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.Commentsandsuggestionsforimprovementsintheseguidesareencour agedatalltimes,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,Ge neral.Singlecopiesofregulatoryguides(whichmaybereproduced)maybeobtainedfreeofchargebywritingtheDistributionServicesSection,U.S.Nuclea rRegulatoryCommission,Washington,DC20555-0001,orbyfaxto(301)415-2289,orbyemailtoDISTRIBUTION@NRC.GOV.Electroniccopiesofthisguide areavailableontheinternetatNRC'shomepageat<

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REGULATORY

GUIDEOFFICEOFNUCLEARREGULATORYRESEARCHREGULATORYGUIDE1.183(DraftwasissuedasDG-1081)ALTERNATIVERADIOLOGICALSOURCETERMSFOREVALUATINGDESIGNBASISACCIDENTSATNUCLEARPOWERREACTORS

iiAVAILABILITYINFORMATIONSinglecopiesofregulatoryguides,bothactiveanddraft,anddraftNUREGdocumentsmaybeobtainedfreeofchargebywritingtheReproductionandDistributionServicesSection,OCIO,

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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,andthetimingofthereleasetothecontainmen

t. TheNRC

staffconsideredtheapplicabilityoftherevisedsourcetermstooperatingreactorsanddeterminedthat thecurrentanalyticalapproachbasedontheTID-14844sourcetermwouldcontinuetobeadequateto protectpublichealthandsafety.Operatingreactorslicensedunderthatapproachwouldnotbe requiredtore-analyzeaccidentsusingtherevisedsourceterms.TheNRCstaffalsodeterminedthat somelicenseesmightwishtouseanASTinanalysestosupportcost-beneficiallicensingactions.

3TheNUREG-1465sourcetermshaveoftenbeenreferredtoasthe"revisedsourceterms."Inrecognitionthattheremaybeadditionalsourcetermsidentifiedinthefuture,10CFR50.67addresses"alternativesourceterms."Thisregulatoryguide endorsesasourcetermderivedfromNUREG-1465andprovidesguidanceontheacceptableattributesofotheralternativesourceterms.1.183-3TheNRCstaff,therefore,initiatedseveralactionstoprovidearegulatorybasisforoperatingreactorstouseanAST

3indesignbasisanalyses.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.REGULATORYPOSITION

1. IMPLEMENTATIONOFAST

1.1GenericConsiderationsAsusedinthisguide,anASTisanaccidentsourcetermthatisdifferentfromtheaccidentsourcetermusedintheoriginaldesignandlicensingofthefacilityandthathasbeenapprovedforuse under10CFR50.67.ThisguideidentifiesanASTthatisacceptabletotheNRCstaffandidentifies significantcharacteristicsofotherASTsthatmaybefoundacceptable.WhiletheNRCstaff recognizesseveralpotentialusesofanAST,itisnotpossibletoforeseeallpossibleuse

s. 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'ssevereaccidentmanagementprogram.

4ThisplanningbasisisalsoaddressedinNUREG-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,theparticularcharacteristicshavebeentailoredspecificallyforDBAanalysisus

e. 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.

1.183-8evaluationwillnecessarilybeafunctionofthespecificproposedfacilitymodification

6 andwhetherafullorselectiveimplementationisbeingpursued.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,asensitivity analysisisanevaluationthatconsidershowtheoverallresultsvaryasaninputparameter(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 changeinanalysismethodologythatwouldrequireNRCapproval.

7

1.183-10Thisguidanceisalsoapplicabletoselectiveimplementationstotheextentthattheaffectedanalysesarewithinthescopeoftheapprovedimplementationasdescribedinthefacilitydesign basis.Inthesecases,thecharacteristicsoftheASTandTEDEcriteriaidentifiedinthefacility designbasisneedtobeconsideredinupdatingtheanalyses.Useofothercharacteristicsofthe ASTorTEDEcriteriathatarenotpartoftheapproveddesignbasis,andchangestopreviously approvedASTcharacteristics,requirespriorNRCstaffapprovalunder10CFR50.67.1.3.5EquipmentEnvironmentalQualificationCurrentenvironmentalqualification(EQ)analysesmaybeimpactedbyaproposedplantmodificationassociatedwiththeASTimplementation.TheEQanalysesthathaveassumptionsor inputsaffectedbytheplantmodificationshouldbeupdatedtoaddresstheseimpact

s. 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),providesadditionalguidanc

e. 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 quantitiesoffissionproducts.

