-analyzedusingthe guidanceinAppendixAofthisguide.AdditionalguidanceonanalysisisprovidedinRegulatory Position1.3ofthisguide.SincetheASTandTEDEcriteriawouldbecomepartofthefacilitydesign basis,newapplicationsoftheASTwouldnotrequirepriorNRCapprovalunlessstipulatedby10 CFR50.59,"Changes,Tests,andExperiments,"orunlessthenewapplicationinvolvedachangetoa technicalspecification.However,achangefromanapprovedASTtoadifferentASTthatisnot approvedforuseatthatfacilitywouldrequirealicenseamendmentunder10CFR50.67.1.2.2SelectiveImplementationSelectiveimplementationisamodificationofthefacilitydesignbasisthat(1)isbasedononeormoreofthecharacteristicsoftheASTor(2)entailsre-evaluationofalimitedsubsetofthedesign basisradiologicalanalyses.TheNRCstaffwillallowlicenseesflexibilityintechnicallyjustified selectiveimplementationsprovidedaclear,logical,andconsistentdesignbasisismaintained.An exampleofanapplicationofselectiveimplementationwouldbeoneinwhichalicenseedesirestouse thereleasetiminginsightsoftheASTtoincreasetherequiredclosuretimeforacontainmentisolation valvebyasmallamount.Anotherexamplewouldbearequesttoremovethecharcoalfiltermedia fromthespentfuelbuildingventilationexhaust.Forthelatter,thelicenseemayonlyneedtore- analyzeDBAsthatcreditedtheiodineremovalbythecharcoalmedia.Additionalanalysisguidance isprovidedinRegulatoryPosition1.3ofthisguide.NRCapprovalfortheAST(andtheTEDEdose criterion)willbelimitedtotheparticularselectiveimplementationproposedbythelicensee.The 5Doseguidelinesof10CFR100.11aresupersededby10CFR50.67forlicenseesthathaveimplementedanAST.1.183-7licenseewouldbeabletomakesubsequentmodificationstothefacilityandchangestoproceduresbasedontheselectedASTcharacteristicsincorporatedintothedesignbasisundertheprovisionsof 10CFR50.59.However,useofothercharacteristicsofanASToruseofTEDEcriteriathatarenot partoftheapproveddesignbasis,andchangestopreviouslyapprovedASTcharacteristics,would requirepriorstaffapprovalunder10CFR50.67.Asanexample,alicenseewithanimplementation involvingonlytiming,suchasrelaxedclosuretimeonisolationvalves,couldnotuse10CFR50.59 asamechanismtoimplementamodificationinvolvingareanalysisoftheDBALOCA.However, thislicenseecouldextenduseofthetimingcharacteristictoadjusttheclosuretimeonisolation valvesnotincludedintheoriginalapproval.1.3ScopeofRequiredAnalyses1.3.1DesignBasisRadiologicalAnalysesThereareseveralregulatoryrequirementsforwhichcomplianceisdemonstrated,inpart,bytheevaluationoftheradiologicalconsequencesofdesignbasisaccidents.Theserequirements include,butarenotlimitedto,thefollowing.EnvironmentalQualificationofEquipment(10CFR50.49)ControlRoomHabitability(GDC-19ofAppendixAto10CFRPart50)EmergencyResponseFacilityHabitability(ParagraphIV.E.8ofAppendixEto10CFRPart50)AlternativeSourceTerm(10CFR50.67)EnvironmentalReports(10CFRPart51)FacilitySiting(10CFR100.11)5Theremaybeadditionalapplicationsoftheaccidentsourcetermidentifiedinthetechnicalspecificationbasesandinvariouslicenseecommitments.Theseinclude,butarenotlimitedto,the followingfromReference2,NUREG-0737.Post-AccidentAccessShielding(NUREG-0737,II.B.2)Post-AccidentSamplingCapability(NUREG-0737,II.B.3)AccidentMonitoringInstrumentation(NUREG-0737,II.F.1)LeakageControl(NUREG-0737,III.D.1.1)EmergencyResponseFacilities(NUREG-0737,III.A.1.2)ControlRoomHabitability(NUREG-0737,III.D.3.4)1.3.2Re-AnalysisGuidanceAnyimplementationofanAST,fullorselective,andanyassociatedfacilitymodificationshouldbesupportedbyevaluationsofallsignificantradiologicalandnonradiologicalimpactsof theproposedactions.Thisevaluationshouldconsidertheimpactoftheproposedchangesonthe facility'scompliancewiththeregulationsandcommitmentslistedaboveaswellasanyother facility-specificrequirements.Theseimpactsmaybedueto(1)theassociatedfacility modificationsor(2)thedifferencesintheASTcharacteristics.Thescopeandextentofthere-

6Forexample,aproposedmodificationtochangethetimingofacontainmentisolationvalvefrom2.5secondsto5.0secondsmightbeacceptablewithoutanydosecalculations.However,aproposedmodificationthatwoulddelaycontainmentsprayactuationcouldinvolverecalculationofDBALOCAdoses,re-assessmentofthecontainmentpressureandtemperaturetransient, recalculationofsumppH,re-assessmentoftheemergencydieselgeneratorloadingsequence,integrateddosestoequipmentin thecontainment,andmore.1.183-8evaluationwillnecessarilybeafunctionofthespecificproposedfacilitymodification6andwhetherafullorselectiveimplementationisbeingpursued.TheNRCstaffdoesnotexpecta completerecalculationofallfacilityradiologicalanalyses,butdoesexpectlicenseestoevaluateall impactsoftheproposedchangesandtoupdatetheaffectedanalysesandthedesignbases appropriately.Ananalysisisconsideredtobeaffectediftheproposedmodificationchangesone ormoreassumptionsorinputsusedinthatanalysissuchthattheresults,ortheconclusionsdrawn onthoseresults,arenolongervalid.Genericanalyses,suchasthoseperformedbyownergroups orvendortopicalreports,maybeusedprovidedthelicenseejustifiestheapplicabilityofthe genericconclusionstothespecificfacilityandimplementation.Sensitivityanalyses,discussed below,mayalsobeanoption.Ifaffecteddesignbasisanalysesaretobere-calculated,allaffected assumptionsandinputsshouldbeupdatedandallselectedcharacteristicsoftheASTandthe TEDEcriteriashouldbeaddressed.Thelicenseamendmentrequestshoulddescribethelicensee's re-analysiseffortandprovidestatementsregardingtheacceptabilityoftheproposed implementation,includingmodifications,againsteachoftheapplicableanalysisrequirementsand commitmentsidentifiedinRegulatoryPosition1.3.1ofthisguide.TheNRCstaffhasperformedanevaluationoftheimpactoftheASTonthreerepresentativeoperatingreactors(Ref.14).Thisevaluationdeterminedthatradiologicalanalysis resultsbasedontheTID-14844sourcetermassumptions(Ref.1)andthewholebodyandthyroid methodologygenerallyboundtheresultsfromanalysesbasedontheASTandTEDEmethodology.

Licenseesmayusetheapplicableconclusionsofthisevaluationinaddressingtheimpactofthe ASTondesignbasisradiologicalanalyses.However,thisdoesnotexemptthelicenseefrom evaluatingtheremainingradiologicalandnonradiologicalimpactsoftheASTimplementationand theimpactsoftheassociatedplantmodifications.Forexample,aselectiveimplementationbased onthetiminginsightsoftheASTmaychangetherequiredisolationtimeforthecontainment purgedampersfrom2.5secondsto5.0seconds.Thisapplicationmightbeacceptablewithout dosecalculations.However,evaluationsmayneedtobeperformedregardingtheabilityofthe dampertocloseagainstincreasedcontainmentpressureortheabilityofductworkdownstreamof thedamperstowithstandincreasedstresses.Forfullimplementation,acompleteDBALOCAanalysisasdescribedinAppendixAofthisguideshouldbeperformed,asaminimum.Otherdesignbasisanalysesareupdatedin accordancewiththeguidanceinthissection.AselectiveimplementationofanASTandanyassociatedfacilitymodificationbasedontheASTshouldevaluatealltheradiologicalandnonradiologicalimpactsoftheproposedactions astheyapplytotheparticularimplementation.Designbasisanalysesareupdatedinaccordance withtheguidanceinthissection.ThereisnominimumrequirementthataDBALOCAanalysisbe performed.Theanalysesperformedneedtoaddressallimpactsoftheproposedmodification,the selectedcharacteristicsoftheAST,andifdosecalculationsareperformed,theTEDEcriteria.For selectiveimplementationsbasedonthetimingcharacteristicoftheAST,e.g.,changeintheclosure timingofacontainmentisolationvalve,re-analysisofradiologicalcalculationsmaynotbe 7Inperformingscreeningsandevaluationspursuantto10CFR50.59,itmaybenecessarytocomparedoseresultsexpressedintermsofwholebodyandthyroidwithnewresultsexpressedintermsofTEDE.Inthesecases,thepreviousthyroiddoseshould bemultipliedby0.03andtheproductaddedtothewholebodydose.TheresultisthencomparedtotheTEDEresultinthe screeningsandevaluations.Thischangeindosemethodologyisnotconsideredachangeinthemethodofevaluationifthe licenseewaspreviouslyauthorizedtouseanASTandtheTEDEcriteriaunder10CFR50.67.1.183-9necessaryifthemodifiedelapsedtimeremainsafraction(e.g.,25%)ofthetimebetweenaccidentinitiationandtheonsetofthegapreleasephase.Longertimedelaysmaybeconsideredonan individualbasis.Forlongertimedelays,evaluationoftheradiologicalconsequencesandother impactsofthedelay,suchasblockagebydebrisinsumpwater,maybenecessary.Ifaffected designbasisanalysesaretobere-calculated,allaffectedassumptionsandinputsshouldbeupdated andallselectedcharacteristicsoftheASTandtheTEDEcriteriashouldbeaddressed.1.3.3UseofSensitivityorScopingAnalysesItmaybepossibletodemonstratebysensitivityorscopingevaluationsthatexistinganalyseshavesufficientmarginandneednotberecalculated.Asusedinthisguide,asensitivityanalysisisanevaluationthatconsidershowtheoverallresultsvaryasaninputparameter(inthiscase,ASTcharacteristics)isvaried.Ascopinganalysisisabriefevaluationthatusesconservative,simplemethodstoshowthattheresultsoftheanalysisboundthoseobtainablefrom amorecompletetreatment.Sensitivityanalysesareparticularlyapplicabletosuitesofcalculations thataddressdiversecomponentsorplantareasbutareotherwiselargelybasedongeneric assumptionsandinputs.Suchcasesmightincludepostaccidentvitalareaaccessdosecalculations, shieldingcalculations,andequipmentenvironmentalqualification(integrateddose).Itmaybe possibletoidentifyaboundingcase,re-analyzethatcase,andusetheresultstodrawconclusions regardingtheremainderoftheanalyses.Itmayalsobepossibletoshowthatforsomeanalysesthe wholebodyandthyroiddosesdeterminedwiththeprevioussourcetermwouldboundtheTEDE obtainedusingtheAST.Wherepresent,arbitrary"designermargins"maybeadequatetobound anyimpactoftheASTandTEDEcriteria.Ifsensitivityorscopinganalysesareused,thelicense amendmentrequestshouldincludeadiscussionoftheanalysesperformedandtheconclusions drawn.Scopingorsensitivityanalysesshouldnotconstituteasignificantpartoftheevaluations forthedesignbasisexclusionareaboundary(EAB),lowpopulationzone(LPZ),orcontrolroom dose.1.3.4UpdatingAnalysesFollowingImplementationFullimplementationoftheASTreplacesthepreviousaccidentsourcetermwiththeapprovedASTandtheTEDEcriteriaforalldesignbasisradiologicalanalyses.The implementationmayhavebeensupportedinpartbysensitivityorscopinganalysesthatconcluded manyofthedesignbasisradiologicalanalyseswouldremainboundingfortheASTandtheTEDE criteriaandwouldnotrequireupdating.Aftertheimplementationiscomplete,theremaybea subsequentneed(e.g.,aplannedfacilitymodification)torevisetheseanalysesortoperformnew analyses.Fortheserecalculations,theNRCstaffexpectsthatallcharacteristicsoftheASTandthe TEDEcriteriaincorporatedintothedesignbasiswillbeaddressedinallaffectedanalysesonan individualas-neededbasis.Re-evaluationusingthepreviouslyapprovedsourcetermmaynotbe appropriate.SincetheASTandtheTEDEcriteriaarepartoftheapproveddesignbasisforthe facility,useoftheASTandTEDEcriteriainnewapplicationsatthefacilitydonotconstitutea changeinanalysismethodologythatwouldrequireNRCapprova .183-10Thisguidanceisalsoapplicabletoselectiveimplementationstotheextentthattheaffectedanalysesarewithinthescopeoftheapprovedimplementationasdescribedinthefacilitydesign basis.Inthesecases,thecharacteristicsoftheASTandTEDEcriteriaidentifiedinthefacility designbasisneedtobeconsideredinupdatingtheanalyses.Useofothercharacteristicsofthe ASTorTEDEcriteriathatarenotpartoftheapproveddesignbasis,andchangestopreviously approvedASTcharacteristics,requirespriorNRCstaffapprovalunder10CFR50.67.1.3.5EquipmentEnvironmentalQualificationCurrentenvironmentalqualification(EQ)analysesmaybeimpactedbyaproposedplantmodificationassociatedwiththeASTimplementation.TheEQanalysesthathaveassumptionsor inputsaffectedbytheplantmodificationshouldbeupdatedtoaddresstheseimpacts.TheNRC staffisassessingtheeffectofincreasedcesiumreleasesonEQdosestodeterminewhether licenseeactioniswarranted.Untilsuchtimeasthisgenericissueisresolved,licenseesmayuse eithertheASTortheTID14844assumptionsforperformingtherequiredEQanalyses.However, noplantmodificationsarerequiredtoaddresstheimpactofthedifferenceinsourceterm characteristics(i.e.,ASTvsTID14844)onEQdosespendingtheoutcomeoftheevaluationofthe genericissue.TheEQdoseestimatesshouldbecalculatedusingthedesignbasissurvivability period.1.4RiskImplicationsTheuseofanASTchangesonlytheregulatoryassumptionsregardingtheanalyticaltreatmentofthedesignbasisaccidents.TheASThasnodirecteffectontheprobabilityofthe accident.UseofanASTalonecannotincreasethecoredamagefrequency(CDF)orthelargeearly releasefrequency(LERF).However,facilitymodificationsmadepossiblebytheASTcouldhave animpactonrisk.IftheproposedimplementationoftheASTinvolveschangestothefacility designthatwouldinvalidateassumptionsmadeinthefacility'sPRA,theimpactontheexisting PRAsshouldbeevaluated.Considerationshouldbegiventotheriskimpactofproposedimplementationsthatseektoremoveordowngradetheperformanceofpreviouslyrequiredengineeredsafeguardsequipmenton thebasisofthereducedpostulateddoses.TheNRCstaffmayrequestriskinformationifthereisa reasontoquestionadequateprotectionofpublichealthandsafety.ThelicenseemayelecttouseriskinsightsinsupportofproposedchangestothedesignbasisthatarenotaddressedincurrentlyapprovedNRCstaffpositions.Forguidance,referto RegulatoryGuide1.174,"AnApproachforUsingProbabilisticRiskAssessmentinRisk-Informed DecisionsonPlant-SpecificChangestotheLicensingBasis"(Ref.15).1.5SubmittalRequirementsAccordingto10CFR50.90,anapplicationforanamendmentmustfullydescribethechangesdesiredandshouldfollow,asfarasapplicable,theformprescribedfororiginal applications.RegulatoryGuide1.70,"StandardFormatandContentofSafetyAnalysisReports forNuclearPowerPlants(LWREdition)"(Ref16),providesadditionalguidance.TheNRC staff'sfindingthattheamendmentmaybeapprovedmustbebasedonthelicensee'sanalyses, 1.183-11sinceitistheseanalysesthatwillbecomepartofthedesignbasisofthefacility.Theamendmentrequestshoulddescribethelicensee'sanalysesoftheradiologicalandnonradiologicalimpactsof theproposedmodificationinsufficientdetailtosupportreviewbytheNRCstaff.Thestaff recommendsthatlicenseessubmitaffectedFSARpagesannotatedwithchangesthatreflectthe revisedanalysesorsubmittheactualcalculationdocumentation.IfthelicenseehasusedacurrentapprovedversionofanNRC-sponsoredcomputercode,theNRCstaffreviewcanbemademoreefficientifthelicenseeidentifiesthecodeusedand submitstheinputsthatthelicenseeusedinthecalculationsmadewiththatcode.Inmanycases, thiswillreducetheneedforNRCstaffconfirmatoryanalyses.Thisrecommendationdoesnot constitutearequirementthatthelicenseeuseNRC-sponsoredcomputercodes.1.6FSARRequirementsRequirementsforupdatingthefacility'sfinalsafetyanalysisreport(FSAR)arein10CFR50.71,"MaintenanceofRecords,MakingofReports."Theregulationsin10CFR50.71(e)require thattheFSARbeupdatedtoincludeallchangesmadeinthefacilityorproceduresdescribedinthe FSARandallsafetyevaluationsperformedbythelicenseeinsupportofrequestsforlicense amendmentsorinsupportofconclusionsthatchangesdidnotinvolveunreviewedsafety questions.Theanalysesrequiredby10CFR50.67aresubjecttothisrequirement.Theaffected radiologicalanalysisdescriptionsintheFSARshouldbeupdatedtoreflectthereplacementofthe designbasissourcetermbytheAST.Theanalysisdescriptionsshouldcontainsufficientdetailto identifythemethodologiesused,significantassumptionsandinputs,andnumericresults.

