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CompositionbyRosemaryGuenther,PAPool,andGroup1S-10

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UC-41Issued:January1986

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I

CONTENTS

ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

I. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A. NeedfortheGuide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B. StatusoftheGuide . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .C. FeaturesofDOENuclearFacilities . . . . . . . . . . . . . . . . . . . . . . . . . .

II. SCOPEANDINTENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

III. CRITERIAANDDEFINITIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .A. DOEOrders . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1. Order5480.lA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2. Order6430.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3. Order5500.3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4. Order5481.lA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

B. Definitions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1. DesignBasisAccident(DBA) . . . . . . . . . . . . . . . . . . . . . . . . . . .2. OffsitePerson. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3. Population Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4. PopulationCenterDistance . . . . . . . . . . . . . . . . . . . . . . . . . . . .5. EffectiveDoseEquivalent. . . . . . . . . . . . . . . . . . , . . . . . . . . . .6. RiskAnalysisandRiskAssessment. . . . . . . . . . . . . . . . . . . . . . . .

IV. DESCRIPTIONOFDESIGNBASISACCIDENTS. . . . . . . . . . . . . . . . . . . . .A. OperationalAccidents. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1. Explosions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2. Fires . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3. CriticalityAccidents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .4. LeakstotheAtmosphere. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5. LeakstotheAquaticEnvironment . . . . . . . . . . . . . . . . . . . . . . . .

B. NaturalPhenomenaEvents. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1. Earthquakes. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

a. SiteSelection. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .b. StructuralAdequacy. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

(1) DescriptionoftheDBE. . . . . . . . . . . . . . . . . . . . . . . . . .(2) StaticResponseMethod . . . . . . . . . . . . . . . . . . . . . . . . .(3) DynamicAnalysisMethod. . . . . . . . . . . . . . . . . . . . . . . .(4) DeterminationofDamping . . . . . . . . . . . . . . . . . . . . . . .(5) StrengthAnalysis(CombinedResponse) . . . . . . . . . . . . . . . .

c. ComponentAdequacy. . . . . . . . . . . . . . . . . . . . . . . . . . . . .d. DamageEstimationandReleaseFract.ion. . . . . . . . . . . . . . . . . .

2. TomadoesandExtremeWinds . . . . . . . . . . . . . . , . . . . . . . . . . .a. SiteSelection. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .b. StructuralAdequacy. . . . . . . . . . . . . . . . . . , . . . . . . . . . . .c. ComponentAdequacy. . . . . . . . . . . . . . . . . . . . . . . . . . . . .d. DamageEstimationandReleaseFraction. . . . . . . . . . . . . . . . . .e. DispersionAssumptions . . . . . . . . . . . . . . . . . . . . . . . . . . .

‘3. OtherNaturalPhenomena. . . . . . . . . . . . . . . . . . . . . . . . . . . . .C. AccidentswithExtemalOrigins . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1. Pipelines,Tankers, Barges,or Rail Cars CarryingHazardousMaterials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. AircraftandAirports. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3. OtherNuclearFacilitiesorReactors . . . . . . . . . . . . . . . . . . . . . . .4. LargeDams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5. ExplosiveorToxicMaterialFacilities . . . . . , . . . . . . . . . . . . . . . .

D. AccidentswithHigherProbabi.1.ity.. . . . . . . . . . . . . . . . . . . . . . . . . .

112 :A-2

2f<

.4 I

444

~

555556666

67789

111111121212121213131313141414141415151515

151616161616

1

.

iv

V. EVALUATIONSOF ACCIDENT CONSEQUENCES . . . . . . . . . . . . . . . . . . .A.

B.c.D.E.

F.

G.

SourceTerms. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1. Radionuclides. . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . . . .2. RadionuclideQuant.ities. . . . . . . . . . . . . . . . . , . . . . . . . . . . . .

ReleaseFractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .ReductionandRemovalFactors. . . . , . . . . . . . . . . . . . . . . . . . . . . .ReleaseDuration . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .MeteorologicalAnalysisandDispersion. . . . . . . . . . . . . . . . . . . . . . . .1.2.3.

4.

5.6.

Gauss;anDispe&ionModel_. . . . . . . . . . . . . . . . . . . . . . . . . . . .DispersionParameters/StabilityClassifications. . . . . . . . . . . . . . . . .ReleaseEffects . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .a. Plume Rise. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .b. Stackand BuildingWake Effects. . . . . . . . . . . . . . . . . . . . . . .DispersionEffects. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .a. RadioactiveDecay. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .b. Inversion Lid. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .c. Fumigation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .d. Terrain . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .e. Dry Deposition. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Meteorology. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .Pomlation Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

RadiologicalDose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1. Inhalation Dose. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2. Direct Irradiation from Cloud Immersion . . . . . . . . . . . . . . . . . . . .3. IngestionDose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Other Consequencesto be Considered. . . . . . . . . . . . . . . . . . . . . . . . .1. EnvironmentalContamination . . . . . . . . . . . . . . . . . . . . . . . . . .2. PopulationDose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3. PublicHealth Effects. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

ACKNOWLEDGMENTS . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

REFERENCES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

APPENDICESA. GLOSSARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B. RADI.ATIONDOSE AND HEALTH EYFECTS. . . . . . . . . . . . . . . . . . .C. RISK ASSESSMENTMETHODS. . . . . . . . . . . . . . . . . . . . . . . . . . .D. DISPERSIONCALCULATIONMETHODS . . . . . . . . . . . . . . . . . . . .E. DOSE METHODS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .F. DOSE CODE COMPARISON. . . . . . . . . . . . . . . . . . . . . . . . . . . . .G. ENVIRONMENTALCONTAMINATION CONCERNS . . . . . . . . . . . . .

ATTACHMENTA. COMMENT SHEET . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

.

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.1.

11.

111.

IV.

TABLES

SUMMARYOF NONREA~OR NUCLEAR FACILITYTYPES . . . . . . . . . . . .

SITING GUIDELINE DOSES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

EXPLOSIONACCIDENT CAUSES . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

FIRE ACCIDENT CATEGORIES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1617171819212324242424242424252525252525252525262626262727

28

28

34394346525861

67

3

5

8

9

v

V. CRITICALITYACCIDENT FISSIONYIELDS. . . . . . . . . . . . . . . . . . . . . . .

VL POTENTIAL RADIOLOGICALDOSEGUIDELINES FOR ACCIDENTEVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

VII. IMPORTANT RADIONUCLIDESFOR ORGAN DOSECALCULATIONS. . . . . .

VIII. SAFETYANALYSISPAIUM ETERSSUMMARY-RELEASE FRAC71’IONS. . . . .

IX. SAFETYANALYSISPARAMETERSSUMMARY-REDUCI’ION AND REMOVALFACTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

B-L LIFETIMERISK OF DYING OF A RADIATION-INDUCED CANCER . . . . . . . .

F-I. DISPERSIONAND DOSE CODES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

F-IL SUMMARYOF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

G-L DECONTAMINATION FACTORSFOR Pu02 . . . . . . . . . . . . . . . . . . . . . . .

G-II. POPULATION DENSITY AND LAND USE FRACTIONS FOR AGRICULTURAL,TOWN, AND CENTRAL CITY AREAS. . . . . . . . . . . . . . . . . . . . . . . . . . .

G-III. BUILDING AND DWELLING PARAMETER ESTIMATES. . . . . . . . . . . . . . .

G-IV. BUILDING TYPE AND LAND USE AREAFRACTION ESTIMATES. . . . . . . . .

G-V. DECONTAMINATION METHODS AND COSTS(1981DOLLARS)FORAGRICULTURAL, SUBURBAN,AND COMMERCIALAREASFOR PUOZ. . . . . .

G-VI. TOTAL DECONTAMINATION COST SUMMARYESTIMATES . . . . . . . . . . .

10

17

20

22

23

41

59

60

62

63

63

64

65

66

FIGURES

Figure 1.Earthquakeanalysissteps. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

Figure2. Accidentreleasesteps. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18

Figure3.Accidentconsequencesteps . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

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vi

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A GUIDE TO IL4DIOLOGICAL ACCIDENT CONSIDEl&iTIONSFOR SITING AND DESIGN OF

DOE NONREACTOR NUCLEAR FACILITIES

by

J. C. Elder, J. M. Graf, J. M. Dewa@T. E. Buh~W. J. Wenze&L. J. Walkex,and A. K. Stoker

ABSTRACT

This Guide was prepared to providethe experiencedsafety analyst with accidentanalysis guidancein greater detail than is possible in Department of Energy (DOE)Orders. The Guideaddressesanalysisofpostulatedseriousaccidentsconsideredin thesiting and selectionof major design features of DOE nuclear facilities.Its scopehasbeen limited to radiologicalaccidents at nonreactor nuclear facilities. The analysisstepsaddressedin the GuideIead to evaluationof radiologicaldoseto exposedpersonsforcomparisonwithsitingguidelinedoses.Other possibleconsequencesconsideredareenvironmentalcontamination,populationdose, and public health effects. Choices ofmodels and parameters leading to estimation of source terms, release fractions,reductionand removal factors, dispersion and dose factors are discnssed. AIthoughrequirementsfor risk analysis have not been established, risk estimates are findingincreaseduse in sitingof majornuclearfacilitiesand are discussedin the Guide.

I. INTRODUCTION

A. Need for the Guide

A DOE guide in the area of postulated radiologicalaccident analysis at nonreactor nuclear facilities wasprovidedto supplymore detail than that practicalin theDOE Orders. Further, collectionof related informationin a guideof this typewasconsideredusefulfor conduct-ingradiologicalaccidentanalysiswithin the DOE com-

-- plex.Major requirements for the siting and design of nu-

clear facilitiesas formulated by the Nuclear Regulatory. Commission(NRC) and DOE are found in the Code of

Federal Regulations(CFR, Title 10) and DOE Ordersrespectivelyand are stated in terms of radiation dose (oreffectivedose equivalent) calculated at a specified lo-cation. Althoughthe sitingcriteria dosesmay be unam-biguous, the many models, parameters, and assump-

tions necessaryfor the intermediate calculationalstepsbetweenthe postulated accident and the resultant doseare not specified.The NRC addresses this situation byissuingsupplementaryguidancesuch as technicalinfor-mation documents,safetyguides,and regulatoryguides.This supplementaryinformation is availableto and hasbeen appropriately used by DOE safety analysts;how-ever, it is specificto the types of facilitieslicensed byNRC, particularlylight-waterreactors(LWRS).That theDOE should apply reactor-basedcriteria to nonreactorfacilitieshas been implied in DOE Orders. However,applying criteria intended for LWR power plants toDOE facilities that may be “first of a kind,” locatedremotelyon a DOE reservation, and already subject tostrictregulatorycontrol is often not appropriate.Recog-nition that a numerical limit shouldbe accompaniedbyguidancefor its applicationled to the preparation of thisGuide. Its scope is limited to accidental release ofradioactivematerial from nonreactor nuclear facilities.

1

B. Status of the Guide

This Guide shouldbe regardedonlyas a guide,not asa regulation or standard or as a statement of DOEpolicy. DOE Orders remain the primary source of re-quirements related to nuclearfacilitysiting,design,andsafetyanalysis.The status of the Guide as a substantialdocumentusefulto the safetyanalysthasbeen enhancedby an effortto obtain carefil peer review.Followingthesuggestionscontained in this Guide should lead tocompliancewith applicableDOE Orders.