8TheuncertaintyfactorusedindeterminingthecoreinventoryshouldbethatvalueprovidedinAppendixKto10CFRPart50,typically1.02.

9Notethatforsomeradionuclides,suchasCs-137,equilibriumwillnotbereachedpriortofueloffload.Thus,themaximuminventoryattheendoflifeshouldbeused.

1.183-12 2.2TheASTmustbeexpressedintermsoftimesandratesofappearanceofradioactivefissionproductsreleasedintocontainment,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.

DeviationsfromthisguidancemustbeevaluatedagainstRegulatoryPosition

2. AftertheNRC

staffhasapprovedanimplementationofanAST,subsequentchangestotheASTwillrequireNRC

staffreviewunder10CFR50.67.3.1FissionProductInventoryTheinventoryoffissionproductsinthereactorcoreandavailableforreleasetothecontainmentshouldbebasedonthemaximumfullpoweroperationofthecorewith,asa minimum,currentlicensedvaluesforfuelenrichment,fuelburnup,andanassumedcorepower equaltothecurrentlicensedratedthermalpowertimestheECCSevaluationuncertainty.

8 Theperiodofirradiationshouldbeofsufficientdurationtoallowtheactivityofdose-significant radionuclidestoreachequilibriumortoreachmaximumvalues.

9ThecoreinventoryshouldbedeterminedusinganappropriateisotopegenerationanddepletioncomputercodesuchasORIGEN

2(Ref.17)orORIGEN-ARP(Ref.18).Coreinventoryfactors(Ci/MWt)providedinTID14844 andusedinsomeanalysiscomputercodeswerederivedforlowburnup,lowenrichmentfueland shouldnotbeusedwithhigherburnupandhigherenrichmentfuels.

10ThereleasefractionslistedherehavebeendeterminedtobeacceptableforusewithcurrentlyapprovedLWRfuelwithapeakburnupupto62,000MWD/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.3.2ReleaseFractions

10Thecoreinventoryreleasefractions,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 Lanthanides0.000.00020.0002Table3 11Non-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 theonsetoftheprojecteddamage.

1.183-15Table4LOCAReleasePhasesPWRsBWRsPhaseOnsetDurationOnsetDurationGapRelease30sec0.5hr2min0.5hrEarlyIn-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 guideprovideadditionaldetails.

13ThepriorpracticeofbasinginhalationexposureononlyradioiodineandnotincludingradioiodineinexternalexposurecalculationsisnotconsistentwiththedefinitionofTEDEandthecharacteristicsoftherevisedsourceterm.

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.

13 4.1.2Theexposure-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

-4cubicmeterspersecond.

14WithregardtotheEABTEDE,themaximumtwo-hourvalueisthebasisforscreeningandevaluationunder10CFR50.59.Changestodosesoutsideofthetwo-hourwindowareonlyconsideredinthecontextoftheirimpactonthemaximumtwo-hour EABTEDE.1.183-17 4.1.4TheDDEshouldbecalculatedassumingsubmergenceinsemi-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.

4.2.3Themodelsusedtotransportradioactivematerialintoandthroughthecontrol room, 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

-4cubicmeterspersecond.

1.183-19 4.2.7ControlroomdosesshouldbecalculatedusingdoseconversionfactorsidentifiedinRegulatoryPosition4.1aboveforuseinoffsitedoseanalyses.TheDDEfromphotonsmaybe correctedforthedifferencebetweenfinitecloudgeometryinthecontrolroomandthesemi- infinitecloudassumptionusedincalculatingthedoseconversionfactors.Thefollowing expressionmaybeusedtocorrectthesemi-infiniteclouddose,DDE,toafiniteclouddose, DDEfinite,wherethecontrolroomismodeledasahemispherethathasavolume,V,incubicfeet,equivalenttothatofthecontrolroom(Ref.22).Equation1 DDEDDEVfinite=¥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.