RegulatoryGuide1.70(Ref.16)providesadditionalguidance.Thedescriptionsofsuperseded analysesshouldberemovedfromtheFSARintheinterestofmaintainingacleardesignbasis.2.ATTRIBUTESOFANACCEPTABLEASTAnacceptableASTisnotsetforthin10CFR50.67.RegulatoryPosition3ofthisguideidentifiesanASTthatisacceptabletotheNRCstaffforuseatoperatingpowerreactors.A substantialeffortwasexpendedbytheNRC,itscontractors,variousnationallaboratories,peer reviewers,andothersinperformingsevereaccidentresearchandindevelopingthesourceterms providedinNUREG-1465(Ref.5).However,futureresearchmayidentifyopportunitiesfor changesinthesesourceterms.TheNRCstaffwillconsiderapplicationsforanASTdifferentfrom thatidentifiedinthisguide.However,theNRCstaffdoesnotexpecttoapproveanysourceterm thatisnotofthesamelevelofqualityasthesourcetermsinNUREG-1465.Tobeconsidered acceptable,anASTmusthavethefollowingattributes:2.1TheASTmustbebasedonmajoraccidents,hypothesizedforthepurposesofdesignanalysesorconsiderationofpossibleaccidentalevents,thatcouldresultinhazardsnot exceededbythosefromotheraccidentsconsideredcredible.TheASTmustaddressevents thatinvolveasubstantialmeltdownofthecorewiththesubsequentreleaseofappreciable quantitiesoffissionproduct TheuncertaintyfactorusedindeterminingthecoreinventoryshouldbethatvalueprovidedinAppendixKto10CFRPart50,typically1.02.9Notethatforsomeradionuclides,suchasCs-137,equilibriumwillnotbereachedpriortofueloffload.Thus,themaximuminventoryattheendoflifeshouldbeused.1.183-122.2TheASTmustbeexpressedintermsoftimesandratesofappearanceofradioactivefissionproductsreleasedintocontainment,thetypesandquantitiesoftheradioactivespecies released,andthechemicalformsofiodinereleased.2.3TheASTmustnotbebaseduponasingleaccidentscenariobutinsteadmustrepresentaspectrumofcrediblesevereaccidentevents.Riskinsightsmaybeused,nottoselecta singlerisk-significantaccident,butrathertoestablishtherangeofeventstobeconsidered.

Relevantinsightsfromapplicablesevereaccidentresearchonthephenomenologyof fissionproductreleaseandtransportbehaviormaybeconsidered.2.4TheASTmusthaveadefensibletechnicalbasissupportedbysufficientexperimentalandempiricaldata,beverifiedandvalidated,andbedocumentedinascrutableformthat facilitatespublicreviewanddiscourse.2.5TheASTmustbepeer-reviewedbyappropriatelyqualifiedsubjectmatterexperts.Thepeer-reviewcommentsandtheirresolutionshouldbepartofthedocumentationsupporting theAST.3.ACCIDENTSOURCETERMThissectionprovidesanASTthatisacceptabletotheNRCstaff.ThedatainRegulatoryPositions3.2through3.5arefundamentaltothedefinitionofanAST.Onceapproved,theAST assumptionsorparametersspecifiedinthesepositionsbecomepartofthefacility'sdesignbasis.

DeviationsfromthisguidancemustbeevaluatedagainstRegulatoryPosition2.AftertheNRC staffhasapprovedanimplementationofanAST,subsequentchangestotheASTwillrequireNRC staffreviewunder10CFR50.67.3.1FissionProductInventoryTheinventoryoffissionproductsinthereactorcoreandavailableforreleasetothecontainmentshouldbebasedonthemaximumfullpoweroperationofthecorewith,asa minimum,currentlicensedvaluesforfuelenrichment,fuelburnup,andanassumedcorepower equaltothecurrentlicensedratedthermalpowertimestheECCSevaluationuncertainty.8Theperiodofirradiationshouldbeofsufficientdurationtoallowtheactivityofdose-significant radionuclidestoreachequilibriumortoreachmaximumvalues.9ThecoreinventoryshouldbedeterminedusinganappropriateisotopegenerationanddepletioncomputercodesuchasORIGEN 2(Ref.17)orORIGEN-ARP(Ref.18).Coreinventoryfactors(Ci/MWt)providedinTID14844 andusedinsomeanalysiscomputercodeswerederivedforlowburnup,lowenrichmentfueland shouldnotbeusedwithhigherburnupandhigherenrichmentfuel ThereleasefractionslistedherehavebeendeterminedtobeacceptableforusewithcurrentlyapprovedLWRfuelwithapeakburnupupto62,000MWD/MTU.Thedatainthissectionmaynotbeapplicabletocorescontainingmixedoxide(MOX)fuel.1.183-13FortheDBALOCA,allfuelassembliesinthecoreareassumedtobeaffectedandthecoreaverageinventoryshouldbeused.ForDBAeventsthatdonotinvolvetheentirecore,thefission productinventoryofeachofthedamagedfuelrodsisdeterminedbydividingthetotalcore inventorybythenumberoffuelrodsinthecore.Toaccountfordifferencesinpowerlevelacross thecore,radialpeakingfactorsfromthefacility'scoreoperatinglimitsreport(COLR)ortechnical specificationsshouldbeappliedindeterminingtheinventoryofthedamagedrods.Noadjustmenttothefissionproductinventoryshouldbemadeforeventspostulatedtooccurduringpoweroperationsatlessthanfullratedpowerorthosepostulatedtooccuratthe beginningofcorelife.Foreventspostulatedtooccurwhilethefacilityisshutdown,e.g.,afuel handlingaccident,radioactivedecayfromthetimeofshutdownmaybemodeled.3.2ReleaseFractions10Thecoreinventoryreleasefractions,byradionuclidegroups,forthegapreleaseandearlyin-vesseldamagephasesforDBALOCAsarelistedinTable1forBWRsandTable2forPWRs.

ThesefractionsareappliedtotheequilibriumcoreinventorydescribedinRegulatoryPosition3.1.Fornon-LOCAevents,thefractionsofthecoreinventoryassumedtobeinthegapforthevariousradionuclidesaregiveninTable3.ThereleasefractionsfromTable3areusedin conjunctionwiththefissionproductinventorycalculatedwiththemaximumcoreradialpeaking factor.Table1BWRCoreInventoryFractionReleasedIntoContainmentGapEarly ReleaseIn-vesselGroupPhasePhaseTotalNobleGases0.050.951.0Halogens0.050.250.3 AlkaliMetals0.050.200.25 TelluriumMetals0.000.050.05Ba,Sr0.000.020.02NobleMetals0.000.00250.0025 CeriumGroup0.000.00050.0005 Lanthanides0.000.00020.0002 11ThereleasefractionslistedherehavebeendeterminedtobeacceptableforusewithcurrentlyapprovedLWRfuelwithapeakburnupupto62,000MWD/MTUprovidedthatthemaximumlinearheatgenerationratedoesnotexceed6.3kw/ftpeakrod averagepowerforburnupsexceeding54GWD/MTU.Asanalternative,fissiongasreleasecalculationsperformedusingNRC-approvedmethodologiesmaybeconsideredonacase-by-casebasis.Tobeacceptable,thesecalculationsmustuseaprojectedpowerhistorythatwillboundthelimitingprojectedplant-specificpowerhistoryforthespecificfuelload.FortheBWRrod dropaccidentandthePWRrodejectionaccident,thegapfractionsareassumedtobe10%foriodinesandnoblegases.12Inlieuoftreatingthereleaseinalinearrampmanner,theactivityforeachphasecanbemodeledasbeingreleasedinstantaneouslyatthestartofthatreleasephase,i.e.,instepincreases.1.183-14Table2PWRCoreInventoryFractionReleasedIntoContainmentGapEarlyReleaseIn-vesselGroupPhasePhaseTotalNobleGases0.050.951.0Halogens0.050.350.4 AlkaliMetals0.050.250.3 TelluriumMetals0.000.050.05 Ba,Sr0.000.020.02 NobleMetals0.000.00250.0025 CeriumGroup0.000.00050.0005 Lanthanides0.000.00020.0002Table311Non-LOCAFractionofFissionProductInventoryinGapGroupFraction I-1310.08 Kr-850.10 OtherNobleGases0.05 OtherHalogens0.05 AlkaliMetals0.123.3TimingofReleasePhasesTable4tabulatestheonsetanddurationofeachsequentialreleasephaseforDBALOCAsatPWRsandBWRs.Thespecifiedonsetisthetimefollowingtheinitiationoftheaccident(i.e.,

time=0).Theearlyin-vesselphaseimmediatelyfollowsthegapreleasephase.Theactivity releasedfromthecoreduringeachreleasephaseshouldbemodeledasincreasinginalinear fashionoverthedurationofthephase.12Fornon-LOCADBAsinwhichfueldamageisprojected,thereleasefromthefuelgapandthefuelpelletshouldbeassumedtooccurinstantaneouslywith theonsetoftheprojecteddamag .183-15Table4LOCAReleasePhasesPWRsBWRsPhaseOnsetDurationOnsetDurationGapRelease30sec0.5hr2min0.5hrEarlyIn-Vessel0.5hr1.3hr0.5hr1.5hrForfacilitieslicensedwithleak-before-breakmethodology,theonsetofthegapreleasephasemaybeassumedtobe10minutes.Alicenseemayproposeanalternativetimefortheonset ofthegapreleasephase,basedonfacility-specificcalculationsusingsuitableanalysiscodesoron anacceptedtopicalreportshowntobeapplicabletothespecificfacility.Intheabsenceof approvedalternatives,thegapreleasephaseonsetsinTable4shouldbeused.3.4RadionuclideCompositionTable5liststheelementsineachradionuclidegroupthatshouldbeconsideredindesignbasisanalyses.Table5RadionuclideGroupsGroupElementsNobleGasesXe,KrHalogensI,Br AlkaliMetalsCs,Rb TelluriumGroupTe,Sb,Se,Ba,Sr NobleMetalsRu,Rh,Pd,Mo,Tc,Co LanthanidesLa,Zr,Nd,Eu,Nb,Pm,PrSm,Y,Cm,AmCeriumCe,Pu,Np3.5ChemicalFormOftheradioiodinereleasedfromthereactorcoolantsystem(RCS)tothecontainmentinapostulatedaccident,95percentoftheiodinereleasedshouldbeassumedtobecesiumiodide(CsI),

4.85percentelementaliodine,and0.15percentorganiciodide.Thisincludesreleasesfromthe gapandthefuelpellets.Withtheexceptionofelementalandorganiciodineandnoblegases, fissionproductsshouldbeassumedtobeinparticulateform.Thesamechemicalformisassumed inreleasesfromfuelpinsinFHAsandfromreleasesfromthefuelpinsthroughtheRCSinDBAs otherthanFHAsorLOCAs.However,thetransportoftheseiodinespeciesfollowingreleasefrom thefuelmayaffecttheseassumedfractions.Theaccident-specificappendicestothisregulatory guideprovideadditionaldetail ThepriorpracticeofbasinginhalationexposureononlyradioiodineandnotincludingradioiodineinexternalexposurecalculationsisnotconsistentwiththedefinitionofTEDEandthecharacteristicsoftherevisedsourceterm.1.183-163.6FuelDamageinNon-LOCADBAsTheamountoffueldamagecausedbynon-LOCAdesignbasiseventsshouldbeanalyzedtodetermine,forthecaseresultinginthehighestradioactivityrelease,thefractionofthefuelthat reachesorexceedstheinitiationtemperatureoffuelmeltandthefractionoffuelelementsfor whichthefuelcladisbreached.AlthoughtheNRCstaffhastraditionallyrelieduponthedeparture fromnucleateboilingratio(DNBR)asafueldamagecriterion,licenseesmayproposeother methodstotheNRCstaff,suchasthosebaseduponenthalpydeposition,forestimatingfuel damageforthepurposeofestablishingradioactivityreleases.TheamountoffueldamagecausedbyaFHAisaddressedinAppendixBofthisguide.