The extent to whichthe Guide willreceivecontinuingreviewand revision has not been determined.This willdepend on its usefulness,the extent of its utilization byfield officesand contractors,and the feedbackreceivedduringthe firstyearor two after its issue.

Acomment sheethas been includedas AttachmentAto encouragethe user to provide the authors with inputas the need arises.

C. FeaturesofDOE NuclearFacilities

DOE nonreactor nuclear facilities house a broadvariety of processes,a number of which contain radio-logical hazards with potential for onsite and offsiteconsequencesin the event of a major accident. Thesepotentialhazardsare generallyof lowermagnitudethanthose associatedwith nuclear reactor facilitiesbecauseofa lowerinventoryof radionuclidesand lowerlevelsofdispersiveenergy.However,carefulanalysisofpotentialaccidents in nuclear facilitiesis required to assure thatthe combination of proper siting and design of safetyfeatureswouldprovide a highdegreeof safetyfor mem-bers of the public.This is accomplishedby consideringsiting criteria and safety features described in latersections.

Nonreactor nuclear facilities are defined in DOEOrder 5480.lA, Chapter V, “Safety of Nuclear Facili-ties,”as

NuclearFacility.A facilitywhoseoperationsinvolve radioactive materials in such formand quantity that a significantnuclear haz-ard potentially exists to the employeesandthe general public. Included are facilitiesthat (1) produce, process, or store radioac-tive liquid or solid waste, fissionablematerials, or tritium; (2) conduct separa-tions operations; (3) conduct irradiatedmaterialsinspection,fiel fabrication,decon-tamination, or recoveryoperations;(4) cOn-duct fuel enrichment operations. Incidental

use of radioactive materials in a facilityoperation (e.g., check sources, radioactivesources, and x-ray machines) does notnecessarilyrequirethe facilityto be includedin this definition. -a

Nuclear facilitiesare further categorizedin DOE Order -6430.1, Chapter IV, as either critical or noncritical ‘f-acilities (DOE 1983A).Critical facilitiesare those for “radioactive material handling, processing, or storage, - Iand other facilities having vital importance to DOEprograms or high dollar value [such as plutoniumprocessing, tritium processing, weapon assembly(HE/Pu), and certain storage facilities].Noncritical fa-cilities are other facilities that meet the definition ofnuclearfacilities(above).

The major categories of DOE nonreactor nuclearfacilitiesand a summary of processes,prominent radio-nuclides,dispersiveener~ potential,and accidenttypesmost likely to be the design basis accident (DBA) arepresentedin Table I. The entriesin Table I are discussedin greater detail in the Standards and Criteria Guide(Brynda1981).There maybe other nuclearfacilitytypeswhichrequireanalysisof a DBA.

II. SCOPE AND INTENT

The Guide focuses on the implementation of DOEOrder 5480.IA, Chapter V, DOE Order 5500.3, andDOE Order 6430.1,“General DesignCriteria Manual.”It also has incidental application to DOE Order5481.1A, “Safety Analysis and Review System.” Itsmost direct application is to DOE Order 6430.1,whichcontainsguidanceon sitingand major designfeaturesofnuclear facilities. Order 6430.1 discusses general re-quirements, largely leaving the choice of analysismethod and parameters to the analyst. The largenum-ber ofanalysismethodsavailablehave resultedin varia-tions among analysts seekingto answer the basic ques-tions:

Does the proposed site meet the siting guidelinedoses?Is the proposed site more suitable than alternativesites, based on consideration of other potential -consequencesof an accident, such as population -dose, environmental contamination, or public -health effects? --

Can emergencyplanning requirements be met at -the proposedsite?Can an existingfacilitysafelyhousea new process? -

.

TABLEI. SUMMARYOFNONREACTORNUCLEARFACILITYTYPES

Relative Dispersion PrincipalFacilityType Operations Radionuclides SourceTerm Potential DBA

PuorEnriched-UP

P

Criticality

TritiumProcessing

RM eoe

Criticality

L

i

u

Criticality,

A C

r

m

Lc 2

F

ECorrosivegas,highpressure

~6 d-e h

I

3

-“

.

.

The Guide has been compiled to aid experiencedanaIystsin app[yinganalysis techniquesconsistentlytoallaccidentswith major potential for radiologicalconse-quences.It is not intendedas a tutorialdocumentor as aguide to writing safety analysis reports (SARS).TheGuide should be usefi.danywhere postulated accidentanalysis is required, that is, for SARS,design analysis,environmental documentation, emergency planning,etc. Its primary utility willprobablybe as a tool for theanalystand will contain useful information and qualityreferences. Areas either vague or not covered by ex-perimentaldata are identified,and in some cases,speci-ficassumptionsare suggested.

Coveringall aspectsof radiologicalaccident analysiswas not considered possible or practical. The analystshould not assume all areas of accident analysis havebeen covered in the Guide, although an attempt hasbeen made to dealwith all important issues.

111.CRITERIA AND DEFINITIONS

A. DOE Orders

1. Environmental Protectio~ Safety, and HealthProtection Program for DOE Operations (OrderS480.lA). DOE Order 5480.lA, Chapter I (DOE1981A),“EnvironmentalProtection,Safety,and HealthProtection Standards,” provides as a recommendedstandard for nuclear facility safety “appropriate por-tions of Reactor Site Criteria” (10 CFR 100, revised,CFR 1962). DOE Order 5480.lA, Chapter V (DOE1981B),“SafetyofNuclearFacilities,”establishessafetyproceduresand requirementsfor nuclearfacilities(reac-tors and accelerators are exceptions in Chapter V) toassure

. . . that nuclear facilitiesare sited,designed,constructed, modified, operated, main-tained, and decommissionedin accordancewith generally uniform standards, guides,and codes that are consistent with thoseapplied to comparable licensed nuclear fa-cilities.

Although the 10 CFR 100 site criteria were issuedspecificallyfor siting stationary pressurized-waterandboiling-waterreactors,theyhave representedforover 20yr the only authoritative sitingguidanceand have beenapplied to nonreactor facilitiesby NRC and DOE [orpreviouslyby the AtomicEnergyCommission(AEC)orEnergy Research and Development Administration

(ERDA)].More recent guidancehas been proposed foruse within DOE in DOE Order 6430.1,as discussedinthe followingsections.

DOE Order 5480.1Aalso requiresthat adequate con-6siderationbe givento environmentalprotection,safety, ~

and health protection matters throughout the life of a -nuclear facility,includingits siting. Its operation must -not create undue environmental protection, safety, or --“-

health protection risks. Evaluation methods applied to -accident effects in these areas are addressed in theGuide.

2. GeneralDesignCriteria (Order 6430.1). The DOEimplements radiological accident guidance for sitingand major designfeaturesin DOE Order 6430.1,“Gen-eral DesignCriteria Manual” (DOE 1983A).DOE pol-icy stated in Order 6430.1requires that “DOE facilitiesbe designedand constructed to be reasonableand ade-quate for their intended purpose and consistent withhealth, safety, and environmental protection require-ments.”

Dose guidelinesproposed for inclusion in Chapter Iof DOE Order 6430.1 to limit a one-time accidentaldose from a major credible accident are as shown inTable II. The proposedguidelinedoses for wholebodyand thyroidare purposelyconsistentwith existingsitingcriteria in 10CFR 100.Bonesurfaceand lungdosesarebasedon ratios of ICRP weightingfactors(0.03forbonesurfaces,0.12 for lungs)to the thyroid weightingfactor(0.03) (ICRP 1977).This rationale was developed byNRC in the Clinch River Breeder Reactor site suit-ability decision (NRC 1977A).Determination of effec-tive doseequivalentis discussedin Section111.B.6.

The following caveat should accompany eachpublication of the guidelines to aid in keeping doseguidelinesin proper perspective(CFR 1962):

The use of these guideline doses is not in-tended to imply that these doses constituteacceptablelimits for emergencydosesto thepublic under accident conditions. Rather,these values are referencevalues to be usedin the evaluationof facilitysiteswith respectto potential accidents of exceedingly lowprobability of occurrence and low risk ofpublic exposure to radiation. They do notapply to facilityoperationsunder normal oremergency status, nor to emergency doseguidelinesthat might be appropriate for thegeneralpublicshouldan accidentoccur.

.

Definitions

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a

The radiologicalguidelines in DOE Order 6430.1,Chapter I, require considerationalso be given to onsitepersonnel when the site and major design features areselected.Considerationof the onsite person is stated inDOE Order 6430.1,Chapter I, to be “prudent measuresassociated with the radiological protection of onsitepersonnel and in conjunction with onsite emergencyplanning,as required through implementation of DOEOrder 5500.3.” The guidance on dose methodologyprovided herein is considered usefil in meeting therequirements of DOE 5500.3 related to emergencyradiologicalresponseplans.

3. Nuclear Facility Emergency Plaming, Prepared-ness, and ResponsePrograms (Order 5500.3).This or-der requiresemergencyactions to respond to the onsiteand offsite consequencesof a radiological emergencyand to assure protection of onsite personnel, publichealthand safety,and the environment.

4. Safety Analysis and Review System (Order5481.1A).Postulated accident considerations are dis-cussed in appropriate detail in DOE Order 5481.1Ainterms of safetyanalysis(DOE 1981C). Accident-relatedtopicsto receiveanalysisor be addressedare

● identificationof h~ds,. potentialaccidents,● probabilityof occurrence,s physical design features and administrative ~n-

trols to prevent or mitigatepotentialaccidents,and● predictedconsequences.

\VariousDOE fieldofficeshave prepared parallelordersto implement DOE Order 5481.1A.These orders are. generallymore speeificto the operational needs of con-tractor activities under a field officeand might not beapplicableacrossthe DOE complex.

The glossary of the Guide (Appendix A) containsdefinitions of many terms used in the Guide. Definit-ions needing further elaboration and discussion areincludedin this section.

Basis Accident (DBA). All credible acci-dents are evaluated for the purpose of establishingtheneed for certain designfatures in a nuclear facilityandapproving its siting.The DBA is that accident causingthe most severeconsequencesand is compared with theguidelinedoses.

Credibility of a potential accident is based on theamua.1frequencyat which the accident is expected tooccur.Accident-frequencydata to support a probabilityestimate may be lacking. In this case, a deterministicapproach similar to TID-14844 (AEC 1962) may beapplied.Assumptionsshouldcontain a suitablelevelofconservativism.Credibilitylimits in the rangeof 10–6to3 X 10-5occurrencesper yearhave been used within theDOE and elsewhere(Lucas 1981,Clemens 1982,ALO1982,ANSI 1976,NRC 1983).This Guide suggestsanapproximateannual frequencyof 10+ be used to estab-lish the credibilityof potential DBAs. The selectionof10+/yris basedprimarilyon a generalconsensusamongrisk analysts to consider a frequency of 10-5/yr as afrequency which should cause concern if the accidentconsequenceis high;conversely,a frequencylowerthan10-7/yr is considered so low as to be almost in-determinate or nonsensical.Therefore, any postulatedmajor aeeident which has an estimated annual fre-quencyapproaching10+/yr shouldbe consideredcredi-ble.