1.183-20Table6 17AccidentDoseCriteriaAccidentorCaseEABandLPZDoseCriteriaAnalysisReleaseDurationLOCA25remTEDE30daysforcontainment,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.Methodologiesthathavebeenusedfordetermining c/QvaluesaredocumentedinRegulatoryGuides1.3and1.4,RegulatoryGuide1.145,"AtmosphericDispersionModelsforPotentialAccidentConsequenceAssessmentsatNuclearPowerPlants,"andthepaper,"NuclearPowerPlantControlRoom VentilationSystemDesignforMeetingGeneralCriterion19"(Refs.6,7,22,and28).

19TheARCON96computercodecontainsprocessingoptionsthatmayyield c/Qvaluesthatarenotsufficientlyconservativeforuseinaccidentconsequenceassessmentsormaybeincompatiblewithreleasepointandventilationintakeconfigurationsatparticularsites.Theapplicabilityoftheseoptionsandassociatedinputparametersshouldbeevaluatedonacase-by-casebasis.

TheassumptionsmadeintheexamplesintheARCON96documentationareillustrativeonlyanddonotimplyNRCstaff acceptanceofthemethodsordatausedintheexample.

1.183-23References22and28shouldbeusediftheFSAR

c/Qvaluesaretoberevisedorifvaluesaretobedeterminedfornewreleasepointsorreceptordistances.FumigationshouldbeconsideredwhereapplicablefortheEABandLPZ.FortheEAB,theassumedfumigationperiod shouldbetimedtobeincludedintheworst2-hourexposureperiod.TheNRCcomputercode PAVAN(Ref.29)implementsRegulatoryGuide1.145(Ref.28)anditsuseisacceptabletothe NRCstaff.ThemethodologyoftheNRCcomputercodeARCON96

19(Ref.26)isgenerallyacceptabletotheNRCstaffforuseindeterminingcontrolroom c/Qvalues.Meteorologicaldatacollectedinaccordancewiththesite-specificmeteorologicalmeasurementsprogramdescribedinthefacilityFSARshouldbeusedingeneratingaccident c/Qvalues.AdditionalguidanceisprovidedinRegulatoryGuide1.23,"OnsiteMeteorologicalPrograms"(Ref.30).Allchangesin

ÿ/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 consequenceanalysesatoperatingpowerreactors.

1.183-24

1.183-25REFERENCES{SeetheinsidefrontcoverofthisguideforinformationonobtainingNRCdocuments.}1.J.J.DiNunnoetal.,"CalculationofDistanceFactorsforPowerandTestReactorSites,"USAECTID-14844,U.S.AtomicEnergyCommission(nowUSNRC),1962.2.USNRC,"ClarificationofTMIActionPlanRequirements,"NUREG-0737,November

1980.3.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,December1978.

1.183-2613.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,March1978.

1.183-2726.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,February1972.

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.SOURCETERMASSUMPTIONS

1.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.ASSUMPTIONSONTRANSPORTINPRIMARYCONTAINMENT

3.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.ASSUMPTIONSONDUALCONTAINMENTS

4.Forfacilitieswithdualcontainmentsystems,theacceptableassumptionsrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromthesecondarycontainmentor enclosurebuildingsareasfollows.

4.1Leakagefromtheprimarycontainmentshouldbeconsideredtobecollected,processedbyengineeredsafetyfeature(ESF)filters,ifany,andreleasedtotheenvironmentviathe secondarycontainmentexhaustsystemduringperiodsinwhichthesecondarycontainment hasanegativepressureasdefinedintechnicalspecifications.Creditforanelevatedrelease shouldbeassumedonlyifthepointofphysicalreleaseismorethantwoandone-halftimes theheightofanyadjacentstructure.