METHODOLOGY

TheNRCstaffhasdeterminedthatthereisanimpliedsynergybetweentheASTsandtotaleffectivedoseequivalent(TEDE)criteria,andbetweentheTID-14844sourcetermsandthewhole bodyandthyroiddosecriteria,andtherefore,theydonotexpecttoallowtheTEDEcriteriatobe usedwithTID-14844calculatedresults.Theguidanceofthissectionappliestoalldose calculationsperformedwithanASTpursuantto10CFR50.67.Certainselectiveimplementations maynotrequiredosecalculationsasdescribedinRegulatoryPosition1.3ofthisguide.4.1OffsiteDoseConsequencesThefollowingassumptionsshouldbeusedindeterminingtheTEDEforpersonslocatedatorbeyondtheboundaryoftheexclusionarea(EAB):4.1.1ThedosecalculationsshoulddeterminetheTEDE.TEDEisthesumofthecommittedeffectivedoseequivalent(CEDE)frominhalationandthedeepdoseequivalent(DDE)

fromexternalexposure.ThecalculationofthesetwocomponentsoftheTEDEshouldconsiderall radionuclides,includingprogenyfromthedecayofparentradionuclides,thataresignificantwith regardtodoseconsequencesandthereleasedradioactivity.134.1.2Theexposure-to-CEDEfactorsforinhalationofradioactivematerialshouldbederivedfromthedataprovidedinICRPPublication30,"LimitsforIntakesofRadionuclidesby Workers"(Ref.19).Table2.1ofFederalGuidanceReport11,"LimitingValuesofRadionuclide IntakeandAirConcentrationandDoseConversionFactorsforInhalation,Submersion,and Ingestion"(Ref.20),providestablesofconversionfactorsacceptabletotheNRCstaff.The factorsinthecolumnheaded"effective"yielddosescorrespondingtotheCEDE.4.1.3Forthefirst8hours,thebreathingrateofpersonsoffsiteshouldbeassumedtobe3.5x10-4cubicmeterspersecond.From8to24hoursfollowingtheaccident,thebreathingrateshouldbeassumedtobe1.8x10-4cubicmeterspersecond.Afterthatanduntiltheendoftheaccident,therateshouldbeassumedtobe2.3x10-4cubicmeterspersecon WithregardtotheEABTEDE,themaximumtwo-hourvalueisthebasisforscreeningandevaluationunder10CFR50.59.Changestodosesoutsideofthetwo-hourwindowareonlyconsideredinthecontextoftheirimpactonthemaximumtwo-hour EABTEDE.1.183-174.1.4TheDDEshouldbecalculatedassumingsubmergenceinsemi-infinitecloudassumptionswithappropriatecreditforattenuationbybodytissue.TheDDEisnominally equivalenttotheeffectivedoseequivalent(EDE)fromexternalexposureifthewholebodyis irradiateduniformly.Sincethisisareasonableassumptionforsubmergenceexposuresituations, EDEmaybeusedinlieuofDDEindeterminingthecontributionofexternaldosetotheTEDE.

TableIII.1ofFederalGuidanceReport12,"ExternalExposuretoRadionuclidesinAir,Water,andSoil"(Ref.21),providesexternalEDEconversionfactorsacceptabletotheNRCstaff.Thefactors inthecolumnheaded"effective"yielddosescorrespondingtotheEDE.4.1.5TheTEDEshouldbedeterminedforthemostlimitingpersonattheEAB.ThemaximumEABTEDEforanytwo-hourperiodfollowingthestartoftheradioactivityrelease shouldbedeterminedandusedindeterminingcompliancewiththedosecriteriain10CFR 50.67.14Themaximumtwo-hourTEDEshouldbedeterminedbycalculatingthepostulateddoseforaseriesofsmalltimeincrementsandperforminga"sliding"sumovertheincrementsfor successivetwo-hourperiods.ThemaximumTEDEobtainedissubmitted.Thetimeincrements shouldappropriatelyreflecttheprogressionoftheaccidenttocapturethepeakdoseinterval betweenthestartoftheeventandtheendofradioactivityrelease(seealsoTable6).4.1.6TEDEshouldbedeterminedforthemostlimitingreceptorattheouterboundaryofthelowpopulationzone(LPZ)andshouldbeusedindeterminingcompliancewiththedose criteriain10CFR50.67.4.1.7Nocorrectionshouldbemadefordepletionoftheeffluentplumebydepositionontheground.4.2ControlRoomDoseConsequencesThefollowingguidanceshouldbeusedindeterminingtheTEDEforpersonslocatedinthecontrolroom:4.2.1TheTEDEanalysisshouldconsiderallsourcesofradiationthatwillcauseexposuretocontrolroompersonnel.Theapplicablesourceswillvaryfromfacilitytofacility,buttypically willinclude:Contaminationofthecontrolroomatmospherebytheintakeorinfiltrationoftheradioactivematerialcontainedintheradioactiveplumereleasedfromthefacility,Contaminationofthecontrolroomatmospherebytheintakeorinfiltrationofairborneradioactivematerialfromareasandstructuresadjacenttothecontrolroom envelope,Radiationshinefromtheexternalradioactiveplumereleasedfromthefacility, 15Theiodineprotectionfactor(IPF)methodologyofReference22maynotbeadequatelyconservativeforallDBAsandcontrolroomarrangementssinceitmodelsasteady-statecontrolroomcondition.Sincemanyanalysisparameterschangeoverthedurationoftheevent,theIPFmethodologyshouldonlybeusedwithcaution.TheNRCcomputercodesHABIT(Ref.23)and RADTRAD(Ref.24)incorporatesuitablemethodologies.16Thisoccupancyismodeledinthe c/QvaluesdeterminedinReference22andshouldnotbecreditedtwice.TheARCON96Code(Ref.26)doesnotincorporatetheseoccupancyassumptions,makingitnecessarytoapplythiscorrectioninthedosecalculations.1.183-18Radiationshinefromradioactivematerialinthereactorcontainment,Radiationshinefromradioactivematerialinsystemsandcomponentsinsideorexternaltothecontrolroomenvelope,e.g.,radioactivematerialbuildupin recirculationfilters.4.2.2Theradioactivematerialreleasesandradiationlevelsusedinthecontrolroomdoseanalysisshouldbedeterminedusingthesamesourceterm,transport,andreleaseassumptionsused fordeterminingtheEABandtheLPZTEDEvalues,unlesstheseassumptionswouldresultinnon- conservativeresultsforthecontrolroom.4.2.3Themodelsusedtotransportradioactivematerialintoandthroughthecontrolroom,15andtheshieldingmodelsusedtodetermineradiationdoseratesfromexternalsources,shouldbestructuredtoprovidesuitablyconservativeestimatesoftheexposuretocontrolroom personnel.4.2.4Creditforengineeredsafetyfeaturesthatmitigateairborneradioactivematerialwithinthecontrolroommaybeassumed.Suchfeaturesmayincludecontrolroomisolationor pressurization,orintakeorrecirculationfiltration.RefertoSection6.5.1,"ESFAtmospheric CleanupSystem,"oftheSRP(Ref.3)andRegulatoryGuide1.52,"Design,Testing,and MaintenanceCriteriaforPostaccidentEngineered-Safety-FeatureAtmosphereCleanupSystemAir FiltrationandAdsorptionUnitsofLight-Water-CooledNuclearPowerPlants"(Ref.25),for guidance.ThecontrolroomdesignisoftenoptimizedfortheDBALOCAandtheprotection affordedforotheraccidentsequencesmaynotbeasadvantageous.Inmostdesigns,controlroom isolationisactuatedbyengineeredsafeguardsfeature(ESF)signalsorradiationmonitors(RMs).

Insomecases,theESFsignaliseffectiveonlyforselectedaccidents,placingrelianceontheRMs fortheremainingaccidents.SeveralaspectsofRMscandelaythecontrolroomisolation, includingthedelayforactivitytobuilduptoconcentrationsequivalenttothealarmsetpointand theeffectsofdifferentradionuclideaccidentisotopicmixesonmonitorresponse.4.2.5Creditshouldgenerallynotbetakenfortheuseofpersonalprotectiveequipmentorprophylacticdrugs.Deviationsmaybeconsideredonacase-by-casebasis.4.2.6Thedosereceptorfortheseanalysesisthehypotheticalmaximumexposedindividualwhoispresentinthecontrolroomfor100%ofthetimeduringthefirst24hoursafter theevent,60%ofthetimebetween1and4days,and40%ofthetimefrom4daysto30days.16Forthedurationoftheevent,thebreathingrateofthisindividualshouldbeassumedtobe3.5x10-4cubicmeterspersecon .183-194.2.7ControlroomdosesshouldbecalculatedusingdoseconversionfactorsidentifiedinRegulatoryPosition4.1aboveforuseinoffsitedoseanalyses.TheDDEfromphotonsmaybe correctedforthedifferencebetweenfinitecloudgeometryinthecontrolroomandthesemi- infinitecloudassumptionusedincalculatingthedoseconversionfactors.Thefollowing expressionmaybeusedtocorrectthesemi-infiniteclouddose,DDE,toafiniteclouddose,DDEfinite,wherethecontrolroomismodeledasahemispherethathasavolume,V,incubicfeet,equivalenttothatofthecontrolroom(Ref.22).Equation1DDEDDEVfinite=¥03381173.4.3OtherDoseConsequencesTheguidanceprovidedinRegulatoryPositions4.1and4.2shouldbeused,asapplicable,inre-assessingtheradiologicalanalysesidentifiedinRegulatoryPosition1.3.1,suchasthosein NUREG-0737(Ref.2).DesignenvelopesourcetermsprovidedinNUREG-0737shouldbe updatedforconsistencywiththeAST.Ingeneral,radiationexposurestoplantpersonnelidentified inRegulatoryPosition1.3.1shouldbeexpressedintermsofTEDE.Integratedradiationexposure ofplantequipmentshouldbedeterminedusingtheguidanceofAppendixIofthisguide.4.4AcceptanceCriteriaTheradiologicalcriteriafortheEAB,theouterboundaryoftheLPZ,andforthecontrolroomarein10CFR50.67.Thesecriteriaarestatedforevaluatingreactoraccidentsofexceedingly lowprobabilityofoccurrenceandlowriskofpublicexposuretoradiation,e.g.,alarge-break LOCA.Thecontrolroomcriterionappliestoallaccidents.Foreventswithahigherprobabilityof occurrence,postulatedEABandLPZdosesshouldnotexceedthecriteriatabulatedinTable6.TheacceptancecriteriaforthevariousNUREG-0737(Ref.2)itemsgenerallyreferenceGeneralDesignCriteria19(GDC19)fromAppendixAto10CFRPart50orspecifycriteria derivedfromGDC-19.Thesecriteriaaregenerallyspecifiedintermsofwholebodydose,orits equivalenttoanybodyorgan.Forfacilitiesapplyingfor,orhavingreceived,approvalfortheuse ofanAST,theapplicablecriteriashouldbeupdatedforconsistencywiththeTEDEcriterionin10CFR50.67(b)(2)(iii).

17ForPWRswithsteamgeneratoralternativerepaircriteria,differentdosecriteriamayapplytosteamgeneratortuberuptureandmainsteamlinebreakanalyses.1.183-20Table617AccidentDoseCriteriaAccidentorCaseEABandLPZDoseCriteriaAnalysisReleaseDurationLOCA25remTEDE30daysforcontainment,ECCS,andMSIV(BWR)leakageBWRMainSteamLineBreakInstantaneouspuffFuelDamageorPre-incidentSpike25remTEDEEquilibriumIodineActivity2.5remTEDEBWRRodDropAccident6.3remTEDE24hours PWRSteamGeneratorTubeRuptureAffectedSG:timetoisolate;UnaffectedSG(s):untilcoldshutdownisestablishedFuelDamageorPre-incidentSpike25remTEDECoincidentIodineSpike2.5remTEDEPWRMainSteamLineBreakUntilcoldshutdownisestablishedFuelDamageorPre-incidentSpike25remTEDE CoincidentIodineSpike2.5remTEDEPWRLockedRotorAccident2.5remTEDEUntilcoldshutdownisestablished PWRRodEjectionAccident6.3remTEDE30daysforcontainmentpathway;untilcoldshutdownisestablishedfor secondarypathwayFuelHandlingAccident6.3remTEDE2hoursThecolumnlabeled"AnalysisReleaseDuration"isasummaryoftheassumedradioactivityreleasedurationsidentifiedintheindividualappendicestothisguide.Refertothese appendicesforcompletedescriptionsofthereleasepathwaysanddurations.5.ANALYSISASSUMPTIONSAND

METHODOLOGY

5.1GeneralConsiderations5.1.1AnalysisQualityTheevaluationsrequiredby10CFR50.67arere-analysesofthedesignbasissafetyanalysesandevaluationsrequiredby10CFR50.34;theyareconsideredtobeasignificantinputto theevaluationsrequiredby10CFR50.92or10CFR50.59.Theseanalysesshouldbeprepared, reviewed,andmaintainedinaccordancewithqualityassuranceprogramsthatcomplywith AppendixB,"QualityAssuranceCriteriaforNuclearPowerPlantsandFuelReprocessingPlants,"

to10CFRPart50.Thesedesignbasisanalyseswerestructuredtoprovideaconservativesetofassumptionstotesttheperformanceofoneormoreaspectsofthefacilitydesign.Manyphysicalprocessesand phenomenaarerepresentedbyconservative,boundingassumptionsratherthanbeingmodeled 18Notethatforsomeparameters,thetechnicalspecificationvaluemaybeadjustedforanalysispurposesbyfactorsprovidedinotherregulatoryguidance.Forexample,ESFfilterefficienciesarebasedontheguidanceinRegulatoryGuide1.52(Ref.25)and inGenericLetter99-02(Ref.27)ratherthanthesurveillancetestcriteriainthetechnicalspecifications.Generally,these adjustmentsaddresspotentialchangesintheparameterbetweenscheduledsurveillancetests.1.183-21directly.Thestaffhasselectedassumptionsandmodelsthatprovideanappropriateandprudentsafetymarginagainstunpredictedeventsinthecourseofanaccidentandcompensateforlarge uncertaintiesinfacilityparameters,accidentprogression,radioactivematerialtransport,and atmosphericdispersion.Licenseesshouldexercisecautioninproposingdeviationsbasedupon datafromaspecificaccidentsequencesincetheDBAswereneverintendedtorepresentany specificaccidentsequence--theproposeddeviationmaynotbeconservativeforotheraccident sequences.5.1.2CreditforEngineeredSafeguardFeaturesCreditmaybetakenforaccidentmitigationfeaturesthatareclassifiedassafety-related,arerequiredtobeoperablebytechnicalspecifications,arepoweredbyemergencypowersources,and areeitherautomaticallyactuatedor,inlimitedcases,haveactuationrequirementsexplicitly addressedinemergencyoperatingprocedures.Thesingleactivecomponentfailurethatresultsin themostlimitingradiologicalconsequencesshouldbeassumed.Assumptionsregardingthe occurrenceandtimingofalossofoffsitepowershouldbeselectedwiththeobjectiveof maximizingthepostulatedradiologicalconsequences.5.1.3AssignmentofNumericInputValuesThenumericvaluesthatarechosenasinputstotheanalysesrequiredby10CFR50.67shouldbeselectedwiththeobjectiveofdeterminingaconservativepostulateddose.Insome instances,aparticularparametermaybeconservativeinoneportionofananalysisbutbe nonconservativeinanotherportionofthesameanalysis.Forexample,assumingminimum containmentsystemsprayflowisusuallyconservativeforestimatingiodinescrubbing,butin manycasesmaybenonconservativewhendeterminingsumppH.Sensitivityanalysesmaybe neededtodeterminetheappropriatevaluetouse.Asaconservativealternative,thelimitingvalue applicabletoeachportionoftheanalysismaybeusedintheevaluationofthatportion.Asingle valuemaynotbeapplicableforaparameterforthedurationoftheevent,particularlyfor parametersaffectedbychangesindensity.Forparametersaddressedbytechnicalspecifications, thevalueusedintheanalysisshouldbethatspecifiedinthetechnicalspecifications.18Ifarangeofvaluesoratolerancebandisspecified,thevaluethatwouldresultinaconservativepostulated doseshouldbeused.Iftheparameterisbasedontheresultsoflessfrequentsurveillancetesting, e.g.,steamgeneratornondestructivetesting(NDT),considerationshouldbegiventothe degradationthatmayoccurbetweenperiodictestsinestablishingtheanalysisvalue.5.1.4ApplicabilityofPriorLicensingBasisTheNRCstaffconsiderstheimplementationofanASTtobeasignificantchangetothedesignbasisofthefacilitythatisvoluntarilyinitiatedbythelicensee.Inordertoissuealicense amendmentauthorizingtheuseofanASTandtheTEDEdosecriteria,theNRCstaffmustmakea currentfindingofcompliancewithregulationsapplicabletotheamendment.Thecharacteristics oftheASTsandthereviseddosecalculationalmethodologymaybeincompatiblewithmanyofthe analysisassumptionsandmethodscurrentlyreflectedinthefacility'sdesignbasisanalyses.The NRCstaffmayfindthatneworunreviewedissuesarecreatedbyaparticularsite-specific 1.183-22implementationoftheAST,warrantingreviewofstaffpositionsapprovedsubsequenttotheinitialissuanceofthelicense.Thisisnotconsideredabackfitasdefinedby10CFR50.109,