2. Offsite Person. The offsiteperson is a member ofthe exposed offsite population and is assumed to belocatedat the site boundary. His dose, which maybe awhole-bodydose,an individual-organdose, or an effec-tive dose equivalent, is compared with the sitingguide-linedosesproposedforDOE Order 6430.1(alsolistedinSection 111.A.2).The exposure reeeived by the offsiteperson should depend on unfavorable,site-speeificme-teorology &ta (methods are discussed in SectionV.E.5.).As a generalrule, the offsiteperson is assumedto be presentat the centerlineof the cloud fora period of2 h unlesscloud passageor evacuation within a shortertime is a reasonableassumption at the proposed site. Itmay be assumed that any person who might be con-sidered the offsite person is lawfully occupying thelocation and will consent to evacuation. The offsiteperson should be assumed to be the individual mostaffected by the released material. In most cases this

5

person is characterized by the ICRP reference man(ICRP 1974). A possible exception is a radioiodineexposurein which the dose to an infant’sthyroid mightresult in a highereffectivedoseequivalent.

3. PopulationDose.Populationdose is an estimate oftotal radiation dose received by members of a popula-tion group exposedto the radionuclidesreleasedby thepostulatedaccident.It can representa collectivewhole-body dose, collective dose to a specific organ, or acollective effective dose equivalent. Consideration ofpopulation dose may be appropriate under some cir-cumstances, such as the case where several sites arebeingcompared. Populationdosesmay showa decisiveadvantageof one siteover the others.

Estimation of population dose and potential publichealth effectsis addressed in SectionV.G.

4. PopulationCenter Distance.Population center dis-tance is defined as the minimum distance from theproposed nonreactor nuclear facility (structure, notboundary)to the nearestpopulationcenter. DOE Order6430.1 does not specify a population center distanceother than by indirect reference to 10 CFR 100. Apopulationcenter is defined in 10CFR 100as a denselypopulated center containing more than about 25000residents.To establisha minimum distanceto a popula-tion center,the 10CFR 100calculation(1.3timesa low-population-zoneouter-boundarydistance) is suggested,if the releaseis delayedover 2 h.

Somepopulationcenterboundariesare vague;that is,sprawlingsuburbsmay approach the proposedsitein anirregular pattern, making a single population centerdistanu difficultto choose.Politicalboundariesare notgenerally invoked as population boundaries withoutconsiderationof actual residences,includingthe possi-bility of fiture expansion bringingresidencescloser tothe facility.It will be necessa~ to evaluate each sitingcaseon its own particular featuresand makejudgmentsthat may not be alwaysconsistentamong analysts.

5. EffectiveDose Equivalent.Effectivedose equiva-lent isa specializeddosevalue relatingdosesreceivedbymultiple organs to a single whole-body dose for thepurposeof comparison with sitingguidelinedoses. It isbased on the ICRP 26 approach that risk of delayedmortalityshouldbe the same whetherthe wholebody isirradiated uniformly or whether several organs receivethe dose (ICRP 1977).The ICRP 26 approach replacesthe critical organ concept, which has been in use formany years.The effectivedose equivalent is the sum ofdoses to each organ receiving dose (including wholebody), after each organ dose is multiplied by an organweightingfactor. These weightingfactors are based on

somatichealth effectsdata derived from many sources.Some organ risk factors are still in question. Healtheffectsdata, the weightingfactors that may be derivedfromthem, and relatedtopicsare discussedin AppendixB.

6. Risk Analysis and Risk Assessment. Risk is thecombination of probability that a serious event willoccur (numlm of events per unit time) and the esti-mated consequencesof that event (commonly cancersor deaths or rem). The common practiceof callingriskthe product of probabilityand consequenceis not con-sidered sufficientlyinformative because both compo-nents are needed to filly describethe nature of the risk.Risk analysisis consideredthe specifictechnique, suchas fault or event tree analysis,which is used to performthe broader processof risk assessment.Since the DOEhasnot specificallyadopteda riskassessmentmethod ordefined an acceptable level of nslq treatment of thissubject in the Guide is limited to discussion of riskassessmentmethods presently in use by DOE contrac-tors: qualitative (informal) methods and probabilistic(formal)methods.Ongoingactivity in this area by DOEand its contractors is encouraged, along with activeinformation exchangeswith other safety analysts notpresentlyequipped to perform probabilisticrisk assess-ment. Sourcesof riskanalysisand nskassessment meth-odsare discussedfurther in AppendixC.

IV. DESCRIPTION OF DESIGN BASIS ACCI-DENTS

Requirements for identification and description ofthe DBAsare set forth in DOE Order 6430.1,Chapter I,and other chapterswhich are facilityspecific.Proposednuclear facilitiesmust be sited and designedto provideconfinement of radioactive materials under normaloperations and DBA conditions. The DBAs are thepostulated accidents and resulting conditions againstwhichthe structure,systems,and equipment must meettheir functionalgoals.The analysisof DBAsservestwomajorpurposes:to determinethe need duringthe designphase for engineered safety features (ESFS)and othercontrols,and to justify that the proposedfacilityinclud-ing the EiSFswilladequatelymeet sitingguidelinedosesin the event of the DBA.

Both internal and external initiating events must beconsidered.Unlessthe facilityis specificallydesignedtowithstand all credible external initiating events, somereleaseof radioactive material from these events mustbe assumed. This release should be ‘estimated andshownto be lower than that which could cause dose in

.

r-

.

6

,---

w

-.-I-

.

excess of the siting guideline doses in DOE Order6430.1,Chapter I (also in Guide Section 111.A.2).Thesteps involved in evaluation of amident consequencesare discussedin SectionV.

The processes in the nuclear facility should becarefullyreviewedto assure that all potential accidentsthat could qualifi as DBAs have been described andanalyzed,both for their probabilitiesand their conse-quences.The traditional deterministicmethod is acmp-tableas wellas some form of risk assessment(a consid-eration of both probabilityand consequence).An infor-mal, qualitative approach or a more formal, quan-titative approach may be used, if a comprehensive,systematic,well-reviewed and welldocumented analy-sis is performed. Two general methods of risk assess-ment (qualitative and quantitative analysis methods)are presently in use within the DOE complex and arediscussedin AppendixC.

The depth of analysis should be in some measureproportional to the level of risk at the facility underevaluation. Certainly, critical nuclear facilities (pluto-nium processing,tritium processing,weaponsassembly,and reactor fuel reprocessing)should be subjected to arigorousformal risk assessmentor a deterministicanal-ysis containing a suitable level of conservatism in theestimation of accident consequences.The followingas-sumptions are presented as examples, all or part ofwhichmightbe includedin formulation of DBAs:

● A worst-caserelease mechanism is assumed, con-sistent with credible but conservative selection ofphysicaland chemical parameters of the releasedmaterial.

. The m~mum amount of dispersiblemateri~ ~.lowed in the facility is assumed available for re-lease.

. Maximum release fraction based on the physicaland chemicalparameters is assumed.

. CreditforESFSand administrativecontrolssuchasevacuation is based on degraded performance un-lessthey are clearlyunaffectedby the accident.

. Unfavorable atmospheric or aquatic dispersionconditionsare assumed.

● Radiologicaldose calculationsare based on a 50-yrdose acmmulation time and selectionof breathingrate, particle size, and chemical volubilityclass,leading to doses which contain a suitable level ofconservatism.

A. O

The DBA could be an operational accident causedcaused by an internal event. Direct causes are usuallypoor design or procedures, operator errors, equipmentfailures, or inadequate technical development (un-

knowns)that lead to the accident. The major accidentcategoriesare explosion,fire, nuclearcriticality,leakstothe atmosphere,and leaks to the aquatic environment.Event histories have been prepared that can aid theanalyst in decidingwhat operational accident could bethe DBAfor the proposedfacility(for example,Perkins1980, 1981).Several other reports contain usefid de-scriptions of accidents and suggestedparameters ap-plicableto many operational DBAs (Selby 1975,Faust1977,Walker 1978,ANSI 1976,ANSI 1980,and ORNL1970).

The processesin the facilitymust bereviewed for potential energy release by explosion orother uncontrolled reaction that could release radioaetivematerialto theatmosphereor aquaticenvironment.The major causes of explosions involving radioactivematerials are listed in Table III. The list should not beconsidered all-inclusive but contains a summary ofmajor accident types either noted as actual events inevent histories or postulated events considered to becrediblethroughaccidentanalysis.

Experimental or explosion accident investigationdata available in the literature have been summarizedby Walker (1978);major sources of dispersal informa-tion are Selby (1975), Mishima (1966, 1970), andCastleman(1969).In the absenceof applicableinforma-tion, simpli&ng but consemative assumptions shouldbe made. For example, maximum airborne concentra-tion of respirableparticleswithin the space into whichsolid particles or liquid droplets are dispersed by anexplosionprobablywillnot exceed 100mg/m3after thefirst 10min (Selby1975).Releasefractionsare discussedtirt.her in SectionV.B.

For facilitiesother than a weaponsassembly(HE/Pu)facility,dispersalbeyond the facility structure may belimited by an assumption of structural integrityif blastbarriersare providedaround conceivableblast locationsand engineered safety fkatures are unaffected. In thiscase, the radioactive material would be dischargedthrough the ventilation system at its normal dischargerate and releasedat effectivestackheight.Otherwise,thereleaseshouldbe assumeda puff releaseat ground level.

HE/Pu assemblycellsdesignedto fullycontain an HEdetonation may release limited amounts of materialthroughfacilitypenetrations.An analysismethod ofthiscase may also be found in the Pantex EnvironmentalImpact Statement (EIS) (DOE 1983B)and supportingdocuments. A detonation accident at existing HE/Pufacilities(weaponsassemblycellsthat cannot iidly con-tain a detonation)shouldbe assumed to release 100%ofPu presentas an aerosoland 20%as a respirableaerosol.A suggestedmethod of accident analysis in which anelevated cloud of accident debris is dispersed may befound in the Pantex EIS and supporting documents(DOE 1983B).

7

TABLE III. EXPLOSION ACCIDENT CAUSES’

ChemicalProcessing

2.

3.

4.

5.

Hzexplosion—H2fromradiolysis,Na-H2reaction,fluoride-zirconiumreactionin dissolveror in a-.-1

reducingfurnace.w

Solventor red oil explosion—organiesin evaporators,concentrators,denigrators...

Hydrazoicacid explosion(hydrazine).

Ion exchangeresin—fwefollowedby explosion.

Unstable compounds—silver-nitrogen-halogencompounds, ammonium nitrate, mercury com-pounds.

WeaDonAssembly1. High-explosivesdetonation—uncasedHE/Pu or HE/U, duringassembly.

General1. Powderedmetals 7. Methane2. Hydrogen 8. Ozone3. Acetylene 9. Picricacid4. Volatileorganicliquids 10. Explosivegas mixtures5. Nitrates 11. Fuels,natural gas6. Peroxides

Selby

Frequencyof explosionsin chemicalprocessing,fielfabrication,and radioactive material processingplantscan be estimated from event histories. A typical ap-proach to estimating accident probability from failurerates and accident frequencies is described by Selby(1975,p. 92).Recent sourcesofdata have been collectedby Perkins(1980,1981).Sourceterms, releasefractions,and evaluation of accident consequencesare discussedfurther in SectionV.