A-4 4.2Leakagefromtheprimarycontainmentisassumedtobereleaseddirectlytotheenvironmentasaground-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).ASSUMPTIONSONESFSYSTEMLEAKAGE

5.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:

FF hh h ff fg=-12Where: h f1istheenthalpyofliquidatsystemdesigntemperatureandpressure;h f2istheenthalpyofliquidatsaturationconditions(14.7psia,212ºF);andh fgistheheatofvaporizationat212ºF.

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).ASSUMPTIONSONMAINSTEAMISOLATIONVALVELEAKAGEINBWRS

6.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.ASSUMPTIONONCONTAINMENTPURGING

7.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,November

1980.A-9J.E.Cline,"MSIVLeakageIodineTransportAnalysis,"LetterReportdatedMarch26,1991.(ADAMSAccessionNumberML003683718)A-10USNRC,"SafetyEvaluationofGETopicalReport,NEDC-31858P(ProprietaryGEreport),Revision2,BWROGReportforIncreasingMSIVLeakageLimitsandEliminationofLeakageControlSystems

,September1993,"letterdatedMarch3,1999,ADAMSAccessionNumber9903110303.

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.

2Containment isolationdoesnotimplycontainmentintegrityasdefinedbytechnicalspecificationsfornon-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 ventilationflowtotheESFfiltrationsystem

1shouldbedeterminedandaccountedforintheradioactivityreleaseanalyses.

4.3TheradioactivityreleasefromthefuelpoolshouldbeassumedtobedrawnintotheESFfiltrationsystemwithoutmixingordilutioninthefuelbuilding.Ifmixingcanbe demonstrated,creditformixinganddilutionmaybeconsideredonacase-by-casebasis.

Thisevaluationshouldconsiderthemagnitudeofthebuildingvolumeandexhaustrate, thepotentialforbypasstotheenvironment,thelocationofexhaustplenumsrelativeto thesurfaceofthepool,recirculationventilationsystems,andinternalwallsandfloorsthat impedestreamflowbetweenthesurfaceofthepoolandtheexhaustplenums.5.FUELHANDLINGACCIDENTSWITHINCONTAINMENTForfuelhandlingaccidentspostulatedtooccurwithinthecontainment,thefollowingassumptionsareacceptabletotheNRCstaff.

5.1Ifthecontainmentisisolated

2duringfuelhandlingoperations,noradiologicalconsequencesneedtobeanalyzed.

5.2Ifthecontainmentisopenduringfuelhandlingoperations,butdesignedtoautomaticallyisolateintheeventofafuelhandlingaccident,thereleasedurationshouldbebasedon

3Thestaffwillgenerallyrequirethattechnicalspecificationsallowingsuchoperationsincludeadministrativecontrolstoclosetheairlock,hatch,oropenpenetrationswithin30minutes.Suchadminstrativecontrolswillgenerallyrequirethatadedicated individualbepresent,withnecessaryequipmentavailable,torestorecontainmentclosureshouldafuelhandlingaccidentoccur.Radiologicalanalysesshouldgenerallynotcreditthismanualisolation.B-3delaysinradiationdetectionandcompletionofcontainmentisolation.Ifitcanbeshownthatcontainmentisolationoccursbeforeradioactivityisreleasedtotheenvironment, 1 noradiologicalconsequencesneedtobeanalyzed.

5.3Ifthecontainmentisopenduringfuelhandlingoperations(e.g.,personnelairlockorequipmenthatchisopen), 3theradioactivematerialthatescapesfromthereactorcavitypooltothecontainmentisreleasedtotheenvironmentovera2-hourtimeperiod.

5.4AreductionintheamountofradioactivematerialreleasedfromthecontainmentbyESFfiltersystemsmaybetakenintoaccountprovidedthatthesesystemsmeettheguidanceof RegulatoryGuide1.52andGenericLetter99-02(Refs.B-2andB-3).Delaysinradiation detection,actuationoftheESFfiltrationsystem,ordiversionofventilationflowtothe ESFfiltrationsystemshouldbedeterminedandaccountedforintheradioactivityrelease analyses.1 5.5Creditfordilutionormixingoftheactivityreleasedfromthereactorcavitybynaturalorforcedconvectioninsidethecontainmentmaybeconsideredonacase-by-casebasis.