"Backfitting."However,priordesignbasesthatareunrelatedtotheuseoftheAST,orare unaffectedbytheAST,maycontinueasthefacility'sdesignbasis.Licenseesshouldensurethat analysisassumptionsandmethodsarecompatiblewiththeASTsandtheTEDEcriteria.5.2Accident-SpecificAssumptionsTheappendicestothisregulatoryguideprovideaccident-specificassumptionsthatareacceptabletothestaffforperforminganalysesthatarerequiredby10CFR50.67.TheDBAs addressedintheseattachmentswereselectedfromaccidentsthatmayinvolvedamagetoirradiated fuel.ThisguidedoesnotaddressDBAswithradiologicalconsequencesbasedontechnical specificationreactororsecondarycoolant-specificactivitiesonly.Theinclusionorexclusionofa particularDBAinthisguideshouldnotbeinterpretedasindicatingthatananalysisofthatDBAis requiredornotrequired.LicenseesshouldanalyzetheDBAsthatareaffectedbythespecific proposedapplicationsofanAST.TheNRCstaffhasdeterminedthattheanalysisassumptionsintheappendicestothisguideprovideanintegratedapproachtoperformingtheindividualanalysesandgenerallyexpects licenseestoaddresseachassumptionorproposeacceptablealternatives.Suchalternativesmaybe justifiableonthebasisofplant-specificconsiderations,updatedtechnicalanalyses,or,insome cases,apreviouslyapprovedlicensingbasisconsideration.Theassumptionsintheappendicesare deemedconsistentwiththeASTidentifiedinRegulatoryPosition3andinternallyconsistentwith eachother.Althoughlicenseesarefreetoproposealternativestotheseassumptionsfor considerationbytheNRCstaff,licenseesshouldavoiduseofpreviouslyapprovedstaffpositions thatwouldadverselyaffectthisconsistency.TheNRCiscommittedtousingprobabilisticriskanalysis(PRA)insightsinitsregulatoryactivitiesandwillconsiderlicenseeproposalsforchangesinanalysisassumptionsbaseduponrisk insights.Thestaffwillnotapproveproposalsthatwouldreducethedefenseindepthdeemed necessarytoprovideadequateprotectionforpublichealthandsafety.Insomecases,thisdefense indepthcompensatesforuncertaintiesinthePRAanalysesandaddressesaccidentconsiderations notadequatelyaddressedbythecoredamagefrequency(CDF)andlargeearlyreleasefrequency (LERF)surrogateindicatorsofoverallrisk.5.3MeteorologyAssumptionsAtmosphericdispersionvalues(c/Q)fortheEAB,theLPZ,andthecontrolroomthatwereapprovedbythestaffduringinitialfacilitylicensingorinsubsequentlicensingproceedingsmaybeusedinperformingtheradiologicalanalysesidentifiedbythisguide.Methodologiesthathavebeenusedfordeterminingc/QvaluesaredocumentedinRegulatoryGuides1.3and1.4,RegulatoryGuide1.145,"AtmosphericDispersionModelsforPotentialAccidentConsequenceAssessmentsatNuclearPowerPlants,"andthepaper,"NuclearPowerPlantControlRoom VentilationSystemDesignforMeetingGeneralCriterion19"(Refs.6,7,22,and28).

19TheARCON96computercodecontainsprocessingoptionsthatmayyieldc/Qvaluesthatarenotsufficientlyconservativeforuseinaccidentconsequenceassessmentsormaybeincompatiblewithreleasepointandventilationintakeconfigurationsatparticularsites.Theapplicabilityoftheseoptionsandassociatedinputparametersshouldbeevaluatedonacase-by-casebasis.

TheassumptionsmadeintheexamplesintheARCON96documentationareillustrativeonlyanddonotimplyNRCstaff acceptanceofthemethodsordatausedintheexample.1.183-23References22and28shouldbeusediftheFSARc/Qvaluesaretoberevisedorifvaluesaretobedeterminedfornewreleasepointsorreceptordistances.FumigationshouldbeconsideredwhereapplicablefortheEABandLPZ.FortheEAB,theassumedfumigationperiod shouldbetimedtobeincludedintheworst2-hourexposureperiod.TheNRCcomputercode PAVAN(Ref.29)implementsRegulatoryGuide1.145(Ref.28)anditsuseisacceptabletothe NRCstaff.ThemethodologyoftheNRCcomputercodeARCON9619(Ref.26)isgenerallyacceptabletotheNRCstaffforuseindeterminingcontrolroomc/Qvalues.Meteorologicaldatacollectedinaccordancewiththesite-specificmeteorologicalmeasurementsprogramdescribedinthefacilityFSARshouldbeusedingeneratingaccidentc/Qvalues.AdditionalguidanceisprovidedinRegulatoryGuide1.23,"OnsiteMeteorologicalPrograms"(Ref.30).Allchangesinÿ/QanalysismethodologyshouldbereviewedbytheNRCstaff.6.ASSUMPTIONSFOREVALUATINGTHERADIATIONDOSESFOREQUIPMENTQUALIFICATIONTheassumptionsinAppendixItothisguideareacceptabletotheNRCstaffforperformingradiologicalassessmentsassociatedwithequipmentqualification.TheassumptionsinAppendixI willsupersedeRegulatoryPositions2.c(1)and2.c(2)andAppendixDofRevision1ofRegulatory Guide1.89,"EnvironmentalQualificationofCertainElectricEquipmentImportanttoSafetyfor NuclearPowerPlants"(Ref.11),foroperatingreactorsthathaveamendedtheirlicensingbasisto useanalternativesourceterm.ExceptasstatedinAppendixI,allotherassumptions,methods, andprovisionsofRevision1ofRegulatoryGuide1.89remaineffective.TheNRCstaffisassessingtheeffectofincreasedcesiumreleasesonEQdosestodeterminewhetherlicenseeactioniswarranted.Untilsuchtimeasthisgenericissueisresolved, licenseesmayuseeithertheASTortheTID14844assumptionsforperformingtherequiredEQ analyses.However,noplantmodificationsarerequiredtoaddresstheimpactofthedifferencein sourcetermcharacteristics(i.e.,ASTvsTID14844)onEQdosespendingtheoutcomeofthe evaluationofthegenericissue.

D. IMPLEMENTATION

ThepurposeofthissectionistoprovideinformationtoapplicantsandlicenseesregardingtheNRCstaff'splansforusingthisregulatoryguide.ExceptinthosecasesinwhichanapplicantorlicenseeproposesanacceptablealternativemethodforcomplyingwiththespecifiedportionsoftheNRC'sregulations,themethodsdescribed inthisguidewillbeusedintheevaluationofsubmittalsrelatedtotheuseofASTsinradiological consequenceanalysesatoperatingpowerreactor .183-24 1.183-25REFERENCES{SeetheinsidefrontcoverofthisguideforinformationonobtainingNRCdocuments.}1.J.J.DiNunnoetal.,"CalculationofDistanceFactorsforPowerandTestReactorSites,"USAECTID-14844,U.S.AtomicEnergyCommission(nowUSNRC),1962.2.USNRC,"ClarificationofTMIActionPlanRequirements,"NUREG-0737,November1980.3.USNRC,"StandardReviewPlanfortheReviewofSafetyAnalysisReportsforNuclearPowerPlants,"NUREG-0800,September1981(orupdatesofspecificsections).4.USNRC,"UseofProbabilisticRiskAssessmentMethodsinNuclearActivities:FinalPolicyStatement,"FederalRegister,Volume60,page42622(60FR42622)August16,1995.5.L.Sofferetal.,"AccidentSourceTermsforLight-WaterNuclearPowerPlants,"NUREG-1465,USNRC,February1995.6.USNRC,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaLossofCoolantAccidentforBoilingWaterReactors."RegulatoryGuide1.3,Revision2, June1974.7.USNRC,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaLossofCoolantAccidentforPressurizedWaterReactors,"RegulatoryGuide1.4,Revision 2,June1974.8.USNRC,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaSteamLineBreakAccidentforBoilingWaterReactors,"RegulatoryGuide1.5,March 1971.9.USNRC,"AssumptionsUsedforEvaluatingthePotentialRadiologicalConsequencesofaFuelHandlingAccidentintheFuelHandlingandStorageFacilityforBoilingand PressurizedWaterReactors,"RegulatoryGuide1.25,March1972.10.USNRC,"AssumptionsUsedforEvaluatingaControlRodEjectionAccidentforPressurizedWaterReactors,"RegulatoryGuide1.77,May1974.11.USNRC,"EnvironmentalQualificationofCertainElectricEquipmentImportanttoSafetyforNuclearPowerPlants,"RegulatoryGuide1.89,Revision1,June1984.12.USNRC,"PlanningBasisfortheDevelopmentofStateandLocalGovernmentRadiologicalEmergencyResponsePlansinSupportofLightWaterNuclearPowerPlants,"

NUREG-0396,December197 .183-2613.USNRC,"CriteriaforPreparationandEvaluationofRadiologicalEmergencyResponsePlansandPreparednessinSupportofNuclearPowerPlants,"NUREG-0654,Revision1 (FEMA-REP-1),November1980.14.USNRC,"ResultsoftheRevised(NUREG-1465)SourceTermRebaseliningforOperatingReactors,"SECY-98-154,June30,1998.15.USNRC,"AnApproachforUsingProbabilisticRiskAssessmentinRisk-InformedDecisionsonPlant-SpecificChangestotheLicensingBasis,"RegulatoryGuide1.174,July 1998.16.USNRC,"StandardFormatandContentofSafetyAnalysisReportsforNuclearPowerPlants(LWREdition),"RegulatoryGuide1.70,Revision3,November1978.17.A.G.Croff,"AUser'sManualfortheORIGEN2ComputerCode,"ORNL/TM-7175,OakRidgeNationalLaboratory,July1980.18.S.M.BowmanandL.C.Leal,"TheORIGNARPInputProcessorforORIGEN-ARP,"AppendixF7.AinSCALE:AModularCodeSystemforPerformingStandardizedAnalysesforLicensingEvaluation,NUREG/CR-0200,USNRC,March1997.19.ICRP,"LimitsforIntakesofRadionuclidesbyWorkers,"ICRPPublication30,1979.

20.K.F.Eckermanetal.,"LimitingValuesofRadionuclideIntakeandAirConcentrationandDoseConversionFactorsforInhalation,Submersion,andIngestion,"FederalGuidance Report11,EPA-520/1-88-020,EnvironmentalProtectionAgency,1988.21.K.F.EckermanandJ.C.Ryman,"ExternalExposuretoRadionuclidesinAir,Water,andSoil,"FederalGuidanceReport12,EPA-402-R-93-081,EnvironmentalProtectionAgency, 1993.22.K.G.MurphyandK.W.Campe,"NuclearPowerPlantControlRoomVentilationSystemDesignforMeetingGeneralCriterion19,"publishedinProceedingsof13thAECAirCleaningConference,AtomicEnergyCommission(nowUSNRC),August1974.23.USNRC,"ComputerCodesforEvaluationofControlRoomHabitability(HABITV1.1),"Supplement1toNUREG/CR-6210,November1998.24.S.L.Humphreysetal.,"RADTRAD:ASimplifiedModelforRadionuclideTransportandRemovalandDoseEstimation,"NUREG/CR-6604,USNRC,April1998.25.USNRC,"Design,Testing,andMaintenanceCriteriaforPostaccidentEngineeredSafetyFeatureAtmosphereCleanupSystemAirFiltrationandAdsorptionUnitsofLight-Water- CooledNuclearPowerPlants,"RegulatoryGuide1.52,Revision2,March197 .183-2726.J.V.RamsdellandC.A.Simonen,"AtmosphericRelativeConcentrationsinBuildingWakes,NUREG-6331,Revision1,USNRC,May1997.27.USNRC,"LaboratoryTestingofNuclear-GradeActivatedCharcoal,"NRCGenericLetter99-02,June3,1999.28.USNRC,"AtmosphericDispersionModelsforPotentialAccidentConsequenceAssessmentsatNuclearPowerPlants,"RegulatoryGuide1.145,Revision1,November 1982.29.T.J.Bander,"PAVAN:AnAtmosphericDispersionProgramforEvaluatingDesignBasisAccidentalReleasesofRadioactiveMaterialsfromNuclearPowerStations,"NUREG-

2858,USNRC,November1982.30.USNRC,"OnsiteMeteorologicalPrograms,"RegulatoryGuide1.23,February197 A-1AppendixAASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFALWRLOSS-OF-COOLANTACCIDENTTheassumptionsinthisappendixareacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofloss-of-coolantaccidents(LOCAs)atlightwaterreactors(LWRs).

Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.AppendixA,"GeneralDesignCriteriaforNuclearPowerPlants,"to10CFRPart50definesLOCAsasthosepostulatedaccidentsthatresultfromalossofcoolantinventoryatrates thatexceedthecapabilityofthereactorcoolantmakeupsystem.Leaksuptoadouble-ended ruptureofthelargestpipeofthereactorcoolantsystemareincluded.TheLOCA,aswithall designbasisaccidents(DBAs),isaconservativesurrogateaccidentthatisintendedtochallenge selectiveaspectsofthefacilitydesign.Analysesareperformedusingaspectrumofbreaksizesto evaluatefuelandECCSperformance.Withregardtoradiologicalconsequences,alarge-break LOCAisassumedasthedesignbasiscaseforevaluatingtheperformanceofreleasemitigation systemsandthecontainmentandforevaluatingtheproposedsitingofafacility.SOURCETERMASSUMPTIONS1.AcceptableassumptionsregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisguide.2.IfthesumporsuppressionpoolpHiscontrolledatvaluesof7orgreater,thechemicalformofradioiodinereleasedtothecontainmentshouldbeassumedtobe95%cesiumiodide(CsI),

4.85percentelementaliodine,and0.15percentorganiciodide.Iodinespecies,includingthose fromiodinere-evolution,forsumporsuppressionpoolpHvalueslessthan7willbeevaluatedon acase-by-casebasis.EvaluationsofpHshouldconsidertheeffectofacidsandbasescreated duringtheLOCAevent,e.g.,radiolysisproducts.Withtheexceptionofelementalandorganic iodineandnoblegases,fissionproductsshouldbeassumedtobeinparticulateform.ASSUMPTIONSONTRANSPORTINPRIMARYCONTAINMENT3.Acceptableassumptionsrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromtheprimarycontainmentinPWRsorthedrywellinBWRsareasfollows:3.1TheradioactivityreleasedfromthefuelshouldbeassumedtomixinstantaneouslyandhomogeneouslythroughoutthefreeairvolumeoftheprimarycontainmentinPWRsorthe drywellinBWRsasitisreleased.Thisdistributionshouldbeadjustedifthereareinternal compartmentsthathavelimitedventilationexchange.Thesuppressionpoolfreeair volumemaybeincludedprovidedthereisamechanismtoensuremixingbetweenthe drywelltothewetwell.Thereleaseintothecontainmentordrywellshouldbeassumedto terminateattheendoftheearlyin-vesselphase.3.2Reductioninairborneradioactivityinthecontainmentbynaturaldepositionwithinthecontainmentmaybecredited.Acceptablemodelsforremovalofiodineandaerosolsare 1Thisdocumentdescribesstatisticalformulationswithdifferinglevelsofuncertainty.Theremovalrateconstantsselectedforuseindesignbasiscalculationsshouldbethosethatwillmaximizethedoseconsequences.ForBWRs,thesimplifiedmodel shouldbeusedonlyifthereleasefromthecoreisnotdirectedthroughthesuppressionpool.Iodineremovalinthesuppression poolaffectstheiodinespeciesassumedbythemodeltobepresentinitially.A-2describedinChapter6.5.2,"ContainmentSprayasaFissionProductCleanupSystem,"oftheStandardReviewPlan(SRP),NUREG-0800(Ref.A-1)andinNUREG/CR-6189,"A SimplifiedModelofAerosolRemovalbyNaturalProcessesinReactorContainments" (Ref.A-2).ThelattermodelisincorporatedintotheanalysiscodeRADTRAD(Ref.A-3).

Thepriorpracticeofdeterministicallyassumingthata50%plateoutofiodineisreleased fromthefuelisnolongeracceptabletotheNRCstaffasitisinconsistentwiththe characteristicsoftherevisedsourceterms.3.3ReductioninairborneradioactivityinthecontainmentbycontainmentspraysystemsthathavebeendesignedandaremaintainedinaccordancewithChapter6.5.2oftheSRP(Ref.

A-1)maybecredited.Acceptablemodelsfortheremovalofiodineandaerosolsare describedinChapter6.5.2oftheSRPandNUREG/CR-5966,"ASimplifiedModelof AerosolRemovalbyContainmentSprays"1(Ref.A-4).ThissimplifiedmodelisincorporatedintotheanalysiscodeRADTRAD(Refs.A-1toA-3).Theevaluationofthecontainmentspraysshouldaddressareaswithintheprimarycontainmentthatarenotcoveredbythespraydrops.Themixingrateattributedtonatural convectionbetweensprayedandunsprayedregionsofthecontainmentbuilding,provided thatadequateflowexistsbetweentheseregions,isassumedtobetwoturnoversofthe unsprayedregionsperhour,unlessotherratesarejustified.Thecontainmentbuilding atmospheremaybeconsideredasingle,well-mixedvolumeifthespraycoversatleast90%

ofthevolumeandifadequatemixingofunsprayedcompartmentscanbeshown.TheSRPsetsforthamaximumdecontaminationfactor(DF)forelementaliodinebasedonthemaximumiodineactivityintheprimarycontainmentatmospherewhenthesprays actuate,dividedbytheactivityofiodineremainingatsometimeafterdecontamination.

TheSRPalsostatesthattheparticulateiodineremovalrateshouldbereducedbyafactor of10whenaDFof50isreached.Thereductionintheremovalrateisnotrequiredifthe removalrateisbasedonthecalculatedtime-dependentairborneaerosolmass.Thereisno specifiedmaximumDFforaerosolremovalbysprays.Themaximumactivitytobeusedin determiningtheDFisdefinedastheiodineactivityinthecolumnslabeled"Total"in Tables1and2ofthisguidemultipliedby0.05forelementaliodineandby0.95for particulateiodine(i.e.,aerosoltreatedasparticulateinSRPmethodology).3.4Reductioninairborneradioactivityinthecontainmentbyin-containmentrecirculationfiltersystemsmaybecreditedifthesesystemsmeettheguidanceofRegulatoryGuide1.52and GenericLetter99-02(Refs.A-5andA-6).Thefiltermedialoadingcausedbythe increasedaerosolreleaseassociatedwiththerevisedsourcetermshouldbeaddressed.3.5ReductioninairborneradioactivityinthecontainmentbysuppressionpoolscrubbinginBWRsshouldgenerallynotbecredited.However,thestaffmayconsidersuchreductionon anindividualcasebasis.Theevaluationshouldconsidertherelativetimingoftheblowdown andthefissionproductreleasefromthefuel,theforcedrivingthereleasethroughthepool, A-3andthepotentialforanybypassofthesuppressionpool(Ref.7).Analysesshouldconsideriodinere-evolutionifthesuppressionpoolliquidpHisnotmaintainedgreaterthan7.3.6Reductioninairborneradioactivityinthecontainmentbyretentioninicecondensers,orotherengineeringsafetyfeaturesnotaddressedabove,shouldbeevaluatedonanindividualcase basis.SeeSection6.5.4oftheSRP(Ref.A-1).3.7Theprimarycontainment(i.e.,drywellforMarkIandIIcontainmentdesigns)shouldbeassumedtoleakatthepeakpressuretechnicalspecificationleakrateforthefirst24hours.

ForPWRs,theleakratemaybereducedafterthefirst24hoursto50%ofthetechnical specificationleakrate.ForBWRs,leakagemaybereducedafterthefirst24hours,if supportedbyplantconfigurationandanalyses,toavaluenotlessthan50%ofthetechnical specificationleakrate.Leakagefromsubatmosphericcontainmentsisassumedtoterminate whenthecontainmentisbroughttoandmaintainedatasubatmosphericconditionasdefined bytechnicalspecifications.ForBWRswithMarkIIIcontainments,theleakagefromthedrywellintotheprimarycontainmentshouldbebasedonthesteamingrateoftheheatedreactorcore,withnocredit forcoredebrisrelocation.Thisleakageshouldbeassumedduringthetwo-hourperiod betweentheinitialblowdownandterminationofthefuelradioactivityrelease(gapandearly in-vesselreleasephases).Aftertwohours,theradioactivityisassumedtobeuniformly distributedthroughoutthedrywellandtheprimarycontainment.3.8Iftheprimarycontainmentisroutinelypurgedduringpoweroperations,releasesviathepurgesystempriortocontainmentisolationshouldbeanalyzedandtheresultingdoses summedwiththepostulateddosesfromotherreleasepaths.Thepurgereleaseevaluation shouldassumethat100%oftheradionuclideinventoryinthereactorcoolantsystemliquidis releasedtothecontainmentattheinitiationoftheLOCA.Thisinventoryshouldbebasedon thetechnicalspecificationreactorcoolantsystemequilibriumactivity.Iodinespikesneednot beconsidered.Ifthepurgesystemisnotisolatedbeforetheonsetofthegapreleasephase, thereleasefractionsassociatedwiththegapreleaseandearlyin-vesselphasesshouldbe consideredasapplicable.ASSUMPTIONSONDUALCONTAINMENTS4.Forfacilitieswithdualcontainmentsystems,theacceptableassumptionsrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromthesecondarycontainmentor enclosurebuildingsareasfollows.4.1Leakagefromtheprimarycontainmentshouldbeconsideredtobecollected,processedbyengineeredsafetyfeature(ESF)filters,ifany,andreleasedtotheenvironmentviathe secondarycontainmentexhaustsystemduringperiodsinwhichthesecondarycontainment hasanegativepressureasdefinedintechnicalspecifications.Creditforanelevatedrelease shouldbeassumedonlyifthepointofphysicalreleaseismorethantwoandone-halftimes theheightofanyadjacentstructur A-44.2Leakagefromtheprimarycontainmentisassumedtobereleaseddirectlytotheenvironmentasaground-levelreleaseduringanyperiodinwhichthesecondarycontainmentdoesnot haveanegativepressureasdefinedintechnicalspecifications.4.3Theeffectofhighwindspeedsontheabilityofthesecondarycontainmenttomaintainanegativepressureshouldbeevaluatedonanindividualcasebasis.Thewindspeedtobe assumedisthe1-houraveragevaluethatisexceededonly5%ofthetotalnumberofhoursin thedataset.Ambienttemperaturesusedintheseassessmentsshouldbethe1-houraverage valuethatisexceededonly5%or95%ofthetotalnumbersofhoursinthedataset, whicheverisconservativefortheintendeduse(e.g.,ifhightemperaturesarelimiting,use thoseexceededonly5%).4.4Creditfordilutioninthesecondarycontainmentmaybeallowedwhenadequatemeanstocausemixingcanbedemonstrated.Otherwise,theleakagefromtheprimarycontainment shouldbeassumedtobetransporteddirectlytoexhaustsystemswithoutmixing.Creditfor mixing,iffoundtobeappropriate,shouldgenerallybelimitedto50%.Thisevaluation shouldconsiderthemagnitudeofthecontainmentleakageinrelationtocontiguousbuilding volumeorexhaustrate,thelocationofexhaustplenumsrelativetoprojectedrelease locations,therecirculationventilationsystems,andinternalwallsandfloorsthatimpede streamflowbetweenthereleaseandtheexhaust.4.5Primarycontainmentleakagethatbypassesthesecondarycontainmentshouldbeevaluatedatthebypassleakrateincorporatedinthetechnicalspecifications.Ifthebypassleakageis throughwater,e.g.,viaafilledpipingrunthatismaintainedfull,creditforretentionofiodine andaerosolsmaybeconsideredonacase-by-casebasis.Similarly,depositionofaerosol radioactivityingas-filledlinesmaybeconsideredonacase-by-casebasis.4.6ReductionintheamountofradioactivematerialreleasedfromthesecondarycontainmentbecauseofESFfiltersystemsmaybetakenintoaccountprovidedthatthesesystemsmeetthe guidanceofRegulatoryGuide1.52(Ref.A-5)andGenericLetter99-02(Ref.A-6).ASSUMPTIONSONESFSYSTEMLEAKAGE5.ESFsystemsthatrecirculatesumpwateroutsideoftheprimarycontainmentareassumedtoleakduringtheirintendedoperation.Thisreleasesourceincludesleakagethroughvalvepacking glands,pumpshaftseals,flangedconnections,andothersimilarcomponents.Thisreleasesource mayalsoincludeleakagethroughvalvesisolatinginterfacingsystems(Ref.A-7).Theradiological consequencesfromthepostulatedleakageshouldbeanalyzedandcombinedwithconsequences postulatedforotherfissionproductreleasepathstodeterminethetotalcalculatedradiological consequencesfromtheLOCA.Thefollowingassumptionsareacceptableforevaluatingthe consequencesofleakagefromESFcomponentsoutsidetheprimarycontainmentforBWRsand PWRs.5.1Withtheexceptionofnoblegases,allthefissionproductsreleasedfromthefueltothecontainment(asdefinedinTables1and2ofthisguide)shouldbeassumedto instantaneouslyandhomogeneouslymixintheprimarycontainmentsumpwater(inPWRs)

orsuppressionpool(inBWRs)atthetimeofreleasefromthecore.Inlieuofthis A-5deterministicapproach,suitablyconservativemechanisticmodelsforthetransportofairborneactivityincontainmenttothesumpwatermaybeused.Notethatmanyofthe parametersthatmakesprayanddepositionmodelsconservativewithregardtocontainment airborneleakagearenonconservativewithregardtothebuildupofsumpactivity.5.2TheleakageshouldbetakenastwotimesthesumofthesimultaneousleakagefromallcomponentsintheESFrecirculationsystemsabovewhichthetechnicalspecifications,or licenseecommitmentstoitemIII.D.1.1ofNUREG-0737(Ref.A-8),wouldrequiredeclaringsuchsystemsinoperable.Theleakageshouldbeassumedtostartattheearliesttimethe recirculationflowoccursinthesesystemsandendatthelatesttimethereleasesfromthese systemsareterminated.Considerationshouldalsobegiventodesignleakagethroughvalves isolatingESFrecirculationsystemsfromtanksventedtoatmosphere,e.g.,emergencycore coolingsystem(ECCS)pumpminiflowreturntotherefuelingwaterstoragetank.5.3Withtheexceptionofiodine,allradioactivematerialsintherecirculatingliquidshouldbeassumedtoberetainedintheliquidphase.5.4Ifthetemperatureoftheleakageexceeds212°F,thefractionoftotaliodineintheliquidthatbecomesairborneshouldbeassumedequaltothefractionoftheleakagethatflashesto vapor.Thisflashfraction,FF,shouldbedeterminedusingaconstantenthalpy,h,process, basedonthemaximumtime-dependenttemperatureofthesumpwatercirculatingoutsidethe containment:FFhhhfffg=-12Where:hf1istheenthalpyofliquidatsystemdesigntemperatureandpressure;hf2istheenthalpyofliquidatsaturationconditions(14.7psia,212ºF);andhfgistheheatofvaporizationat212ºF.5.5Ifthetemperatureoftheleakageislessthan212°Forthecalculatedflashfractionislessthan10%,theamountofiodinethatbecomesairborneshouldbeassumedtobe10%ofthetotal iodineactivityintheleakedfluid,unlessasmalleramountcanbejustifiedbasedonthe actualsumppHhistoryandareaventilationrates.5.6Theradioiodinethatispostulatedtobeavailableforreleasetotheenvironmentisassumedtobe97%elementaland3%organic.Reductioninreleaseactivitybydilutionorholdupwithin buildings,orbyESFventilationfiltrationsystems,maybecreditedwhereapplicable.Filter systemsusedintheseapplicationsshouldbeevaluatedagainsttheguidanceofRegulatory Guide1.52(Ref.A-5)andGenericLetter99-02(Ref.A-6).ASSUMPTIONSONMAINSTEAMISOLATIONVALVELEAKAGEINBWRS6.ForBWRs,themainsteamisolationvalves(MSIVs)havedesignleakagethatmayresultinaradioactivityrelease.TheradiologicalconsequencesfrompostulatedMSIVleakageshouldbe analyzedandcombinedwithconsequencespostulatedforotherfissionproductreleasepathsto A-6determinethetotalcalculatedradiologicalconsequencesfromtheLOCA.ThefollowingassumptionsareacceptableforevaluatingtheconsequencesofMSIVleakage.6.1Forthepurposeofthisanalysis,theactivityavailableforreleaseviaMSIVleakageshouldbeassumedtobethatactivitydeterminedtobeinthedrywellforevaluating containmentleakage(seeRegulatoryPosition3).Nocreditshouldbeassumedfor activityreductionbythesteamseparatorsorbyiodinepartitioninginthereactorvessel.6.2AlltheMSIVsshouldbeassumedtoleakatthemaximumleakrateabovewhichthetechnicalspecificationswouldrequiredeclaringtheMSIVsinoperable.Theleakage shouldbeassumedtocontinueforthedurationoftheaccident.Postulatedleakagemay bereducedafterthefirst24hours,ifsupportedbysite-specificanalyses,toavaluenot lessthan50%ofthemaximumleakrate.6.3ReductionoftheamountofreleasedradioactivitybydepositionandplateoutonsteamsystempipingupstreamoftheoutboardMSIVsmaybecredited,buttheamountof reductioninconcentrationallowedwillbeevaluatedonanindividualcasebasis.