2. Fires. The processes in the facility must be re-viewed for potential releaseof radioactive material byfiredamage.DOE Order 6430.1specifiesdesigncriteriafor nuclear facilities(Chapter X and in speefic facilitychapters).Fire resistancerequirementsare stated as a 2-h minimum fire barrier in the major walls, floors, andceilingsacting as a secondary or tertiary eontinementsystem.Four-hour barriers are required if the potentialcost of fire damage exceeds $75 million. It may beassumed that the fire does not breach the structure ifproposedfire protection (sprinldersand other systems)and restricted fire loading in the buildingare such thatthe fire should be extinguished within 2 h. A plume

release at effective release height may be assumed. Ifbreachof the buildingcannot be excluded,breach of theconfinement system followed by ground-level plume-model release (adjusted for thermal plume rise, if ap-propriate)shouldbe assumed.

Assumptionsregardingfire-causedreleaseof radioa&tive materialfromconfinementwillvaryamong nuclearfacilitiesand shouldbe substantiated for each case.Thefollowinggeneralstatements will usuallyapply to mostfacilities:

. All systems designed as cntieal systems operatenormallythroughoutthe event. 1-

● h cleaningsystems operate normally during the .fire if they receiveadequate protection from direct --fire damage of system components; HEPA tiltersoperate with accident-case efficiencies(stated in -Section V.C.) or with higher efficiencyif adequate -protection is provided by sprinkler systems, de- -misters,and prefilters.

. Any installed fire protection system functions asdesigned.

8

I

-.---

.

0

8

Amounts of flammable materials and radioactivematerials involved in the fire are the maximumamounts actually present during fidl-capacityoperation.Case-specificlocationof flammablema-terial, separate from radioactivematerials, may beused to mitigatethe consequencesof the tire;other-wise, estimates should include the more con-servativeassumptions.The amount of radioactive material in the sourceterm includesthe total amount of dispersiblemate-rial normally in processwithin an area surroundedby a 2-or 4-h firebarrier.

Selection of release fractions should be based onappropriate experimental or historical &ta where@sible. Since the amount of material released isdependenton the form and volatilityof the material andthe air velocity across the material, no single releaseIiactioncan be assumed.Sourceterms, releasefractions,and evaluation of accident consequencesare discussedfurther in SectionV.

Major potential sources of fires in nuclear facilitiesare summarized in Table IV. These major categoriesarebased on analysisof event historiesand accident analy-ses (Perkins 1980,1981).The probabilityof f~es with apotential for release of radioactive material may beestimatedfromevent histories.An approach to estimat-ingfirefrequencieshasbeen drawn from firestatisticsinthe chemicalindustry (Selby1975).

3. CriticalityAccidents.Acriticalityevent isin manycasesa localevent without offsiteimpact; however, the

processesin the facilityshouldbe reviewedfor potentialreleaseof radioactivityand for direct radiation at onsitelocations.Descriptionof this DBAshould include

●

●

●

form of the critiml material (liquid solid, or amixture),total number of fissionsexpecte~ and anamount of eachradionuclidein dispersibleform;times over which fissioningand radionuclide dis-persaloccur;containment of dispersible radionuclides andshieldingof direct radiation;andreductionand removal factorsapplied to dispersedradionuclideamounts.

Table V contains initial burst yields and total yieldsbased on Stratton’s (1967) review of criticality ami-dents.These yieldswerechosenby Woodcock(1966)aspotentialmagnitudesfor emergencyplanningpurposes.Subsequent guidance was issued by NRC (1977B,1979A,1979B)defining a “minimum” criticality acci-clentin a solution in which the total fission yield (1019fissions)dif%crsslightlyfrom the 3 X 1019fissions de-rived by Woodcock. The yields in Table V are con-sideredsuitablefor useby the analystwhen case-specificanalysis is not practical. Case-specificanalysis may beaided by experimental results and estimation methodsprovided by L.ecorche(1973),Hooper (1974),and Tuck(1974).

Radioactivity can be dispersed during a criticalityaccident by fission product generation, or by physicalaction on a solution containing actinidesor other solidradionuclides.The radioactivity available for dispersal

TABLE IV. FIRE ACCIDENT CATEGORIES

Fire Source

General

Zirconium

Organicsolvent

Hydrogen

Electrical

Sodium

Pyrophoricmetal

Cellulose

vehiclefuel,welding,poor housekeeping

fuel reprocessingfacilityin fuelbundleshearing

fuelreprocessingfacility,solventseparationcolumns,solventrecoverytanks, pipingleaks

radiolysisofprocesssolution

sourceof fue spread, lossof services

liquidsodiumspills

U, Pu metal production;scrap recovery;inerting at-mosphereloss

spontaneouscombustionof cellulosewipesand nitricacid

9

TABLE V. CRITICALITY ACCIDENT FISSIONYIELDS’

YieldSystem : (fissions) (fissions)

Solutionsunder 1 x 1017 3 x 101S100gal (0.46m3)

Solutionsover 1x 10I8 3 x 1019100gal (0.46m3)

Liquid/powderb 3 x 1020 3 x 1020

Liquid/metalpiecesc 3 x 10’S 1x 10’9

Soliduranium 3 x 1019 3 x 10’9

Solidplutonium 1x 1018 1x 1018

Large storagearraysd None 1x 1019(belowpromptcritical)

Largestoragea&aysd 3 x 1022 3 x 1022(abovepromptcritical)

---

as w a a powder~yer co~d result ‘n

progressivelyhigherreactivityinsertion.CAd s a w

by the criticality accident can be calculated by theORIGEN2 computer code (Croff 1980)for amounts offission products and actinides initially present in adissolved reactor fiel solution or by the RIBD code(Gumprecht 1968) for amounts of fission productsproduced by the excursion. Actual radionuclideamountsmade airborneby physicalaction (evaporationof solution)or by production during fissioningmay becalculatedon an individual basis or establishedby theassumptions made in the applicable NRC regulatoryguidecitedabove.The NRC assumptionsare as follows:

●

●

10

AUof the noble gas fission products (except thoseremoved before the excursion),25%of the radio-iodine, and O.1%of the ruthenium resulting fromthe excursion or initially present in the spent fielare released directly to the cell (or room) at-mosphere.An aerosol,generatedfrom evaporationof solutionduring the excursion,is releaseddirectly to the cellatmosphere. The aerosol comprises 0.05%of salt(solute)content of the solution that is evaporated.

Persons outside the facility may receive significantdose. Calculationsby Selby(1975)indicate major dosecontributionsat distancesas far as 100m resultingfromprompt neutrons and dose to internal organs frominhalation.

Radionuclides released from a criticality accidentwould normally be discharged to the atmospherethrough the facility stack. Choice of puff or plumemodel will depend on duration of the event (Hanna1982,P.42).

Frequencyof occurrenceof a criticalityaccident maybe estimated flom historic data for the types ofprocessesplanned.Probabilityofacriticzdityaccidentin ~a Pu orU fuelfabricationfacilitywasestimatedby Selby(1975) to be approximately 9 X 10-3 per plant year. ~-Becausethis estimate was based on criticalityaccidents _in solutions and a nearly equal number occurred insolutionsin other facilitytypes,this rate (10-2)per plant ‘=year is considered a suitable frequency for use in risk .assessment.The analystmay choose to perform a case-specificanalysisif a lower rate is expected.Probabilitystudies based on historical criticality safety violation

--

-.---

.-

data indicate the frequency could be assumed muchlower, depending on how soon a fault in a criticalitysafety program might be detected (Lloyd 1979).Thisduration of a faultwouldbe a site-specificor operation-specificdetermination.

Source terms, release fractions, and evaluation ofcriticalityaccident consequencesare discussed firtherin SectionV.

4. Leaks to the Atmosphere.A DBAmay result froma major leak to the atmosphere caused by equipmentfailure or operator error. Filter failure, by-passing ofremoval systems, storage tank rupture, dropped caskscausing fhel disruption, material spills outside of con-finement areas, and short coolingof spent reactor fuelare examples of this accident type. A detailed list ofpotential accidents based on event histories in fuelreprocessingplants is included in Perkins (1980).Leaksresulting from natural phenomena or other externaleventsare discussedin SectionsIV.B.and IV.C., respeetively.

The dispersal mechanisms for this potential DBAinclude mechanical ejection of disrupted material,atomization of radioactive solutions under pressure,aerosolizationof powders,surfam evaporation produc-ing an aerosol, and perhaps others. A summary ofparametersaffectingreleaseand transport of radioactivematerial is included in Walker (1978).Selby(1975)alsodiscussespotential releases from accidents caused byinternal initiators.

Frequencyof occurrenceof accidentscausingleakstothe atmospherecan be estimated from historicaloccur-rence data. Selby (1975) discusses failure-rate datasourcesas valuable predictors of release probability incaseswhereapplicabledata are available.

Source terms, release tlactions, and evaluation ofsuch accident consequences are discussed fim-therinSectionV.

5. Leaks to the Aquatic Environment.Some nuclearfacilities have the combination of a liquid mediumcontaminated with radionuclides and a nearby lake,river, or stream that mightreceivean accidentalrelease.Intake by ingestionof drinking water, consumption offish, immersion in the contaminated water, and con-sumption of food crops irrigated with contaminatedwaterare potentialpathwaysto be considered.

The discussion of accidental releases of liquid ef-fluents in Section 2.4.13 of the Stan&rd Review Plan(NRC 1981A)and its referencesmay be helpfid to theanalystin the evaluationof this typeof accident.Sourceterms, releasefictions, and evaluation of suchaccidentconsequencesare discussedfirther in SectionV.

Natural phenomena, particularly earthquakes andtornadoes, may be capable of acting as initiators ofmajor accidents at nuclear facilities.These events areevaluated both for their capability of disrupting thecxmfinementsystem (structure and engineered safetyfatures) and triggering other failures or effects con-tributingto dispersalof radioactivematerialbeyond thefacility. The discussion applies to evaluation ofproposed critical facilities or reevaluation of existingfacilities being significantlymodified for new opera-tions. In some cases, loadings may be higher than or-iginal design requirements. The facility modificationsshouldbe designedto withstand thesehigherloadings.

This section reviews major considerations in theanalysis of postulated accidents initiated by naturalphenomena:

Site selection-proximity to a known active seis-mic fault location in a region of high tornadofrequency,in a flood-proneregion, or in a regionvulnerableto hurricanes.Structural adequacy—resistanceof an existing ordesign-phasestructureto natural phenomena.Component adequacy—resistanceof critical com-ponents to natural phenomena and adequacy ofthese components in allowing the facility to beplacedat the proposedsite.Damageestimationand releasefraction-amountsof radioactivematerialwhichcouldbe releasedas aresultof postulateddamage to structuresand com-ponents.

DOE Order 6430.1, Chapter IV, provides generalrequirements for site selectionand designingresistanceto natural phenomena into all types of DOE facilities,including facilitiesfor handling processing or storingradioactivematerial, or other facilitiesconsideredcriti-calby virtue of their vital importance to DOE programs(Pu, 3H, and HE/Pu facilities) or high dollar value.Structural design of buildings other than these criticalnuclearfacilitiesmust complywith the latest edition ofANSI A58.1(ANSI 1982)for wind loads and the Uni-form BuildingCode(UBC)for earthquake loads (ICBO1982).