Suchcreditisgenerallylimitedto50%ofthecontainmentfreevolume.Thisevaluation shouldconsiderthemagnitudeofthecontainmentvolumeandexhaustrate,thepotential forbypasstotheenvironment,thelocationofexhaustplenumsrelativetothesurfaceof thereactorcavity,recirculationventilationsystems,andinternalwallsandfloorsthat impedestreamflowbetweenthesurfaceofthereactorcavityandtheexhaustplenums.

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,June3,1999.

1Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,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. Ifnoorminimal

1fueldamageispostulatedforthelimitingevent,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%perday

2foraperiodof24hours,atwhichtimetheleakageisassumedtoterminate.Nocreditshouldbeassumedfordilutionorholdupwithinthe C-2turbinebuilding.Radioactivedecayduringholdupintheturbineandcondensermaybe assumed.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%organic.

1Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingdoseequivalentI-131(DEI-

131),onlytheradioiodineassociatedwithnormaloperationsoriodinespikesshouldbeincluded.Activityfromprojectedfuel damageshouldnotbeincluded.D-1AppendixDASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFABWRMAINSTEAMLINEBREAKACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofamainsteamlineaccidentatBWRlightwaterreactors.These assumptionssupplementtheguidanceprovidedinthemainbodyofthisguid

e. SOURCETERM

1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisguide.Thereleasefrom thebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimateofthe numberoffuelrodsbreached.

2. Ifnoorminimal

1fueldamageispostulatedforthelimitingevent,thereleasedactivityshouldbethemaximumcoolantactivityallowedbytechnicalspecification.Theiodine concentrationintheprimarycoolantisassumedtocorrespondtothefollowingtwocasesinthe nuclearsteamsupplysystemvendor'sstandardtechnicalspecifications.

2.1Theconcentrationthatisthemaximumvalue(typically4.0µCi/gmDEI-131)permittedandcorrespondstotheconditionsofanassumedpre-accidentspike,and

2.1Theconcentrationthatisthemaximumequilibriumvalue(typically0.2µCi/gmDEI-131)permittedforcontinuedfullpoweroperation.

3.Theactivityreleasedfromthefuelshouldbeassumedtomixinstantaneouslyandhomogeneouslyinthereactorcoolant.Noblegasesshouldbeassumedtoenterthesteamphase instantaneously.TRANSPORT 4.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows.

4.1Themainsteamlineisolationvalves(MSIV)shouldbeassumedtocloseinthemaximumtimeallowedbytechnicalspecifications.

4.2Thetotalmassofcoolantreleasedshouldbeassumedtobethatamountinthesteamlineandconnectinglinesatthetimeofthebreakplustheamountthatpassesthroughthe valvespriortoclosure.

D-2 4.3Alltheradioactivityinthereleasedcoolantshouldbeassumedtobereleasedtotheatmosphereinstantaneouslyasaground-levelrelease.Nocreditshouldbeassumedfor plateout,holdup,ordilutionwithinfacilitybuildings.

4.4Theiodinespeciesreleasedfromthemainsteamlineshouldbeassumedtobe95%CsIasanaerosol,4.85%elemental,and0.15%organic.

1Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity,"foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.

2Theactivityassumedintheanalysisshouldbebasedontheactivityassociatedwiththeprojectedfueldamageorthemaximumtechnicalspecificationvalues,whichevermaximizestheradiologicalconsequences.IndeterminingdoseequivalentI-131(DEI-

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

Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguide.