Generally,themodelshouldbebasedontheassumptionofwell-mixedvolumes,but othermodelssuchasslugflowmaybeusedifjustified.6.4IntheabsenceofcollectionandtreatmentofreleasesbyESFssuchastheMSIVleakagecontrolsystem,orasdescribedinparagraph6.5below,theMSIVleakageshouldbe assumedtobereleasedtotheenvironmentasanunprocessed,ground-levelrelease.

Holdupanddilutionintheturbinebuildingshouldnotbeassumed.6.5AreductioninMSIVreleasesthatisduetoholdupanddepositioninmainsteampipingdownstreamoftheMSIVsandinthemaincondenser,includingthetreatmentofair ejectoreffluentbyoffgassystems,maybecreditedifthecomponentsandpipingsystems usedinthereleasepatharecapableofperformingtheirsafetyfunctionduringand followingasafeshutdownearthquake(SSE).Theamountofreductionallowedwillbe evaluatedonanindividualcasebasis.ReferencesA-9andA-10provideguidanceon acceptablemodels.ASSUMPTIONONCONTAINMENTPURGING7.Theradiologicalconsequencesfrompost-LOCAprimarycontainmentpurgingasacombustiblegasorpressurecontrolmeasureshouldbeanalyzed.Iftheinstalledcontainment purgingcapabilitiesaremaintainedforpurposesofsevereaccidentmanagementandarenot creditedinanydesignbasisanalysis,radiologicalconsequencesneednotbeevaluated.Ifthe primarycontainmentpurgingisrequiredwithin30daysoftheLOCA,theresultsofthisanalysis shouldbecombinedwithconsequencespostulatedforotherfissionproductreleasepathsto determinethetotalcalculatedradiologicalconsequencesfromtheLOCA.Reductioninthe amountofradioactivematerialreleasedviaESFfiltersystemsmaybetakenintoaccount providedthatthesesystemsmeettheguidanceinRegulatoryGuide1.52(Ref.A-5)andGeneric Letter99-02(Ref.A-6).

A-7AppendixAREFERENCESA-1USNRC,"StandardReviewPlanfortheReviewofSafetyAnalysisReportsforNuclearPowerPlants,"NUREG-0800.A-2D.A.Powersetal,"ASimplifiedModelofAerosolRemovalbyNaturalProcessesinReactorContainments,"NUREG/CR-6189,USNRC,July1996.A-3S.L.Humphreysetal.,"RADTRAD:ASimplifiedModelforRadionuclideTransportandRemovalandDoseEstimation,"NUREG/CR-6604,USNRC,April1998.A-4D.A.PowersandS.B.Burson,"ASimplifiedModelofAerosolRemovalbyContainmentSprays,"NUREG/CR-5966,USNRC,June1993.A-5USNRC,"Design,Testing,andMaintenanceCriteriaforPostaccidentEngineered-Safety-FeatureAtmosphereCleanupSystemAirFiltrationandAdsorptionUnitsofLight- Water-CooledNuclearPowerPlants,"RegulatoryGuide1.52,Revision2,March1978.A-6USNRC,"LaboratoryTestingofNuclearGradeActivatedCharcoal,"GenericLetter99-02,June3,1999.A-7USNRC,"PotentialRadioactiveLeakagetoTankVentedtoAtmosphere,"InformationNotice91-56,September19,1991.A-8USNRC,"ClarificationofTMIActionPlanRequirements,"NUREG-0737,November1980.A-9J.E.Cline,"MSIVLeakageIodineTransportAnalysis,"LetterReportdatedMarch26,1991.(ADAMSAccessionNumberML003683718)A-10USNRC,"SafetyEvaluationofGETopicalReport,NEDC-31858P(ProprietaryGEreport),Revision2,BWROGReportforIncreasingMSIVLeakageLimitsandEliminationofLeakageControlSystems,September1993,"letterdatedMarch3,1999,ADAMSAccessionNumber990311030 B-1AppendixBASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAFUELHANDLINGACCIDENTThisappendixprovidesassumptionsacceptabletothestaffforevaluatingtheradiologicalconsequencesofafuelhandlingaccidentatlightwaterreactors.Theseassumptionssupplement theguidanceprovidedinthemainbodyofthisguide.1.SOURCETERMAcceptableassumptionsregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisguide.Thefollowingassumptionsalso apply.1.1Thenumberoffuelrodsdamagedduringtheaccidentshouldbebasedonaconservativeanalysisthatconsidersthemostlimitingcase.Thisanalysisshouldconsiderparameters suchastheweightofthedroppedheavyloadortheweightofadroppedfuelassembly (plusanyattachedhandlinggrapples),theheightofthedrop,andthecompression, torsion,andshearstressesontheirradiatedfuelrods.Damagetoadjacentfuel assemblies,ifapplicable(e.g.,eventsoverthereactorvessel),shouldbeconsidered.1.2ThefissionproductreleasefromthebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimateofthenumberoffuelrodsbreached.Allthegapactivityin thedamagedrodsisassumedtobeinstantaneouslyreleased.Radionuclidesthatshould beconsideredincludexenons,kryptons,halogens,cesiums,andrubidiums.1.3Thechemicalformofradioiodinereleasedfromthefueltothespentfuelpoolshouldbeassumedtobe95%cesiumiodide(CsI),4.85percentelementaliodine,and0.15percent organiciodide.TheCsIreleasedfromthefuelisassumedtocompletelydissociateinthe poolwater.BecauseofthelowpHofthepoolwater,theiodinere-evolvesaselemental iodine.Thisisassumedtooccurinstantaneously.TheNRCstaffwillconsider,ona case-by-casebasis,justifiablemechanistictreatmentoftheiodinereleasefromthepool.2.WATERDEPTHIfthedepthofwaterabovethedamagedfuelis23feetorgreater,thedecontamina-tionfactorsfortheelementalandorganicspeciesare500and1,respectively,givinganoverall effectivedecontaminationfactorof200(i.e.,99.5%ofthetotaliodinereleasedfromthe damagedrodsisretainedbythewater).Thisdifferenceindecontaminationfactorsforelemental (99.85%)andorganiciodine(0.15%)speciesresultsintheiodineabovethewaterbeing composedof57%elementaland43%organicspecies.Ifthedepthofwaterisnot23feet,the decontaminationfactorwillhavetobedeterminedonacase-by-casemethod(Ref.B-1).

1Theseanalysesshouldconsiderthetimefortheradioactivityconcentrationtoreachlevelscorrespondingtothemonitorsetpoint,instrumentlinesamplingtime,detectorresponsetime,diversiondamperalignmenttime,andfiltersystemactuation,as applicable.2Containmentisolationdoesnotimplycontainmentintegrityasdefinedbytechnicalspecificationsfornon-shutdownmodes.Thetermisolationisusedherecollectivelytoencompassbothcontainmentintegrityandcontainmentclosure,typicallyinplace duringshutdownperiods.Tobecreditedintheanalysis,theappropriateformofisolationshouldbeaddressedintechnical specifications.B-23.NOBLEGASESTheretentionofnoblegasesinthewaterinthefuelpoolorreactorcavityisnegligible(i.e.,decontaminationfactorof1).Particulateradionuclidesareassumedtoberetainedbythe waterinthefuelpoolorreactorcavity(i.e.,infinitedecontaminationfactor).4.FUELHANDLINGACCIDENTSWITHINTHEFUELBUILDINGForfuelhandlingaccidentspostulatedtooccurwithinthefuelbuilding,thefollowingassumptionsareacceptabletotheNRCstaff.4.1Theradioactivematerialthatescapesfromthefuelpooltothefuelbuildingisassumedtobereleasedtotheenvironmentovera2-hourtimeperiod.4.2Areductionintheamountofradioactivematerialreleasedfromthefuelpoolbyengineeredsafetyfeature(ESF)filtersystemsmaybetakenintoaccountprovidedthese systemsmeettheguidanceofRegulatoryGuide1.52andGenericLetter99-02(Refs.B-2, B-3).Delaysinradiationdetection,actuationoftheESFfiltrationsystem,ordiversionof ventilationflowtotheESFfiltrationsystem1shouldbedeterminedandaccountedforintheradioactivityreleaseanalyses.4.3TheradioactivityreleasefromthefuelpoolshouldbeassumedtobedrawnintotheESFfiltrationsystemwithoutmixingordilutioninthefuelbuilding.Ifmixingcanbe demonstrated,creditformixinganddilutionmaybeconsideredonacase-by-casebasis.

Thisevaluationshouldconsiderthemagnitudeofthebuildingvolumeandexhaustrate, thepotentialforbypasstotheenvironment,thelocationofexhaustplenumsrelativeto thesurfaceofthepool,recirculationventilationsystems,andinternalwallsandfloorsthat impedestreamflowbetweenthesurfaceofthepoolandtheexhaustplenums.5.FUELHANDLINGACCIDENTSWITHINCONTAINMENTForfuelhandlingaccidentspostulatedtooccurwithinthecontainment,thefollowingassumptionsareacceptabletotheNRCstaff.5.1Ifthecontainmentisisolated2duringfuelhandlingoperations,noradiologicalconsequencesneedtobeanalyzed.5.2Ifthecontainmentisopenduringfuelhandlingoperations,butdesignedtoautomaticallyisolateintheeventofafuelhandlingaccident,thereleasedurationshouldbebasedon 3Thestaffwillgenerallyrequirethattechnicalspecificationsallowingsuchoperationsincludeadministrativecontrolstoclosetheairlock,hatch,oropenpenetrationswithin30minutes.Suchadminstrativecontrolswillgenerallyrequirethatadedicated individualbepresent,withnecessaryequipmentavailable,torestorecontainmentclosureshouldafuelhandlingaccidentoccur.Radiologicalanalysesshouldgenerallynotcreditthismanualisolation.B-3delaysinradiationdetectionandcompletionofcontainmentisolation.Ifitcanbeshownthatcontainmentisolationoccursbeforeradioactivityisreleasedtotheenvironment,1noradiologicalconsequencesneedtobeanalyzed.5.3Ifthecontainmentisopenduringfuelhandlingoperations(e.g.,personnelairlockorequipmenthatchisopen),3theradioactivematerialthatescapesfromthereactorcavitypooltothecontainmentisreleasedtotheenvironmentovera2-hourtimeperiod.5.4AreductionintheamountofradioactivematerialreleasedfromthecontainmentbyESFfiltersystemsmaybetakenintoaccountprovidedthatthesesystemsmeettheguidanceof RegulatoryGuide1.52andGenericLetter99-02(Refs.B-2andB-3).Delaysinradiation detection,actuationoftheESFfiltrationsystem,ordiversionofventilationflowtothe ESFfiltrationsystemshouldbedeterminedandaccountedforintheradioactivityrelease analyses.15.5Creditfordilutionormixingoftheactivityreleasedfromthereactorcavitybynaturalorforcedconvectioninsidethecontainmentmaybeconsideredonacase-by-casebasis.

Suchcreditisgenerallylimitedto50%ofthecontainmentfreevolume.Thisevaluation shouldconsiderthemagnitudeofthecontainmentvolumeandexhaustrate,thepotential forbypasstotheenvironment,thelocationofexhaustplenumsrelativetothesurfaceof thereactorcavity,recirculationventilationsystems,andinternalwallsandfloorsthat impedestreamflowbetweenthesurfaceofthereactorcavityandtheexhaustplenum B-4AppendixBREFERENCESB-1.G.Burley,"EvaluationofFissionProductReleaseandTransport,"StaffTechnicalPaper,1971.(NRCAccessionnumber8402080322inADAMSorPARS)B-2.USNRC,"Design,Testing,andMaintenanceCriteriaforPostaccidentEngineered-Safety-FeatureAtmosphereCleanupSystemAirFiltrationandAdsorptionUnitsofLight-Water- CooledNuclearPowerPlants,"RegulatoryGuide1.52,Revision2,March1978.B-3.USNRC,"LaboratoryTestingofNuclearGradeActivatedCharcoal,"GenericLetter99-02,June3,199 Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingthedoseequivalentI-131 (DEI-131),onlytheradioiodineassociatedwithnormaloperationsoriodinespikesshouldbeincluded.Activityfromprojected fueldamageshouldnotbeincluded.2Ifthereareforcedflowpathsfromtheturbineorcondenser,suchasunisolatedmotorvacuumpumpsorunprocessedairejectors,theleakagerateshouldbeassumedtobetheflowrateassociatedwiththemostlimitingofthesepaths.Creditfor collectionandprocessingofreleases,suchasbyoffgasorstandbygastreatment,willbeconsideredonacase-by-casebasis.C-1AppendixCASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFABWRRODDROPACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofaroddropaccidentatBWRlight-waterreactors.These assumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryareprovidedinRegulatoryPosition3ofthisguide.Fortheroddropaccident,thereleasefromthebreachedfuel isbasedontheestimateofthenumberoffuelrodsbreachedandtheassumptionthat10%ofthe coreinventoryofthenoblegasesandiodinesisinthefuelgap.Thereleaseattributedtofuel meltingisbasedonthefractionofthefuelthatreachesorexceedstheinitiationtemperaturefor fuelmeltingandontheassumptionthat100%ofthenoblegasesand50%oftheiodines containedinthatfractionarereleasedtothereactorcoolant.2.Ifnoorminimal1fueldamageispostulatedforthelimitingevent,thereleasedactivityshouldbethemaximumcoolantactivity(typically4µCi/gmDEI-131)allowedbythetechnical specifications.3.TheassumptionsacceptabletotheNRCstaffthatarerelatedtothetransport,reduction,andreleaseofradioactivematerialfromthefuelandthereactorcoolantareasfollows.3.1Theactivityreleasedfromthefuelfromeitherthegaporfromfuelpelletsisassumedtobeinstantaneouslymixedinthereactorcoolantwithinthepressurevessel.3.2Creditshouldnotbeassumedforpartitioninginthepressurevesselorforremovalbythesteamseparators.3.3Oftheactivityreleasedfromthereactorcoolantwithinthepressurevessel,100%ofthenoblegases,10%oftheiodine,and1%oftheremainingradionuclidesareassumedto reachtheturbineandcondensers.3.4Oftheactivitythatreachestheturbineandcondenser,100%ofthenoblegases,10%oftheiodine,and1%oftheparticulateradionuclidesareavailableforreleasetothe environment.Theturbineandcondensersleaktotheatmosphereasaground-level releaseatarateof1%perday2foraperiodof24hours,atwhichtimetheleakageisassumedtoterminate.Nocreditshouldbeassumedfordilutionorholdupwithinthe C-2turbinebuilding.Radioactivedecayduringholdupintheturbineandcondensermaybeassumed.3.5Inlieuofthetransportassumptionsprovidedinparagraphs3.2through3.4above,amoremechanisticanalysismaybeusedonacase-by-casebasis.Suchanalysesaccountforthe quantityofcontaminatedsteamcarriedfromthepressurevesseltotheturbineand condensersbasedonareviewoftheminimumtransporttimefromthepressurevesselto thefirstmainsteamisolation(MSIV)andconsidersMSIVclosuretime.3.6Theiodinespeciesreleasedfromthereactorcoolantwithinthepressurevesselshouldbeassumedtobe95%CsIasanaerosol,4.85%elemental,and0.15%organic.Therelease fromtheturbineandcondensershouldbeassumedtobe97%elementaland3%organi Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingdoseequivalentI-131(DEI-