DOE has initiated a program to prepare site-specifictornado and earthquake hazard models. Compilationsofearthquake and extremewind/tornado hazard curvesfor many siteshave been publishedby LawrenceLiver-more National Laboratory (LLNL) (Coats 1984A,1984B).Related reports are becoming available forearthquake hazards (for example, Tera 1984). The

11

Purpo5esof these site-specfic models are twofold:iira~to tie a probability of occurrence to a maximum ex-pected magnitude an~ second, to characterize the de-sign basis event more specificallyin a locality thanpermittedby existingmethodsfound in NRC regulato~guides,ANSI standards,or buildingcodes.

Present guidance to be applied to critical items isexpressedin DOEOrder 6430.1as recurrencetime. Thisis 106yr for tornadoes; no direct recommendation ismade for earthquakes although 104yr is implied inChapter IV, p. IV-8, of Order 6430.1. A conservativeintfrim value for earthquakes is considered to be 104v.

1. Earthquakes. DOE Order 6430.1 specifies thatseismic resistance be provided in critical facilities towithstanda designbasisearthquake(DBE).The DBE isalso defined as a safe shutdown earthquake (SSE).Jointly occurring accidents should be considered if ajoint eventis likelyto be causedby the eathquake, suchas a tire or explosion.That is, a tire or explosionshouldbeassumedto occurunlessmitigationispresen~suchasnegligiblecombustibleloading or absence of explosivematerials.Becausethe DBEmust be assumedto occuratany time, certain loads, such as common wind loading,snow loading or intermittent maximum loadings(storage tanks, vaults, cooling pools, and the like),shouldbe added to earthquake loading.A detailed DBEanalysis may not be needed if a conservative simpleanalysisshowsanother accidentto clearlybe the DBA.

& Site Selection. DOE Order 6430.1 requires, asquantitative design basis for nuclear facilities againstearthquakes,proceduressimilar to 10CFR 100,Appen-dix& “Seismicand GeologicSitingCriteria forNuclearPower Plants” (CFR 1962).These siting requirementsas relatedto capablefaultsare stated as minimum lengthof fault to be considered versus distance from the site.Site suitability received Iirt.her elaboration by NRC(NRC 1975A)in RegulatoryGuide 4.7, “General SiteSuitabilityfor Nuclear PowerStations,”as follows:

. Sitesthat includecapablefaultsare not suitablefora nuclearpowerstation.

. Sites within about 5 miles (8 km) of a surfacecapable fault ~eater than 1000 ft (305 m) inlength] are generally not suitable for a nuclearpowerstation.

These conclusions are also considered applicable tocritical DOE nuclear facilities based on DOE Order5480.lA specification of comparability with licensedfacilities.

WhisvaluesuppliedbyDavidW.Coats,LawrenceL.iv-ermoreNationalLaborstory,November1983.

12

b.StructuralAdequaqy.Adequacyofcriticaland non-criticalnuclearfacilitystructuresto withstand vibratoryground motion shall be verified, according to DOEOrder 6430.1,using a suitable dynamic analysis tech-nique,exceptwhere the static responsetechniqueof the ‘-Uniform BuildingCode (ICBO 1982)can be shown to -provide a conservative estimation. A common and -recommended design approach is sizing of structural L-.members to meet static loading, then applying a dy- -narnicanalysismethod to checkoveralladequacyof the I

structure. The maximum (peak) loads determined bydynamic analysis are then combined with dead loadsand other live loadsas describedin DOE Order 6430.1.

Figure 1 is a representation of logicalsteps requiredforanalysisof criticalnuclear facilities.

(1)Descriptionof theDBE.The DBE is describedbysite-specificspectra and recurrence time versus peakacceleration data, if available (Coats 1984A).If site-specific&ta are not available,one of the followingmaybe used:

●

●

●

The simplest (and acceptable if shown to be ade-quatelyconservative),the UBC SeismicZone 3 (4in California and Nevada) description for staticanalysis(ICBO 1982).Regionalresponsespectra based on historical list-ing of all known earthquake activity in the region(200-mile radius) supplemented by geologicalevidencebeyondthe historicalrecord(10CFR 100,AppendixA, CFR 1962).The characteristics of a single historical earth-quake, which in the absenceof specifichistorical&ta for the region, is believed to conservativelyrepresent the most serious earthquake expectedatthe site. The El Centro earthquake of 1940 is anexample of earthquake characteristicsselected forthis purpose(Newmark 1978).

(2) StaticResponseMethod.Earthquakeresistanceofa noncritical facility structure may be determined bystaticmethods describedin the Uniform BuildingCode(ICBO 1982).Eagling*suggeststhat Zone 3 classifica-tion be the minimum selected,regardlessof the locationof the facility, and that Zone 4 should be selected inregionsof Californiaand Nevada. Althoughthis sugges-tion may lead to apparent overdesignat some locations,it acknowledges the uncertainties in predicting theseventy of the DBE for a given site. It is also question-able whether major savingsin building costs would berealizedif a lowerzone classificationwere assumed.Anindication of conservatism contained in structures de-

UBC-relatedcodesis discussedin the SeismicSafetyGuide, p. 4-1(Eagling1983).

%s work providedby D. G. l%glin~ LawrenceBerkeleyLaborstory,December1,1983.

--/

.—-

F L C S A

+- ‘-I

t1

] M

I R E I

I D C I

(3) DynamicAnalysisMethod.A variety of dynamicanalysis methods are now available. Descriptions ofdynamic analysis methods are found elsewhere(New-mark 1973,ASME 1980,and Clough 1975).

(4) Determinationof Damping. Calculation of per-eent of damping in the structure is subject to majoruncertainties. Conservative damping values based on

. experimentaldata have been determined (Bohm 1973,Hart 1973,Newmark 1978).These values maybe used

-- unlesshighervaluesem be justified.-1-

(5) Strength Analysis (Combined Response). The.- eombined response from all sources of loading during

the DBE should be accounted for. These responsescaused by the earthquake must be combined with thedead load of the structure in the manner described inDOE Order 6430.1.Specificequations are provided forreinfore.edconcrete structures (elastic only) and steel

structures (elastic or plastic design method). The ade-quacy of critical structures and components should beverified for horizontal and vertieal motions, with themt.ioof vertical to horizontal accelerationset at 2/3 byDOE Order 6430.1 unless a different ratio ean bejustified. The effketsof tipping, tilting, and rotation ofthe ground during an earthquakehave not been studiedextensivelyand usuallyare not analyzed.

c. Compowst A&quaqv. Dynamic analysis of ESF,safety, and confinement system components in a nu-clear facility is performed to assure their continuedoperation throughout the DBE. Failure modes are ex-amined to evaluate integrityof the gloveboxes,vaults,pools, tanks, and other confinement components.l~ossiblef~l~e modes include ftilure Of SUppOr@ortiedowns,window breakage, filter seal breakage, inter-ruption of ESFoperation through lossof power,and thelike.

13

The floordesign, response-spectramethod and thetime-histoxymethods similar to those applied to thebuilding structure are used in analysis of componentadequacy. A basic approach to these methods is dis-cussedin RegulatoryGuide 1.122(NRC 1978A)and theASMEBoilerand Pressure VesselCode (ASME 1980).Newmark (1978) has provided usefid discussion andreferencesregardingcomponentadequacy.

d. DamageEstimationand ReleaseFraction.Failureof structuredmembers indicated by one of the analysismethods discussed earlier may or may not affect theconfinement system or otherwise cause an accidentalreleaseof radioactive material. Assumptions regardingreleaseamounts may depend on severalconditions:

●

●

●

proximity of a failed member to the confinementsystem(that is, does the roof collapseonto a glovebox?);availabilityof an energysource that transports thematerial through a breach in the confinement sys-tem; andsizeof the postulatedbreach.

Release fictions should contain a suitable level ofconservatismin the absenceof analyticalor experime-ntaldata. Analysesby Selby(1975)and M.ishima(1979,1980,1981)may be usefidin arriving at suitablereleasefictions. For example,a tlaction (rather than all)of theradioactive material in crushed or perforated gloveboxes may be dispersed (Mishima 1981,Mehta 1978).Asan alternative,a releaseto the room maybe assumedto reach some upper value of airborne concentration inthe room, such as 100 mg/m3 Pu aerosol in therespirable size range (Selby 1975).Removal by the aircleaningsystemmay be assumedaccordingto the redu~tion and removal tlactions listed in SectionV.C. only ifanalysis shows that the air cleaning system remainsintact.

2. Tornadoes DOE Order6430.1 specifies that critical items and systems in anuclear facilitybe designed to provide confinement ofradioactive material under DBA conditions, one ofwhichis the designbasistornado (DBT).ANSI/ANS2.3also discusses guidelines for determining tornadoparameters (ANSI 1983).The DBT shall not cause thesiting guideline doses to the offsite person to be ex-ceeded.A determination of dose requires knowledgeofor assumption of the amount of radioactive materialreleasedand its dispersion under the chaotic meteoro-logicalconditionsof the tornado.

a. Site Sektion. Tornadoesand extremewindshavenot playeda major role in siting of nuclear facilitiesornuclear power plants, because safety-relatedstructuresand systemscan be designedto resistmost atmospheric -extremes (RegulatoryGuide 4.7, NRC 1975A).How- ~ever, a site in a regionwith relativelyhigh frequencyof .strongtornadoesshouldnot receiveequalconsideration :.=with a site in a more favorable location. Strength and .effects of tornadoes are not well known, and present -design methods cannot guarantee a “tornado-proof’ 1nuclear facility.Guidance on what tornado strength is \too greatto allowdesignofa facilityof reasonablecost islimited. However, this should be considered in the siteselectionprocess.

b. StructuralAdequacy.Design of a critical nuclearfacilityis specifiedbyDOE Order 6430.1to resista DBThavingthe characteristicsof a tornado with a recurrencetime of 106yr. Curvesof recurrenm time and maximumwind velocityare provided for major DOE sites (Coats1984A).Other sites without site-specific tornado orhigh-windprobability data may use characteristics ofthe DBT in the appropriategeographicalregionsshownin DOE Order 6430.1.Tornado resistancecalculationsmust considercombinedloads resultingfrom rotationalplustranslationalwind speed,rate of pressuredrop, andmissiles. Combining these loads is accomplished ac-cording to the equations shown in DOE Order 6430.1.The componentsof tornado or wind-loadcombinationWtare discussedby Vellozziand Healey(Vellozzi1973).The wind-load forces developed on various waUsandroof of a structure hit by a DBT are calculated by themethodsdescribedby ANSI A58.I (1982).

c. ComponentAdequacy.Adequacyofcriticalcompo-nents (ESFand safetysystemscomponentsrequired forsafe shutdown and confinement of radioactivematerials) in resistingthe DBT should be assessedun-less the building remains intact, missile barriers areprovided at ventilation intakes and exhausts, and ade-quate strength of the components to resist pressureeffectsis demonstrated. Experimentaldata from whichcomponent strength might be estimated are limited.Simulated tornado effect experiments have beenperformed on HEPA filter systems that indicate as-sumptionsof filter type and filter loadinginfluencethe

.