1SOURCETERMS

1.AssumptionsacceptabletotheNRCstaffregardingcoreinventoryandthereleaseofradionuclidesfromthefuelareprovidedinRegulatoryPosition3ofthisregulatoryguide.The releasefromthebreachedfuelisbasedonRegulatoryPosition3.2ofthisguideandtheestimate ofthenumberoffuelrodsbreached.Thefueldamageestimateshouldassumethatthehighest worthcontrolrodisstuckatitsfullywithdrawnposition.

2. Ifnoorminimal

2fueldamageispostulatedforthelimitingevent,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.Theactivityreleasedfromthefuelshouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughtheprimarycoolant.

3Inthisappendix,rupturedreferstothestateofthesteamgeneratorinwhichprimary-to-secondaryleakageratehasincreasedtoavaluegreaterthantechnicalspecifications.FaultedreferstothestateofthesteamgeneratorinwhichthesecondarysidehasbeendepressurizedbyaMSLBsuchthatprotectivesystemresponse(mainsteamlineisolation,reactortrip,safetyinjection,etc.)hasoccurred.PartitioningCoefficientisdefinedas:

PCmassofIperunitmassofliquidmassofIperunitmassofgas

=2 2E-2 4.Thechemicalformofradioiodinereleasedfromthefuelshouldbeassumedtobe95%cesiumiodide(CsI),4.85percentelementaliodine,and0.15percentorganiciodide.Iodine releasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elemental and3%organic.Thesefractionsapplytoiodinereleasedasaresultoffueldamageandtoiodine releasedduringnormaloperations,includingiodinespiking.TRANSPORT 3 5.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)shouldbeconsistentwiththebasisoftheparameterbeingconverte

d. TheARC

leakratecorrelationsaregenerallybasedonthecollectionofcooledliquid.Surveillance testsandfacilityinstrumentationusedtoshowcompliancewithleakratetechnical specificationsaretypicallybasedoncooledliquid.Inmostcases,thedensityshouldbe assumedtobe1.0gm/cc(62.4lbm/ft

3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureofthe leakageislessthan100°C(212°F).Thereleaseofradioactivityfromunaffectedsteam generatorsshouldbeassumedtocontinueuntilshutdowncoolingisinoperationand releasesfromthesteamgeneratorshavebeenterminated.

5.4Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.

5.5Thetransportmodeldescribedinthissectionshouldbeutilizedforiodineandparticulatereleasesfromthesteamgenerators.ThismodelisshowninFigureE-1andsummarized

below:

E-3SteamSpaceBulkWaterPrimaryLeakageScrubbingPartitioningReleaseFigureE-1TransportModel

5.5.1Aportionoftheprimary-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.

5.5.3Theleakagethatdoesnotimmediatelyflashisassumedtomixwiththebulk water.5.5.4Theradioactivityinthebulkwaterisassumedtobecomevaporataratethatisthefunctionofthesteamingrateandthepartitioncoefficient.Apartitioncoefficient foriodineof100maybeassumed.Theretentionofparticulateradionuclidesin thesteamgeneratorsislimitedbythemoisturecarryoverfromthesteam generators.

5.6Operatingexperienceandanalyseshaveshownthatforsomesteamgeneratordesigns,tubeuncoverymayoccurforashortperiodfollowinganyreactortrip(Ref.E-3).The potentialimpactoftubeuncoveryonthetransportmodelparameters(e.g.,flashfraction, scrubbingcredit)needstobeconsidered.Theimpactofemergencyoperatingprocedure restorationstrategiesonsteamgeneratorwaterlevelsshouldbeevaluated.

E-4AppendixEREFERENCESE-1USNRC,"SteamGeneratorTubeIntegrity,"DraftRegulatoryGuideDG-1074,December

1998.E-2.USNRC,"IodineBehaviorinaPWRCoolingSystemFollowingaPostulatedSteamGeneratorTubeRuptureAccident,"NUREG-0409,May1985.E-3USNRC,"SteamGeneratorTubeRuptureAnalysisDeficiency,"InformationNotice88-31,May25,1988.

1Facilitieslicensedwith,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. Ifnoorminimal

2fueldamageispostulatedforthelimitingevent,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,shouldbeassumedtobereleasedinstantaneouslyandhomogeneouslythroughtheprimarycoolant.