131),onlytheradioiodineassociatedwithnormaloperationsoriodinespikesshouldbeincluded.Activityfromprojectedfuel damageshouldnotbeincluded.D-1AppendixDASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFABWRMAINSTEAMLINEBREAKACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofamainsteamlineaccidentatBWRlightwaterreactors.These assumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.SOURCETERM1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisguide.Thereleasefrom thebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimateofthe numberoffuelrodsbreached.2.Ifnoorminimal1fueldamageispostulatedforthelimitingevent,thereleasedactivityshouldbethemaximumcoolantactivityallowedbytechnicalspecification.Theiodine concentrationintheprimarycoolantisassumedtocorrespondtothefollowingtwocasesinthe nuclearsteamsupplysystemvendor'sstandardtechnicalspecifications.2.1Theconcentrationthatisthemaximumvalue(typically4.0µCi/gmDEI-131)permittedandcorrespondstotheconditionsofanassumedpre-accidentspike,and2.1Theconcentrationthatisthemaximumequilibriumvalue(typically0.2µCi/gmDEI-131)permittedforcontinuedfullpoweroperation.3.Theactivityreleasedfromthefuelshouldbeassumedtomixinstantaneouslyandhomogeneouslyinthereactorcoolant.Noblegasesshouldbeassumedtoenterthesteamphase instantaneously.TRANSPORT4.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows.4.1Themainsteamlineisolationvalves(MSIV)shouldbeassumedtocloseinthemaximumtimeallowedbytechnicalspecifications.4.2Thetotalmassofcoolantreleasedshouldbeassumedtobethatamountinthesteamlineandconnectinglinesatthetimeofthebreakplustheamountthatpassesthroughthe valvespriortoclosur D-24.3Alltheradioactivityinthereleasedcoolantshouldbeassumedtobereleasedtotheatmosphereinstantaneouslyasaground-levelrelease.Nocreditshouldbeassumedfor plateout,holdup,ordilutionwithinfacilitybuildings.4.4Theiodinespeciesreleasedfromthemainsteamlineshouldbeassumedtobe95%CsIasanaerosol,4.85%elemental,and0.15%organi Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity,"foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.2Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingdoseequivalentI-131(DEI-

131),onlytheradioiodineassociatedwithnormaloperationsoriodinespikesshouldbeincluded.Activityfromprojectedfuel damageshouldnotbeincluded.E-1AppendixEASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRMAINSTEAMLINEBREAKACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofamainsteamlinebreakaccidentatPWRlightwaterreactors.

Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.1SOURCETERMS1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisregulatoryguide.The releasefromthebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimate ofthenumberoffuelrodsbreached.Thefueldamageestimateshouldassumethatthehighest worthcontrolrodisstuckatitsfullywithdrawnposition.2.Ifnoorminimal2fueldamageispostulatedforthelimitingevent,theactivityreleasedshouldbethemaximumcoolantactivityallowedbythetechnicalspecifications.Twocasesof iodinespikingshouldbeassumed.2.1Areactortransienthasoccurredpriortothepostulatedmainsteamlinebreak(MSLB)andhasraisedtheprimarycoolantiodineconcentrationtothemaximumvalue(typically 60µCi/gmDEI-131)permittedbythetechnicalspecifications(i.e.,apreaccidentiodine spikecase).2.2TheprimarysystemtransientassociatedwiththeMSLBcausesaniodinespikeintheprimarysystem.Theincreaseinprimarycoolantiodineconcentrationisestimatedusinga spikingmodelthatassumesthattheiodinereleaseratefromthefuelrodstotheprimary coolant(expressedincuriesperunittime)increasestoavalue500timesgreaterthanthe releaseratecorrespondingtotheiodineconcentrationattheequilibriumvalue(typically 1.0µCi/gmDEI-131)specifiedintechnicalspecifications(i.e.,concurrentiodinespike case).Aconcurrentiodinespikeneednotbeconsiderediffueldamageispostulated.

Theassumediodinespikedurationshouldbe8hours.Shorterspikedurationsmaybe consideredonacase-by-casebasisifitcanbeshownthattheactivityreleasedbythe8- hourspikeexceedsthatavailableforreleasefromthefuelgapofallfuelpins.3.Theactivityreleasedfromthefuelshouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughtheprimarycoolan Inthisappendix,rupturedreferstothestateofthesteamgeneratorinwhichprimary-to-secondaryleakageratehasincreasedtoavaluegreaterthantechnicalspecifications.FaultedreferstothestateofthesteamgeneratorinwhichthesecondarysidehasbeendepressurizedbyaMSLBsuchthatprotectivesystemresponse(mainsteamlineisolation,reactortrip,safetyinjection,etc.)hasoccurred.PartitioningCoefficientisdefinedas:PCmassofIperunitmassofliquidmassofIperunitmassofgas=22E-24.Thechemicalformofradioiodinereleasedfromthefuelshouldbeassumedtobe95%cesiumiodide(CsI),4.85percentelementaliodine,and0.15percentorganiciodide.Iodine releasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elemental and3%organic.Thesefractionsapplytoiodinereleasedasaresultoffueldamageandtoiodine releasedduringnormaloperations,includingiodinespiking.TRANSPORT35.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows.5.1Forfacilitiesthathavenotimplementedalternativerepaircriteria(seeRef.E-1,DG-1074),theprimary-to-secondaryleakrateinthesteamgeneratorsshouldbeassumedto betheleakratelimitingconditionforoperationspecifiedinthetechnicalspecifications.

Forfacilitieswithtraditionalgeneratorspecifications(bothpergeneratorandtotalofall generators),theleakageshouldbeapportionedbetweenaffectedandunaffectedsteam generatorsinsuchamannerthatthecalculateddoseismaximized.5.2Thedensityusedinconvertingvolumetricleakrates(e.g.,gpm)tomassleakrates(e.g.,lbm/hr)shouldbeconsistentwiththebasisoftheparameterbeingconverted.TheARC leakratecorrelationsaregenerallybasedonthecollectionofcooledliquid.Surveillance testsandfacilityinstrumentationusedtoshowcompliancewithleakratetechnical specificationsaretypicallybasedoncooledliquid.Inmostcases,thedensityshouldbe assumedtobe1.0gm/cc(62.4lbm/ft3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureofthe leakageislessthan100°C(212°F).Thereleaseofradioactivityfromunaffectedsteam generatorsshouldbeassumedtocontinueuntilshutdowncoolingisinoperationand releasesfromthesteamgeneratorshavebeenterminated.5.4Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.5.5Thetransportmodeldescribedinthissectionshouldbeutilizedforiodineandparticulatereleasesfromthesteamgenerators.ThismodelisshowninFigureE-1andsummarized below:

E-3SteamSpaceBulkWaterPrimaryLeakageScrubbingPartitioningReleaseFigureE-1TransportModel5.5.1Aportionoftheprimary-to-secondaryleakagewillflashtovapor,basedonthethermodynamicconditionsinthereactorandsecondarycoolant.*Duringperiodsofsteamgeneratordryout,alloftheprimary-to-secondaryleakageisassumedtoflashtovaporandbereleasedtotheenvironmentwithnomitigation.*Withregardtotheunaffectedsteamgeneratorsusedforplantcooldown,theprimary-to-secondaryleakagecanbeassumedtomixwiththesecondarywaterwithoutflashingduringperiodsoftotaltube submergence.5.5.2Theleakagethatimmediatelyflashestovaporwillrisethroughthebulkwaterofthesteamgeneratorandenterthesteamspace.Creditmaybetakenforscrubbing inthegenerator,usingthemodelsinNUREG-0409,"IodineBehaviorinaPWR CoolingSystemFollowingaPostulatedSteamGeneratorTubeRuptureAccident" (Ref.E-2),duringperiodsoftotalsubmergenceofthetubes.5.5.3Theleakagethatdoesnotimmediatelyflashisassumedtomixwiththebulkwater.5.5.4Theradioactivityinthebulkwaterisassumedtobecomevaporataratethatisthefunctionofthesteamingrateandthepartitioncoefficient.Apartitioncoefficient foriodineof100maybeassumed.Theretentionofparticulateradionuclidesin thesteamgeneratorsislimitedbythemoisturecarryoverfromthesteam generators.5.6Operatingexperienceandanalyseshaveshownthatforsomesteamgeneratordesigns,tubeuncoverymayoccurforashortperiodfollowinganyreactortrip(Ref.E-3).The potentialimpactoftubeuncoveryonthetransportmodelparameters(e.g.,flashfraction, scrubbingcredit)needstobeconsidered.Theimpactofemergencyoperatingprocedure restorationstrategiesonsteamgeneratorwaterlevelsshouldbeevaluate E-4AppendixEREFERENCESE-1USNRC,"SteamGeneratorTubeIntegrity,"DraftRegulatoryGuideDG-1074,December1998.E-2.USNRC,"IodineBehaviorinaPWRCoolingSystemFollowingaPostulatedSteamGeneratorTubeRuptureAccident,"NUREG-0409,May1985.E-3USNRC,"SteamGeneratorTubeRuptureAnalysisDeficiency,"InformationNotice88-31,May25,198 Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity"(USNRC,December1998),foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.2Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingdoseequivalentI-131(DEI-

131),onlytheradioiodineassociatedwithnormaloperationsoriodinespikesshouldbeincluded.Activityfromprojectedfuel damageshouldnotbeincluded.F-1AppendixFASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRSTEAMGENERATORTUBERUPTUREACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofasteamgeneratortuberuptureaccidentatPWRlight-water reactors.Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.1SOURCETERM1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareinRegulatoryPosition3ofthisguide.Thereleasefromthe breachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimateofthenumberof fuelrodsbreached.2.Ifnoorminimal2fueldamageispostulatedforthelimitingevent,theactivityreleasedshouldbethemaximumcoolantactivityallowedbytechnicalspecification.Twocasesofiodine spikingshouldbeassumed.2.1Areactortransienthasoccurredpriortothepostulatedsteamgeneratortuberupture(SGTR)andhasraisedtheprimarycoolantiodineconcentrationtothemaximumvalue (typically60µCi/gmDEI-131)permittedbythetechnicalspecifications(i.e.,a preaccidentiodinespikecase).2.2TheprimarysystemtransientassociatedwiththeSGTRcausesaniodinespikeintheprimarysystem.Theincreaseinprimarycoolantiodineconcentrationisestimatedusing aspikingmodelthatassumesthattheiodinereleaseratefromthefuelrodstotheprimary coolant(expressedincuriesperunittime)increasestoavalue335timesgreaterthanthe releaseratecorrespondingtotheiodineconcentrationattheequilibriumvalue(typically 1.0µCi/gmDEI-131)specifiedintechnicalspecifications(i.e.,concurrentiodinespike case).Aconcurrentiodinespikeneednotbeconsiderediffueldamageispostulated.

Theassumediodinespikedurationshouldbe8hours.Shorterspikedurationsmaybe consideredonacase-by-casebasisifitcanbeshownthattheactivityreleasedbythe8- hourspikeexceedsthatavailableforreleasefromthefuelgapofallfuelpins.3.Theactivityreleasedfromthefuel,ifany,shouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughtheprimarycoolan Inthisappendix,rupturedreferstothestateofthesteamgeneratorinwhichprimary-to-secondaryleakageratehasincreasedtoavaluegreaterthantechnicalspecifications.F-24.Iodinereleasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elementaland3%organic.TRANSPORT35.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows:5.1Theprimary-to-secondaryleakrateinthesteamgeneratorsshouldbeassumedtobetheleakratelimitingconditionforoperationspecifiedinthetechnicalspecifications.The leakageshouldbeapportionedbetweenaffectedandunaffectedsteamgeneratorsinsuch amannerthatthecalculateddoseismaximized.5.2Thedensityusedinconvertingvolumetricleakrates(e.g.,gpm)tomassleakrates(e.g.,lbm/hr)shouldbeconsistentwiththebasisofsurveillancetestsusedtoshowcompliance withleakratetechnicalspecifications.Thesetestsaretypicallybasedoncoolliquid.