tornado-inducedpressurepulse strongenough to break -a HEPA filter (Gregory 1982,Horak 1982).Although ‘>these experiments were performed on single filtersrather than multiple-stagebanks of tilters, their results _-indicatepossiblebreakageofhigh-capacityHEPA filtersifan averagemaximum differentialpressureof approx-imately 1.6 psi (11.0 kpa) is exceeded; for standard

14

HEPA filters,2.4 psi (16.5 lcpa).A reduction of colle(ition efficiencycan alsobe exmcted if the pressurepulseis survivedwithoutstructuralbreakageof the tilter.Thiscollectionefficiencywas measured to be approximately

,--., 70%and representsa reasonable removal fi-action(perstage) if other design and damage considerations in-dicate the DBT will not cause breach of the facilityor4pressurepulsesare not high enough to break the HEPAiilters.

d. Damage Estimation and Release Fractwn. Thequantity of radioactive material released from the fa-cilityby a DBTwilldepend on the sourceterm and whatreductionand removal fractionsare assigned.Extent ofdamagefrom tornadoesto structuresand gloveboxesorsimilarconfinementcomponents is discussedby Mish-ima (1979, 1980, 1981). This information suggestedvarious fractions of the contents of the confinementsystemwhichmightbe assumedreleasedto the buildingair cleaningsystem.Intact filter stagesmay be assumed70%efficient,as discussed in c. above. Gregov et al.noted partial release of existing filter load nom intactfilters, which is a release source not yet well defined(Gregory1982).Tornado-inducedfloweffectsand reen-trainment have been studied by lhdrae et al. (Andrae1979)usingthe TVENT computer code (Andrae 1978).

e. DisperswnAssumptions.Three approaches to me-teorologicaldispersion of released radioactive materialhavebeen used in past DBT accidentanalyses:

●

o

●

dispersionby straightwindsaccompanyingthe tor-nado;uptake of the material by the funnel, followedbyreturn to the ground in heavy precipitation;anduptake of the material by the funnel, foliowedbyinfinitedispersion(assumednegligibledose).

This third approach is not consideredacceptablefor theDBT accident analysis. Dispersion by straight windsaccompanyingthe tornado was considered appropriateforan HE/Pu detonation casein the Pantex EIS(Dewart1982).Uptake by the fimnelcloud followedby depositon the groundby precipitationpresentsa difficultprob-lem in establishingthe location of the exposed offsiteperson because the DBT has translational motion andunspecified nonuniform precipitation. A natural-.

,- phenomena analysis of a plutonium fuel fabricationfacility(Pepper 1978)and the Rocky Flats safetyanrdy-

._ sesand risk assessment(RFP 1982)used this approach..

3 Other Natural Phenomena.The natural phenom-ena other than earthquake and tornado have not beentreated in comparable depth because the hazard ofradioactivematerial releasedby a relatedaccidentis notexpected to be comparable. According to DOE Order6430.1,“designloadsand considerationsfor other natu-ral phenomena shall provide a conservative margin ofstiety greater than the maximum historical levels re-corded for the site. Protection against floodingshall bebased on no less than the probable maximum flood(PMF)for the area as definedby the Corpsof Engineers.The possibilityof seismicallyinduced damageor failureof upstream dams shallbe taken into account in assess-ing the nature of flood protection required for thefacility.”

C. AccidentswithExternal Origins

Each nuclear facility exists under some probabilitythat an offsite or onsite external hazard may cause abreach of the confinement systemresultingin a releaseof radioactive material. DOE Order 6430.1 specifiesthat each facilitybe evaluated for all hazards, such asfire, explosions,gas mains, explosives in large quan-tities, flammablegases(wewould add largeonsite vehi-cles),and potentiallyhazardous external (offsite)opera-tions such as airports and private industry. Evaluationshouldincludeboth an estimate of the probabilityof anexternal occurrenceand the magnitude of a release ofradioactivematerial from damage caused by the occur-rence.

site should be evaluated forany potential release of radioactive material caused byan explosion,leakage,or tire at any nearby volatile fuelor toxic chemical transportation route. The accidentalrelease could result from missiles, direct blast forces,fire, leakageof volatilefuelnear or into the facility,or aleak of toxic chemicals rendering the facility un-inhabitable.Determination of structural or componentdamageand assumptionsregardingreleaseamounts canbe approached similarly to tornado missile analysis(Section IV.B.2), to fire-related releases (SectionIV.A.2), or to blast force damage similar to tornadodifferentialpressure forcedamage. Uninhabitability is-sues should be based on inability to perform safe shut-down followingexposureto leakingchemicals.Variousaspects of external hazards are discussed in the NRCStandard ReviewPlan, Section2.2.1(NRC 1981A),andin severalregulatoryguidesreferencedtherein.

15

■

2. Aircraftand Airports.The risk of releaseofrad.io-active material caused by an aircrafl crashing into thefacility increaseswith proximity to an airport. Unlessthe crash risk can be shown probabilisticallyto be lessthan 10+ per year, the facility should be analyzed forvulnerability of the confinement system to damage.Assumptionsregardingamounts of radioactivematerialreleased to the atmosphere should be consistent withamounts releasedby other eventscausingdamageto theconfinement system, such as natural phenomena (Sec-tion IV.B.) and operational events (Section IV.A.). Inparticular, an accompanying fuel fwe should be as-sumed. Various aspects of aircraft and other missilehazards discussed in the NRC Standard Review Plan(Section3.5.1.6)and its referencesmay be helpfulto theanalyst in evaluatingthe hazard of aircraft crash (NRC1981A).

3. Other Nuclear Facilities or Reactors. The risk ofexposureof the facilityto radioactive material releasedaccidentallyfrom nearby nuclear facilities or reactorsinvolves possible overexposure of personnel, loss ofhabitabilityof the facility,and possibleextendedlossofimportant operations. An analysis should demonstratethat an accident at a nearby nuclear facility will notcausean accidentat the proposedfacilityand that it canbe safelyshutdown and evacuated.

4. Large Dams. Presenceofa largedam upstream ofa facilityis cause for an evaluation of risk to the facilityin the event of dam failure. Section IV.B.3 containsrelated requirements stated in DOE Order 6430.1.Sec-tions 2.4.2and 2.4.4of the NRC Standard ReviewPlan(NRC 1981A)and its referencesmay be helpfld to theanalystin evaluationof this potential accident.

5. Explosive or Toxic Material Facilities. Thepossible effect of a major explosion at a nearby ex-plosives facility would be analyzed like the explosionhazard fromfiel transportation (SectionIV.C.1),that is,for vulnerabilityof the proposedfacilityto blast effects,missiles,or fire.halysis of hazards from a nearby toxicmaterial facility would be approached on the basis ofpossible release of radioactive material due to un-inhabitabilityof the facilitywithout safeshutdown.

D. AccidentswithHigher Probability

Doses resulting from an accident with lower conse-quencesand higher probability of occurrence than theDBA may also be compared, where appropriate, withother dose guideLines lower than the guidelinesproposed for DOE Order 6430.1.These accidents withlowerconsequencesmay deserveevaluation becauseof

a potential for exposure of the public and workers atnearbyfacilities.The guidelinedosesor designgoalsfortheseaccidentswill then depend on their probabilityofoccurrence.Determination of this probabilityshouldbebased on failure data where possible. A quantitativemethod currently in use (Durant 1980, 1981) basesaccidentprobabilitieson incident frequenciesrecordedin extensive data bases. Because the relationship be-tweenfrequencyof incidents(componentfailures,oper-ator errors, etc.) and accident probability is not wellknown, subjective judgment cannot be completelyeliminatedfrom the process.Related discussionswhichmay be of use to the analystare found elsewhere(Swain1983,on human reliability analysis; Briscoe 1982,ongeneraltopicsof risk management).

Several approaches to establishing a structure ofsafetyguidelinesin terms of probability of occurrenceand consequence (dose, in this case) exist within theDOE complex. A qualitative evaluation approach iscurrently being implemented by several field offkes(Lucas 1981,ALO 1982,and ORO 1984)and has beensuggestedseparately by Brynda et al. (Brynda 1981).This method requires subjectivejudgment in assigningan accident to a probability class (that is, anticipated,unlikely, extremely unlikely, or incredible). Theseclassesare assigneda range of dose guidelinesthat arefractions of the siting guideline doses. The fractionultimatelyselecteddependson the analyst’sdetermina-tion of probability;the more probable the accident, thelowerthe guidelinedoseselected.Althoughthis methodleaves more to the analyst’sjudgment, it provides asystematic approach to evaluation of lesser accidentsthan the DBA.

TableVI representspotentialcategoriesofprobabilityand ranges of dose. These values are similar to thosereferenced above but were adapted to current DOEsiting guideline doses. A similar approach in ANSIStandard N287 (ANSI 1976)and in field offke ordersfor implementationof DOE Order 5481.1A(ALO 1982,ORO 1984)maybe helpfil to the analyst in devisingasuitable structure of safety guidelines for these acci-dents.

V. EVALUATIONS OF ACCIDENT CONSE-QUENCES

The analyst should predict with acceptableaccuracythe behavior of a DBA under the assumed conditionsmost appropriate for the proposed nonreactor nuclearfacility. Verification of assumptions, where possible,may be derived from comparisonswith existingverifi-able experiences/experimental results and in somecases,with experimentalresultsyet to be gathered.Theconfidence one can have in predictions of accidentbehavior is primarily based on these comparisonswith

.

.

---

16

—-

TABLE VL POTENTL4LR4DIOLOGICAL DOSE GUIDELINES FORACCIDENT EVALUATION

DoseGuideline(rem)

NominalRangeProbability of Probability Whole Bone OtherCategory

A >1O-* <0.01 <0.03 <0.12 <0.12 <0.06

Unlikely 1 0.01-0.50 0.03-1.5 0.12-6 0.12-6 0.06-3

Extremelyunlikel~ 1o-C1O-4 0.5-25 1.5-75 6-300 6-300 3-150

Incrediblee <lo-d

r

facfity.~~s mtegoryincludes

a

experienceand experimentaldata. Experimentalverifi-cation, when available, should demonstrate that theindividual mechanisms or processes(accidents)affect-ing a release are adequately described analytically;allsignificant mechanisms (accidents) are included; andinteractionsamong individual mechanisms (accidents)or processesare properlydescribed.

Evaluationof potentialDBAsof the proposedfacilityor operation involves

@descriptionof the sourceterm,* calculation of the dispersion factor ~Q) of dis-

persedmaterial at the point of interest,* calculationof doseat the point of interest, and● estimation of other c w s

considered.

Figures2 and 3 provide a graphicalmodel of the stepsinvolved in each accident analysis.The model may beadjusted appropriately to accommodate variations-among DBAs,but these figurescontain the major cate-gories of analysis and description. The followingsec---tions provide additional detail and discussion of these\major categories.

--

The sourceterm is defined as the amount of radioacitive material available for release after the fraction ofreleasefrom primary confinement is applied. It maybe

the total amount of radioactive material in process orstoragebut is usuallya smalleramount followingmodi-fication by the release fraction. Release fraction is thefractionof the total availableradioactivematerial that isreleased from primaxy confinement in a readily dis-persible form. It is assumed that readily dispersibleradioactive material is capable of causing radiologicaldose, either by direct radiation or by inhalation of therespirable fraction or by ingestion. The source termdescriptionusually includesa list of radionuclides,thequantity (Kg or Ci or Bq) of each, particle sizecharacteristics, and chemical form. These latter twofeaturesare discussed in Section V.F. in considerationof their role in radiologicaldose.