3Inthisappendix,rupturedreferstothestateofthesteamgeneratorinwhichprimary-to-secondaryleakageratehasincreasedtoavaluegreaterthantechnicalspecifications.F-2 4.Iodinereleasesfromthesteamgeneratorstotheenvironmentshouldbeassumedtobe97%elementaland3%organic.TRANSPORT 3 5.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/ft

3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureoftheleakageislessthan100

°C(212°F).Thereleaseofradioactivityfromtheunaffectedsteamgeneratorsshouldbeassumedtocontinueuntilshutdowncoolingisinoperationandreleasesfromthesteamgeneratorshavebeenterminated.

5.4Thereleaseoffissionproductsfromthesecondarysystemshouldbeevaluatedwiththeassumptionofacoincidentlossofoffsitepower.

5.5Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.

5.6ThetransportmodeldescribedinRegulatoryPositions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulates.

1Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity"(USNRC,December1998),foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.G-1AppendixGASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRLOCKEDROTORACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofalockedrotoraccidentatPWRlightwaterreactors.

1 Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguid

e. SOURCETERM

1.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,includingiodinespikin

g. RELEASETRANSPORT

5.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialtotheenvironmentareasfollows.

5.1Theprimary-to-secondaryleakrateinthesteamgeneratorsshouldbeassumedtobetheleak-rate-limitingconditionforoperationspecifiedinthetechnicalspecifications.The leakageshouldbeapportionedbetweenthesteamgeneratorsinsuchamannerthatthe calculateddoseismaximized.

5.2Thedensityusedinconvertingvolumetricleakrates(e.g.,gpm)tomassleakrates(e.g.,lbm/hr)shouldbeconsistentwiththebasisofsurveillancetestsusedtoshowcompliance withleakratetechnicalspecifications.Thesetestsaretypicallybasedoncoolliquid.

G-2Facilityinstrumentationusedtodetermineleakageistypicallylocatedonlinescontainingcoolliquids.Inmostcases,thedensityshouldbeassumedtobe1.0gm/cc(62.4lbm/ft

3).5.3Theprimary-to-secondaryleakageshouldbeassumedtocontinueuntiltheprimarysystempressureislessthanthesecondarysystempressure,oruntilthetemperatureoftheleakageislessthan100

°C(212°F).Thereleaseofradioactivityshouldbeassumedtocontinueuntilshutdowncoolingisinoperationandreleasesfromthesteamgeneratorshavebeenterminated.

5.4Thereleaseoffissionproductsfromthesecondarysystemshouldbeevaluatedwiththeassumptionofacoincidentlossofoffsitepower.

5.5Allnoblegasradionuclidesreleasedfromtheprimarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.

5.6Thetransportmodeldescribedinassumptions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulates.

1Facilitieslicensedwith,orapplyingfor,alternativerepaircriteria(ARC)shouldusethissectioninconjunctionwiththeguidancethatisbeingdevelopedinDraftRegulatoryGuideDG-1074,"SteamGeneratorTubeIntegrity"(USNRC,December1998),foracceptableassumptionsandmethodologiesforperformingradiologicalanalyses.H-1AppendixHASSUMPTIONSFOREVALUATINGTHERADIOLOGICALCONSEQUENCESOFAPWRRODEJECTIONACCIDENTThisappendixprovidesassumptionsacceptabletotheNRCstaffforevaluatingtheradiologicalconsequencesofarodejectionaccidentatPWRlightwaterreactors.

1 Theseassumptionssupplementtheguidanceprovidedinthemainbodyofthisguid

e. SOURCETERM

1.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%organic.

H-2TRANSPORTFROMCONTAINMENT

6.AssumptionsacceptabletotheNRCstaffrelatedtothetransport,reduction,andreleaseofradioactivematerialinandfromthecontainmentareasfollows.