Facilityinstrumentationusedtodetermineleakageistypicallylocatedonlinescontaining coolliquids.Inmostcases,thedensityshouldbeassumedtobe1.0gm/cc(62.4lbm/ft3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureoftheleakageislessthan100°C(212°F).Thereleaseofradioactivityfromtheunaffectedsteamgeneratorsshouldbeassumedtocontinueuntilshutdowncoolingisinoperationandreleasesfromthesteamgeneratorshavebeenterminated.5.4Thereleaseoffissionproductsfromthesecondarysystemshouldbeevaluatedwiththeassumptionofacoincidentlossofoffsitepower.5.5Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.5.6ThetransportmodeldescribedinRegulatoryPositions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulate Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity"(USNRC,December1998),foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.G-1AppendixGASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRLOCKEDROTORACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofalockedrotoraccidentatPWRlightwaterreactors.1Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.SOURCETERM1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareinRegulatoryPosition3ofthisregulatoryguide.Therelease fromthebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimateofthe numberoffuelrodsbreached.2.Ifnofueldamageispostulatedforthelimitingevent,aradiologicalanalysisisnotrequiredastheconsequencesofthiseventareboundedbytheconsequencesprojectedforthe mainsteamlinebreakoutsidecontainment.3.Theactivityreleasedfromthefuelshouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughtheprimarycoolant.4.Thechemicalformofradioiodinereleasedfromthefuelshouldbeassumedtobe95%cesiumiodide(CsI),4.85percentelementaliodine,and0.15percentorganiciodide.Iodine releasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elemental and3%organic.Thesefractionsapplytoiodinereleasedasaresultoffueldamageandtoiodine releasedduringnormaloperations,includingiodinespiking.RELEASETRANSPORT5.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows.5.1Theprimary-to-secondaryleakrateinthesteamgeneratorsshouldbeassumedtobetheleak-rate-limitingconditionforoperationspecifiedinthetechnicalspecifications.The leakageshouldbeapportionedbetweenthesteamgeneratorsinsuchamannerthatthe calculateddoseismaximized.5.2Thedensityusedinconvertingvolumetricleakrates(e.g.,gpm)tomassleakrates(e.g.,lbm/hr)shouldbeconsistentwiththebasisofsurveillancetestsusedtoshowcompliance withleakratetechnicalspecifications.Thesetestsaretypicallybasedoncoolliqui G-2Facilityinstrumentationusedtodetermineleakageistypicallylocatedonlinescontainingcoolliquids.Inmostcases,thedensityshouldbeassumedtobe1.0gm/cc(62.4lbm/ft3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureoftheleakageislessthan100°C(212°F).Thereleaseofradioactivityshouldbeassumedtocontinueuntilshutdowncoolingisinoperationandreleasesfromthesteamgeneratorshavebeenterminated.5.4Thereleaseoffissionproductsfromthesecondarysystemshouldbeevaluatedwiththeassumptionofacoincidentlossofoffsitepower.5.5Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.5.6Thetransportmodeldescribedinassumptions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulate Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity"(USNRC,December1998),foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.H-1AppendixHASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRRODEJECTIONACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofarodejectionaccidentatPWRlightwaterreactors.1Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.SOURCETERM1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryareinRegulatoryPosition3ofthisguide.Fortherodejectionaccident,thereleasefromthebreachedfuelisbased ontheestimateofthenumberoffuelrodsbreachedandtheassumptionthat10%ofthecore inventoryofthenoblegasesandiodinesisinthefuelgap.Thereleaseattributedtofuelmelting isbasedonthefractionofthefuelthatreachesorexceedstheinitiationtemperatureforfuel meltingandtheassumptionthat100%ofthenoblegasesand25%oftheiodinescontainedin thatfractionareavailableforreleasefromcontainment.Forthesecondarysystemrelease pathway,100%ofthenoblegasesand50%oftheiodinesinthatfractionarereleasedtothe reactorcoolant.2.Ifnofueldamageispostulatedforthelimitingevent,aradiologicalanalysisisnotrequiredastheconsequencesofthiseventareboundedbytheconsequencesprojectedforthe loss-of-coolantaccident(LOCA),mainsteamlinebreak,andsteamgeneratortuberupture.3.Tworeleasecasesaretobeconsidered.Inthefirst,100%oftheactivityreleasedfromthefuelshouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughthe containmentatmosphere.Inthesecond,100%oftheactivityreleasedfromthefuelshouldbe assumedtobecompletelydissolvedintheprimarycoolantandavailableforreleasetothe secondarysystem.4.Thechemicalformofradioiodinereleasedtothecontainmentatmosphereshouldbeassumedtobe95%cesiumiodide(CsI),4.85%elementaliodine,and0.15%organiciodide.If containmentspraysdonotactuateorareterminatedpriortoaccumulatingsumpwater,orifthe containmentsumppHisnotcontrolledatvaluesof7orgreater,theiodinespeciesshouldbe evaluatedonanindividualcasebasis.EvaluationsofpHshouldconsidertheeffectofacids createdduringtherodejectionaccidentevent,e.g.,pyrolysisandradiolysisproducts.Withthe exceptionofelementalandorganiciodineandnoblegases,fissionproductsshouldbeassumed tobeinparticulateform.5.Iodinereleasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elementaland3%organi H-2TRANSPORTFROMCONTAINMENT6.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromthecontainmentareasfollows.6.1Areductionintheamountofradioactivematerialavailableforleakagefromthecontainmentthatisduetonaturaldeposition,containmentsprays,recirculatingfilter systems,dualcontainments,orotherengineeredsafetyfeaturesmaybetakeninto account.RefertoAppendixAtothisguideforguidanceonacceptablemethodsand assumptionsforevaluatingthesemechanisms.6.2Thecontainmentshouldbeassumedtoleakattheleakrateincorporatedinthetechnicalspecificationsatpeakaccidentpressureforthefirst24hours,andat50%ofthisleakrate fortheremainingdurationoftheaccident.Peakaccidentpressureisthemaximum pressuredefinedinthetechnicalspecificationsforcontainmentleaktesting.Leakage fromsubatmosphericcontainmentsisassumedtobeterminatedwhenthecontainmentis broughttoasubatmosphericconditionasdefinedintechnicalspecifications.TRANSPORTFROMSECONDARYSYSTEM7.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromthesecondarysystemareasfollows.7.1Aleakrateequivalenttotheprimary-to-secondaryleakratelimitingconditionforoperationspecifiedinthetechnicalspecificationsshouldbeassumedtoexistuntil shutdowncoolingisinoperationandreleasesfromthesteamgeneratorshavebeen terminated.7.2Thedensityusedinconvertingvolumetricleakrates(e.g.,gpm)tomassleakrates(e.g.,lbm/hr)shouldbeconsistentwiththebasisofsurveillancetestsusedtoshowcompliance withleakratetechnicalspecifications.Theseteststypicallyarebasedoncooledliquid.

Thefacility'sinstrumentationusedtodetermineleakagetypicallyislocatedonlines containingcoolliquids.Inmostcases,thedensityshouldbeassumedtobe1.0gm/cc (62.4lbm/ft3).7.3Allnoblegasradionuclidesreleasedtothesecondarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.7.4Thetransportmodeldescribedinassumptions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulate I-1AppendixIASSUMPTIONSFOREVALUATINGRADIATIONDOSESFOREQUIPMENTQUALIFICATIONThisappendixaddressesassumptionsassociatedwithequipmentqualificationthatareacceptabletotheNRCstaffforperformingradiologicalassessments.AsstatedinRegulatory Position6ofthisguide,thisappendixsupersedesRegulatoryPositions2.c.(1)and2.c.(2)and AppendixDofRevision1ofRegulatoryGuide1.89,"EnvironmentalQualificationofCertain ElectricEquipmentImportanttoSafetyforNuclearPowerPlants"(USNRC,June1984),for operatingreactorsthathaveamendedtheirlicensingbasistouseanalternativesourceterm.

Exceptasstatedinthisappendix,otherassumptions,methods,andprovisionsofRevision1of RegulatoryGuide1.89remaineffective.BASICASSUMPTIONS1.Gammaandbetadosesanddoseratesshouldbedeterminedforthreetypesofradioactivesourcedistributions:(1)activitysuspendedinthecontainmentatmosphere,(2)activityplatedout oncontainmentsurfaces,and(3)activitymixedinthecontainmentsumpwater.Agivenpieceof equipmentmayreceiveadosecontributionfromanyorallofthesesources.Theamountofdose contributedbyeachofthesesourcesisdeterminedbythelocationoftheequipment,thetime- dependentandlocation-dependentdistributionofthesource,andtheeffectsofshielding.ForEQ componentslocatedoutsideofthecontainment,additionalradiationsourcesmayincludepiping andcomponentsinsystemsthatcirculatecontainmentsumpwateroutsideofcontainment.

Activitydepositedinventilationandprocessfiltermediamaybeasourceofpost-accidentdose.2.Theintegrateddoseshouldbedeterminedfromestimateddoseratesusingappropriateintegrationfactorsdeterminedforeachofthemajorsourceterms(e.g.,containmentsump, containmentatmosphere,ECCS,normaloperation).Theperiodofexposureshouldbeconsistent withthesurvivabilityperiodfortheEQequipmentbeingevaluated.Thesurvivabilityperiodis themaximumduration,post-accident,thattheparticularEQcomponentisexpectedtooperate andperformitsintendedsafetyfunction.Theperiodofexposurefornormaloperationdoseis generallythedurationoftheplantlicense,i.e.,40years.FISSIONPRODUCTCONCENTRATIONS3.Theradiationenvironmentresultingfromnormaloperationsshouldbebasedontheconservativesourcetermestimatesreportedinthefacility'sSafetyAnalysisReportorshouldbe consistentwiththeprimarycoolantspecificactivitylimitscontainedinthefacility'stechnical specifications.Theuseofequilibriumprimarycoolantconcentrationsbasedon1%fuelcladding failureswouldbeoneacceptablemethod.Inestimatingtheintegrateddosefrompriornormal operations,appropriatehistoricaldoseratedatamaybeusedwhereavailable.4.Theradioactivityreleasedfromthecoreduringadesignbasisloss-of-coolantaccident(LOCA)shouldbebasedontheassumptionsprovidedinRegulatoryPosition3andAppendixA ofthisregulatoryguide.AlthoughthedesignbasisLOCAisgenerallylimitingforradiological I-2environmentalqualification(EQ)purposes,theremaybecomponentsforwhichanotherdesignbasisaccidentmaybelimiting.Inthesecases,theassumptionsprovidedinAppendicesB throughHofthisregulatoryguide,asapplicable,shouldbeused.Applicablefeaturesand mechanismsmaybeassumedinEQcalculationsprovidedthatanyprerequisitesandlimitations identifiedregardingtheirusearemet.Thereareadditionalconsiderations:*ForPWRicecondensercontainments,thesourceshouldbeassumedtobeinitiallyreleasedtothelowercontainmentcompartment.Thedistributionoftheactivityshould bebasedontheforcedrecirculationfanflowratesandthetransferratesthroughtheice bedsasfunctionsoftime.*ForBWRMarkIIIdesigns,alltheactivityshouldbeassumedinitiallyreleasedtothedrywellareaandthetransferofactivityfromtheseregionsviacontainmentleakagetothe surroundingreactorbuildingvolumeshouldbeusedtopredictthequalificationlevels withinthereactorbuilding(secondarycontainment).DOSEMODELFORCONTAINMENTATMOSPHERE5.Thebetaandgammadoseratesandintegrateddosesfromtheairborneactivitywithinthecontainmentatmosphereandfromtheplateoutofaerosolsoncontainmentsurfacesgenerally shouldbecalculatedforthemidpointinthecontainment,andthisdoserateshouldbeusedforall exposedcomponents.Radiationshieldingaffordedbyinternalstructuresmaybeneglectedfor modelingsimplicity.Itisexpectedthattheshieldingaffordedbythesestructureswouldreduce thedoseratesbyfactorsoftwoormoredependingonthespecificlocationandgeometry.More detailedcalculationsmaybewarrantedforselectedcomponentsifacceptabledoseratescannot beachievedusingthesimplermodelingassumptions.6.Becauseoftheshortrangeofthebetasinair,theairbornebetadoseratesshouldbecalculatedusinganinfinitemediummodel.Othermodels,suchasfinitecloudandsemi-infinite cloud,maybeapplicabletoselectedcomponentswithsufficientjustification.Theapplicability ofthesemi-infinitemodelwoulddependonthelocationofthecomponent,availableshielding, andreceptorgeometry.Forexample,betadoseratesforequipmentlocatedonthecontainment wallsoronlargeinternalstructuresmightbeadequatelyassessedusingthesemi-infinitemodel.

Useofafinitecloudmodelwillbeconsideredonacase-by-casemethod.7.Allgammadoseratesshouldbemultipliedbyacorrectionfactorof1.3toaccountfortheomissionofthecontributionfromthedecaychainsoftheradionuclides.Thiscorrectionis particularlyimportantfornon-gamma-emittingradionuclideshavinggammaemittingprogeny, forexample,Cs-137decaytoBa-137m.Thiscorrectionmaybeomittedifthecalculational methodexplicitlyaccountsfortheemissionsfrombuildupanddecayoftheradioactiveprogeny.DOSEMODELFORCONTAINMENTSUMPWATERSOURCES8.Withtheexceptionofnoblegases,alltheactivityreleasedfromthefuelshouldbeassumedtobetransportedtothecontainmentsumpasitisreleased.Thisactivityshouldbe assumedtomixinstantaneouslyanduniformlywithotherliquidsthatdraintothesump.This I-3transportcanalsobemodeledmechanisticallyasthetime-dependentwashoutofairborneaerosolsbytheactionofcontainmentsprays.Radionuclidesthatdonotbecomeairborneon releasefromthereactorcoolantsystem,e.g.,theyareentrainedinnon-flashedreactorcoolant, shouldbeassumedtobeinstantaneouslytransportedtothesumpandbeuniformlydistributedin thesumpwater.9.Thegammaandbetadoseratesandtheintegrateddosesshouldbecalculatedforapointlocatedonthesurfaceofthewateratthecenterlineofthelargepoolofsumpwater.Theeffects ofbuildupshouldbeconsidered.Moredetailedmodelingwithshieldinganalysiscodesmaybe performed.DOSEMODELFOREQUIPMENTLOCATEDOUTSIDECONTAINMENT10.EQequipmentlocatedoutsideofcontainmentmaybeexposedto(1)radiationfromsourceswithinthecontainmentbuilding,(2)radiationfromactivitycontainedinpipingand componentsinsystemsthatre-circulatecontainmentsumpwateroutsideofcontainment(e.g.,

ECCS,RHR,samplingsystems),(3)radiationfromactivitycontainedinpipingandcomponents insystemsthatprocesscontainmentatmosphere(e.g.,hydrogenrecombiners,purgesystems),(4)

radiationfromactivitydepositedinventilationandprocessfiltermedia,and(5)radiationfrom airborneactivityinplantareasoutsideofthecontainment(i.e.,leakagefromrecirculation systems).Theamountofdosecontributedbyeachofthesesourcesisdeterminedbythelocation oftheequipment,thetime-dependentandlocation-dependentdistributionofthesource,andthe effectsofshielding.11.BecauseofthelargeamountofEQequipmentandthecomplexityofsystemandcomponentlayoutinplantbuildings,itisgenerallynotreasonabletomodeleachEQcomponent.

Areasonableapproachistodeterminethelimitingdoseratefromallsourcesinaparticularplant area(e.g.,cubicle,floor,building)toarealorhypotheticalreceptorandtobasetheintegrated dosesforallcomponentsinthatareaonthispostulateddoserate.Individualdetailedmodeling ofselectedequipmentmaybeperformed.12.Theintegrateddosesfromcomponentsandpipinginsystemsrecirculatingsumpwatershouldassumeasourcetermbasedonthetime-dependentcontainmentsumpsourceterm describedabove.Similarly,thedosesfromcomponentsthatcontainairfromthecontainment atmosphereshouldassumeasourcetermbasedonthetime-dependentcontainmentatmosphere sourcetermdescribedabove.13.Analysesofintegrateddosescausedbyradiationfromthebuildupofactivityonventilationandprocessfiltermediainsystemscontainingcontainmentsumpwateroratmosphere orbothshouldassumethattheventilationorprocessflowisatitsnominaldesignvalueandthat thefiltermediais100%efficientforiodineandparticulates.Thedurationofflowthroughthe filtermediashouldbeconsistentwiththeplantdesignandoperatingprocedures.Radioactive decayinthefiltermediashouldbeconsidered.Shieldingbystructuresandcomponentsbetween thefilterandtheEQequipmentmaybeconsidere K-1AppendixKAcronymsASTAlternativesourcetermBWRBoilingwaterreactor CDFCoredamagefrequency CEDECommittedeffectivedoseequivalent COLRCoreoperatinglimitsreport DBADesignbasisaccident DDEDeepdoseequivalent DNBRDeparturefromnucleateboilingratio EABExclusionareaboundary EDEEffectivedoseequivalent EPAEnvironmentalProtectionAgency EQEnvironmentalqualification ESFEngineeredsafetyfeature FHAFuelhandlingaccident FSARFinalsafetyanalysisreport IPFIodineprotectionfactor LERFLargeearlyreleasefraction LOCALoss-of-coolantaccident LPZLowpopulationzone MOXMixedoxide MSLBMainsteamlinebreak NDTNon-destructivetesting NSSSNuclearsupplysystemsupplier PRAProbabilisticriskassessment PWRPressurizedwaterreactor RMSRadiationmonitoringsystem SGSteamgenerator SGTRSteamgeneratortuberupture TEDETotaleffectivedoseequivalent TIDTechnicalinformationdocument TMIThreeMileIsland VALUE/IMPACTSTATEMENTAseparatevalue/impactanalysishasnotbeenpreparedforthisRegulatoryGuide1.183.Avalue/impactanalysiswasincludedintheregulatoryanalysisfortheproposedamendmentsto 10CFRParts21,50,and54publishedonMarch11,1999(64FR12117).Thisregulatory analysiswasupdatedaspartofthefinalamendmentsto10CFRParts21,50,and54,published inDecember1999(64FR71998).Copiesofbothregulatoryanalysesareavailablefor inspectionorcopyingforafeeintheCommission'sPublicDocumentRoomat2120LStreet NW,Washington,DC,underRGINAG12.ADAMSAccessionNumberML003716792