1. Radionuclides.R.adionuclidesof interest may beoriginally present or are fission products released bydisruption of spent fiel or are produced by a criticalityaccident. They may be in a pure form or mixed withother radionuclides.The list should include all radio-nuclides contributing more than a few percent of theactivity of sourceterm after coolingtime or decay timeis allowedfrom the time of the accident to the time theradionuclide is inhaled by the exposed person. Eachrad.ionuclideshould be screened according to its con-tribution to the dose.A suitablemethod can be based onrankingof the ratio of quantity at the time of intake andthe annual limit on intake (ALI) of the nuclide flomICRP 30(ICRP 1979).Anothermethod is the rankingofrad.ionuclidesby the product of quantity (Ci or Bq) atthe time of intake and organ dose factor (rem/Ci or

E I

N M

E AT EF

OI I 1

EA

R

v IS C

r

A, R

// / /,

E C / / z

I I I \ ~D D D D

R M Q N RQ P F F FL F P S TI F V O

Sv/Bq) appropriate to the intake pathway. The radio-nuclidesto be included in dose calculationscould thenbe all those contributingsignificantlytoward total doseto any organ (for example, to 99%or some lowervalueconsistentwith the uncertaintyof the dose calculation).Nuclide quantities and dose evaluation are discussedfurther in SectionsV.A.2.and V.F., respectively.

Table VII contains a list of radionuclides (fissionproducts, activation products, and actinides) that arelikelyto figurein accidentaldosecalculations.The list istaken from analyses involving accidents followed byvery little decay time (reactor accident, short-cooledfuel,or criticalityaccidents)and with longerdecay time(reactor fhel reprocessingaccidents)(WASH 1400,Ap-pendix VI, pp. 13-21,NRC 1975B).Although this listshould not be consideredall-inclusivefor all combina-tions of nuclidesand potential accidents,these nuclidesshouldbeconsideredin the sourceterm descriptionsformajor accidents. Sources of information on nuclideswith major dose potential are WASH 1400 for spentreactor tiel (NRC 1975B)and ICRP 30 (ICRP 1979,1980).

Activation products are of limited importance formostaccidentsin DOE nuclearfacilitiesbecauseof their

general lack of mobility. However, special cases mayexistthat shouldbe investigated.

2. RadionuclideQuantities. The source term for aradionuclide in dispersible form will normally be thetotal activity present multiplied by a release fraction(discussedlater). Quantities of fissionproducts, activa-tion products,or actinidesthat could be dispersedfromspent fiel are determined on the basis of power history(burnup) of the feed material and its cooling time. Asuitable source for fission product and actinide inven-tories is the ORIGEN2 code (Croff 1980). RIBD(Gumprecht 1968)is also used to obtain fissionproductinventory. These codes require as input the fissileen-richment of the fuel, the average neutron flux (neu-trons/cm2. s), the total uranium or other heavymetal inthe core, and the irradiation time. Code output is atabulation of actinide, activation product (ORIGEN2only), and fission product activities as a function ofcoolingtime. If ORIGEN2 or a comparablecode is notavailable, a simplified determination of the fissionproductand actinidesourceterms of cooledreactor fuelmay be performedby scalingfrom exampleequilibriumsource terms for a standard set of conditions, if

-.

---

.-...-

18

---

---

.

.-

3ATMOS–P

R

A

R

A>

D L;

7

CONSEQUENCES

I E

C I

F=l ED D D

M D I

D P I

D M I

M M

M

similaritybetweenthe reactorsis observed.Scalingfrom The solid fission procluctcategory was apparently in-example-values in the ORIGEN2 manual is p~ssiblebecausethe inventory of the long-livedradionuclidesisproportional to burnup (powerdensity times time) andis not sensitiveto powerdensityat any given time.

B. ReleaseFractions

The release fraction is that fraction applied to thetotal radioactivematerial in processto obtain the sourceterm (that amount released from primary confinementin dispersible form). Release ffactions established inTID 14844(AEC 1962)to allowcalculationof dosesforcomparison with 10 CFR 100 dose criteria are asfollows:

noblegases 100%,halogens 50%,andsolid fissionproducts 1%.

tended to also include the s~rnivolatile‘solid fissionproducts(Ru, Rb, Cs, Te, Tc, and Se).A later additionto this list wasa Pu and U or other particulateactinidescategory of 1%(NRC 1977A).These release fractionswere nonrnechanistically determined for use in sitesuitabilitycalculationsfor light-waterreactorswherethesystemisquite complex.Their applicabilityto accidentsin nonreactor nuclear facilitiesis not clear; their use asan upper limit in SARSis common. However, nonrea-tor nuclearoperationsin DOE facilitiesare, for the mostpa~ relatively straightforward.The accident sequencemaybe readilyunderstandable,compared with the reac-tor case,and use of other lowervalues maybe justified.In specific accident cases for which applicable ex-perimentaldata exist, lowervalues may be considered,if conservative.Table VIII presents an excerpted sum-mary of recommended safetyanalysisparameters com-piled by Walker (1978). These values are based onexistingexperimental data. Some values are justifiably

19

TABLE VII. IMPORTANT RADIONUCLIDES FOR ORGANDOSE CALCULATIONS*

ReIease Whole Bone BoneRadionuclide Bodyb Marrow Lung Surface Otheti

He-3 1Co-58co-doKr-85

Kr-85mKr-87 1Kr-88 2

Rb-86Sr-89 2 2 1Sr-90 1 2

Sr-91 1Y-90Y-91 1 1 1

Zr-95 1 1Zr-97Nb-95 1

Mo-99 1Tc-99mRu-103 1 1 1 1

Ru-105 1RU-106 2 1Rh-lo5

Te-127Te-127m 1Te-129

1 1 11 1 11 2 2 2 2

----

---

Sb-127 1Sb-129 11-131 1 1 1 1 1

.

1-132 2 1 1 11-133 2 1 1 11-134 1

1-135 2 1 1 1Xe-133 1xe-135 1

--.

20

--

--

TABLE VII. (cent)

CS-134 2 1 2 2CS-136 1 1CS-137 1 1 2

Ba-140 2 1 1La-140 1Ce-141 1

Ce-143Ce-144 2Pr-143

Nd-147NP-239 1Pu-238 2 1

Pu-239 2 1Pu-240 2 1Pu-241 1

Am-241Cm-242 1 1Cm-244 1

2 =1=

= c

a

.-

lower than the TID 14844criteria (AEC 1962)and areconsideredsuitableexceptas notedbelow.

● Total fissions from a criticality in a large liquidsystemmight be as high as 3 X 1019fissions;fromcriticality in a liquid/powder system, 3 X l~”fissions(SectionIV.A.3).

● Nonvolatile solids released from solutions spilledwithinbuildingconfinementare expectedto be lessthan 10-2%.*

The experienceat TMI-2 and reeent experiments atORNL and elsewhereindicate the 50%halogen releasefmctionmaybe too large(NRC 198IB).Smallerreleasesare expecteddue to a largeamount of elemental iodine(12)reacting with cesium to form CSI, a much lessvolatileand more water-solubleform than 12.Althoughit is apparent that a reduction in release fraction is

forthcoming,a consensuson a smallerreleasefraction islackingat the present time.

Semivolatile solids (called volatile solids in TableVIII, and includingRu, Sb, Rb, Cs, Te, Tc, and Se) arereleased over a broad range of release fractions (up to80%). Beeause of the radiological importance of thisnuclidegroup,each caseshould be reviewed.

Release fractions of other materials, such as pluto-nium or uranium compounds in powder form, havebeen reviewedand summarized by Walker (1978)andSelby (1975).Their release fractions depend on manyvariables, including temperature, air velocity,and par-ticle size. However, a simpli~ing assumption of 100mg/m3 airborne concentration atler settling and ag-glomerationof particleshas been used.

C. Reduction RemovalFactors

~is workprovidedbyW.S.Durant,SavannahRiverLabo-ratory,August1983.

Reduction and removal factors are those factors bywhich the released radionuclides may be reduced by

21

S

C PP

—

—

~ X

—

3 x4 X

———

—

1 x x

3 x

—

a

1x xx x

1x7 h

e

x

x

d

* p

1 1 1

either natural or engineeredmeans. A natural reductionfactor commonly applied to halogen release is 50%plateoutfactorallowedby NRC (1977B,1979A,1979B).This 50%of the 50%inventory releaseresults in 25%ofthe halogeninventory reachingthe containment barrieror the removalsystem.Anothernatural removalprocessis depositionof particlesfrom a cloud of releasedmate-rial, both inside and outside of the facility.Depositionoutside the facilityis consideredmost important and isdiscussedin SectionV.E.

Engineeredsafety featuresreceivecredit for removalif they are designed, installed, tested, and maintainedaccordingto prescribedstandards (Brynda 1981).TableIX provides a summary of removal factors from theliterature (Walker 1978).Although not all these factorswere investigatedindependentlyduring preparation ofthe Guide, the recommended values in Table IXprovide a basis for common usage, with the possibleexception of HEPA filter efficiency.Three situationscan be envisionedunder whichdecisionsmust be maderegarding credit for HEPA filter stages. First should

22

creditbe allowedfor any stageof HEPA filterswhich isnot protectedagainstheavysmokeloading?In this case,protection is not likely to be realized at any stagebecauseupstreamHEPAstagesare likelyto fail.Second,what credit shouldbe allowedif the postulatedaccidentisunlikelyto affectthe HEPAfilters(no common failureoccurs)? In this case, it appears acceptable to allowcredit commensurate with in-place test results, that is,approximately99.95%.Third, an unclear situation mayexist in which degraded performance should be as-sumed. The practice set by DOE-AL (ALO 1971)usesthe followingefficienciesfor HEPA faltersunder acci-dent conditions,if it maybe assumedtheyare testableasinstalled and protected by prefilters, sprinklers (orequivalent),and demisters:

● 1ststage 99.9%,. 2nd stage 99.8%,. 3rd stage 99.8%,and. 4th stage 99.8%.

.—

—

--

I

S

P. e

—

—e

< 70YOC>

@ <@ >

--

-.———

——

r%

ndsorber.

RegulatoryGuide 1.52recommends 99%in case of anaccident(NRC 1978B);this is consideredtoo low.Faustet al. assumed only one stage of operable filtrationproviding 99.9% removal (Faust 1977); this is con-sidered too restrictive when other tested stages arepresent. A general discussion of HEPA filter creditunder accident conditions is contained in the NuclearAir CleaningHandbook (Burchsted 1976).

Sand filter eficiency is expected to be 99.5%,basedon extensive operating experience at the SavannahRiver Plant (Orth 1980).This value is the 97th percen-tileefficiencyexperiencedwith operatingfilters;99.97%is the median efficiencyfrom the same data. This value(99.5%)may be used to calculate accidental releases

. (including fire, if the filter has been suitably sized orprotectedfrom pluggingby combustionproducts).