6.1Areductionintheamountofradioactivematerialavailableforleakagefromthecontainmentthatisduetonaturaldeposition,containmentsprays,recirculatingfilter systems,dualcontainments,orotherengineeredsafetyfeaturesmaybetakeninto account.RefertoAppendixAtothisguideforguidanceonacceptablemethodsand assumptionsforevaluatingthesemechanisms.

6.2Thecontainmentshouldbeassumedtoleakattheleakrateincorporatedinthetechnicalspecificationsatpeakaccidentpressureforthefirst24hours,andat50%ofthisleakrate fortheremainingdurationoftheaccident.Peakaccidentpressureisthemaximum pressuredefinedinthetechnicalspecificationsforcontainmentleaktesting.Leakage fromsubatmosphericcontainmentsisassumedtobeterminatedwhenthecontainmentis broughttoasubatmosphericconditionasdefinedintechnicalspecifications.TRANSPORTFROMSECONDARYSYSTEM

7.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/ft

3).7.3Allnoblegasradionuclidesreleasedtothesecondarysystemareassumedtobereleasedtotheenvironmentwithoutreductionormitigation.

7.4Thetransportmodeldescribedinassumptions5.5and5.6ofAppendixEshouldbeutilizedforiodineandparticulates.

I-1AppendixIASSUMPTIONSFOREVALUATINGRADIATIONDOSESFOREQUIPMENTQUALIFICATIONThisappendixaddressesassumptionsassociatedwithequipmentqualificationthatareacceptabletotheNRCstaffforperformingradiologicalassessments.AsstatedinRegulatory Position6ofthisguide,thisappendixsupersedesRegulatoryPositions2.c.(1)and2.c.(2)and AppendixDofRevision1ofRegulatoryGuide1.89,"EnvironmentalQualificationofCertain ElectricEquipmentImportanttoSafetyforNuclearPowerPlants"(USNRC,June1984),for operatingreactorsthathaveamendedtheirlicensingbasistouseanalternativesourceterm.

Exceptasstatedinthisappendix,otherassumptions,methods,andprovisionsofRevision1of RegulatoryGuide1.89remaineffectiv

e. BASICASSUMPTIONS

1.Gammaandbetadosesanddoseratesshouldbedeterminedforthreetypesofradioactivesourcedistributions:(1)activitysuspendedinthecontainmentatmosphere,(2)activityplatedout oncontainmentsurfaces,and(3)activitymixedinthecontainmentsumpwater.Agivenpieceof equipmentmayreceiveadosecontributionfromanyorallofthesesources.Theamountofdose contributedbyeachofthesesourcesisdeterminedbythelocationoftheequipment,thetime- dependentandlocation-dependentdistributionofthesource,andtheeffectsofshieldin

g. 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.FISSIONPRODUCTCONCENTRATIONS

3.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).DOSEMODELFORCONTAINMENTATMOSPHERE

5.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.DOSEMODELFORCONTAINMENTSUMPWATERSOURCES

8.Withtheexceptionofnoblegases,alltheactivityreleasedfromthefuelshouldbeassumedtobetransportedtothecontainmentsumpasitisreleased.Thisactivityshouldbe assumedtomixinstantaneouslyanduniformlywithotherliquidsthatdraintothesump.This I-3transportcanalsobemodeledmechanisticallyasthetime-dependentwashoutofairborneaerosolsbytheactionofcontainmentsprays.Radionuclidesthatdonotbecomeairborneon releasefromthereactorcoolantsystem,e.g.,theyareentrainedinnon-flashedreactorcoolant, shouldbeassumedtobeinstantaneouslytransportedtothesumpandbeuniformlydistributedin thesumpwater.

9.Thegammaandbetadoseratesandtheintegrateddosesshouldbecalculatedforapointlocatedonthesurfaceofthewateratthecenterlineofthelargepoolofsumpwater.Theeffects ofbuildupshouldbeconsidered.Moredetailedmodelingwithshieldinganalysiscodesmaybe

performed.DOSEMODELFOREQUIPMENTLOCATEDOUTSIDECONTAINMENT

10.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 thefilterandtheEQequipmentmaybeconsidered.

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