Halogen removal is affected by humidity, loading,and its chemicalform. The accidentdescription shouldincludeestimation of each effectand choiceof etlicien--_cies which lead to conservative estimates of release..Those suggestedin Table IX are considered suitable

except for 30% efficiency against CH31 at highhumidities.Althoughrecommendedby NRC in Regula-tory Guide 1.52(NRC 1978B), the 30%figureis at thelow end of experimental data summarized by Walker(1978)and is not considered conservative.An alterna-tive value in the range 50 to 8096is considered moreappropriate.

D. ReleaseDuration

Releaseduration for the accident case in nonreactorfacilitieswill usually be short (less than 8 h). In mostcases, the total release may be averaged over the totaltime of release;however,someeffort should be made tocharacterizethe timing of the release.If a peak concen-tration exists during a short period of the release, thatperiod and concentrationmay be appropriate for use inthe dispersion and dose calculations. Evacuation orother mitigation may affect the concentration selectedfor the exposure.

2

E. MeteorologicalAnalysisand Dispersion

Dispersionof radioactivematerial in the atmospherefollowinga major postulated accident requires use ofmodels that account for the major features of the acci-dental release: its relatively short duration, its cloudmore often a puff than a plume, and a potentiallyhighconcentration of radioactive material. The postulatedaccidentcaseis simplifiedsomewhatby its brevity;thatis, it may be assumed in most cases that no majorchange in meteorologyoccurs during the duration ofclouddispersion.

This section deals with a suitable model, its generalacceptability, and the approximate range of its ap-plicability. Choice of model and the adjustments tomodels may have a major impact on the result. It issuggestedthat the analyst consider each adjustment asan important sourceof additional accuracy,whilekeep-ing in mind the generallevel of uncertainty inherent ineachcalculationleadingto the dose.Asin all other stepsof dose calculation, conservativism in the selection ofdispersionmodelsand parameters is suggested.

Gaussian Dispersion Model. The straight-lineGaussiandispersionequation (Slade 1968)is in generaluse to model dispersion of chronic and accident re-leases. Its use for estimating the time-integrated airconcentration ~(Ci”s/m3 or Bq” s/m3) at downwindlocationsis recommendedfor most accidentconditions,unless site-specificmodels of some other form havebeen developed and verified. The basic Gaussian dis-persion equation has been adjusted by various authorsto accommodate release and dispersion effects.Theseadjustments are discussed briefly in the followingseetions and in greaterdetail in AppendixD. The range ofdistances over which the Gaussian model should beused varies with conditions (see Section C-I in Appen-dix D) but the model is consideredgenerallyapplicableover the range 0.1 to 10-20km. Beyond 20 km resultsshould not be considered better than order-of-magni-tude estimates.

2. Dispersion Parameters/Stability Classifications.The horizontal and vertical dispersion coefficients,CYand an required in the Gaussiandispersionequation areobtained either from site-specific meteorologicalmeasurements (standard deviations of wind angles)orby estimating an atmospheric stability class for whichstandard coeffkients have been established. Methodsfor using site-specificdata are cited in Appendix D. Ifthe necessary meteorological measurements are notavailable, several methods for determining stabilityclass may be used. Examples are contained in Turner1970,NRC 1974,and NRC 1983.See Appendix D forothers.The vertical temperature gradient method is no

longer recommended by the AMS (see Appendix D).Assumptionof a conservativestabilityclass (for exam-ple,slightlystableor moderatelystable)is an acceptableand frequentlyused method for ground-levelreleases.

Determination of CJYand ISZfrom existingcurves iscommon and acceptable practice. For CY,the AMS(1978)has suggestedthat the Pasquill-Gifford(Gifford1961) curves and the McElroy-Pooler (1968) urbancurvesare acceptableif adjustments are made for sam-plingduration and surfaceroughness.Curvespresentedby Briggs(1973) have combined data from Pasquill-Giffordand severalother sourcesto describedispersionofelevatedreleases.

The differencesbetween puff and plume dispersionbythe Gaussiandispersionequationcan be (and usuallyshould be) accounted for in the accident case. Methodsfor calculating puff dispersion coefilcients have beenaddressedby Gifford (1977),Hanna (1982),and Turner(1970).

3. Release Effects.The dispersionequation may re-quire modifications due to release effects, notablyplume or puff rise (buoyant or momentum) and build-ingwakeeffects.

IZ PlumeRise. Credit for plume rise has not alwaysbeen taken in SARSand regulatory guides. However,calculation of an effective release height above stackheight is considered reasonable for the accident case.Equations appropriate for the descriptions of buoyantand momentum plume rise have been presented byBriggs(1969, 1975).An expressionfor cloud rise fromhigh-explosive detonation has been presented byChurch(1969).

b. Stack andBuildingWakeEff&ts.Stack tip effectswill be observed if exit velocity does not exceed 1.5times the wind velocity.Buildingwake effectswill alsobe observed if the releaseheight does not exceed 2-1/2times the height of nearby structures. Adjustments fortheseeffectsare discussedin AppendixD.

4. DispersionEffects.As the cloud of releasedmate-rial moves downwind, several dispersion effects mayalter the air concentrations obtained by the Gaussianequation. Of these, radioactive decay, plume trappingby an inversion, dry deposition, and fumigation areconsidered of possible importance and are discussedhere and in Appendix D. Wet deposition, althoughconsideredan inappropriate cloud depletion effect forthe accident case, is discussed briefly in Appendix D.Chemical change of state by absorption of a gas byatmosphericwateror the chemicalreactionwith vegeta-tion may be identified removal mechanisms for some

..-

.

.-

.

-

24

compounds; these are not expected to be importantremovalmechanisms.

a. RadioactiveDecay.Radioactive decay will be im-portant only in depleting short-lived radionuclides re-

-.. leased from a criticality accident. Decay effects areaccountedfor in the decay and buildup modelscited inSectionV.A., SourceTerms.

--

h InversionLid. As a generalrule, unlimited mixingdepth should not be used. Dispersion analysisof somelocations with chronic inversions lasting more than afew hours must include some restriction of mixingdepth (Turner 1970).Severalalternatives for determin-ingmixinglayerdepth are discussedin AppendixD.

c. Fumigation.Fumigation conditionsare capable ofcausing locally high concentrations. Modifications tothe Gaussian equation to account for frequent fumiga-tion conditions are presented in Turner (1970) andHanna (1982).Further discussionis provided in Appen-dix D.

d. Terrain.Effectivereleaseheightshouldbe reducedby the terrain height or the releaseheightdivided by 2,whicheveris smaller (Briggs1973).Combined effectsofstabilityand terrain are discussedin AppendixD.

e. Dry Deposition.Dry depositionof particulatemat-ter or gases may be of interest for ground depositionestimates or for reduction of the source term at adownwinddistance.However,depletionof the cloud bydepositionof respirableparticles(<10-~m aerodynamicequivalentdiameter) will not have significanteffectoninhalation doses due to the low deposition velocity ofrespirable particles. Ground deposition is discussedfurther in AppendixD.

5. Meteorology.Two meteorologicalcategories,me-dian and unfavorable, are used to establish a range ofreleasedispersionfrom expectedto extremeconditions.These categoriescorrespond to 50% and 0.5% sectorprobabilitydistribution of z/Qs, respectively.Specificdefinitions of median and unfavorable dispersion fac-tors are provided in AppendixD.

6. PopulationDose. If population dose estimatesareneeded, dispersion factors used to calculate con-

-- servative population doses may be assumed in thedirection having the largest nearby population. Theunfavorable dispersion factor for this sector could be---

. used to calculatethe collectivedose to the peopleresid-ing in that sector (out to some minimum dose contour,such as 25 mrem). If there are population centers inseveral directions from the release, several sectorevaluationsmay be appropriate.

It is not appropriateto considerthe entire populationexposed to the concentration at the cloud centerline.Integratedactivityconcentrationsover the area contain-ing people and the actual population density in thoseareaswouldbe appropriate for populationdose calcula-tions.

F. RadiologicalDose

Choiceof modelsand input parametersare importantifdoseconversionfactorsare to be obtainedand appliedin a reasonablyconsistent manner. Doses to the wholebody and to specific organs are calculated for com-parisonwith the doseguidelinesproposed for Chapter Iof DOE Order 6430.1.This sectiondeals with the mod-els and parameters used to calculate doses from acci-dent-causedexposuresthat are expectedto be the majorcontributors to accident consequences: inhalation ofradioactivematerial during cloud passage,direct radia-tion receivedby immersion in the cloud, and ingestiondose. Dose methods are discussed in greater detail inAppendix E. A comparison of several dispersion anddose codes was made using common input for twopostulatedreleasecases.The results of this comparisonare presentedin AppendixF.

Inhalation Dose. The inhalation model describedby the ICRP Task Group on LungDynamics(TGLD) isincorporated in most major computer codes for calcu-latinginhalationdoses(ICRP 1966).Formulation of theTGLD modelwasbased only on the need for protectionagainst harmful effects of radiation and is notnecessarily an accurate detailed description of thebehavior of inhaled mdionuclides.This model is, how-ever, considered suitable for the purpose of analyzinghypothetical accidents, that is, estimating fractionaldeposition of important radionuclides in the compart-mentsof the respiratorysystemand subsequenttransferto other organs.The TGLD model is intended for usewith particledistributionsthat have an activity medianaerodynamicdiameter (AMAD)between0.2 and 10pmwith geometricstandard deviationsof lessthan 4.5. Forthe unusual distribution having an AMAD greater than20 pm, complete nasopharyngeal deposition can beassumed. The model does not apply to aerosols withAMADsbelow0.1 ~m.

The TGLD model is included in the ICRP 30 dosemodel (ICRP 1979),a major upgradingof the ICRP 2dose model (ICRP 1959).A transition to the ICRP 30model, which accounts for dose to an organ fromgamma and beta emitters deposited in the organ itselfand in nearby organs, is occurring within the DOEcomplex. This transition is expected to continue ascomputing facilitiesacquire the computer codes and asvarious radiation standards such as 10 CFR 20 and

25

DOE Order 5480.lA, Chapter XI, are revised usingICRP 30or similarmodels.In the interim, the following(elaboratedin AppendixE)are providedfor appropriateuse:

. Codesbasedon ICRP 2 dose modelsmaybe modi-fied to use organ massesfrom ICRP referenceman(ICRP 1974) and quality factors from ICRP 26(ICRP 1977).

. Transfer fractions and biological half-times inICRP 30 models may be incorporated in modelswherepossible.

● Volubilityclassesofcompoundsin ICRP 30may beused in all calculations.

. Bone doses calculated for comparison with thesurface bone guideline dose proposed for DOEOrder 6430.1may be calculatedby ICRP 30 modelor by the ICRP 2 model of volumebone dosetimesa distributionfactorof n = 5.

● The exposed person may be considered ICRPreferenceman (ICRP 1974)(seedefinitionsin Seetions 111.B.2.and 111.B.3.).

. The AMAD chosen as input to the TGLD modelrepresents the respirable fraction of the particlecloudrather than the total particlemass.

. A 50-yr dose accumulation time may be used totake advantage of occupational dose conversiontabulations (differencesbetween 50-yr and 70-yrconversionfactorsare minor).

. Dosesto more than one organfrom a singleinhala-tion exposure may be combined into an effectivedose equivalent by a method similar to the ICRP26 method (ICRP 1977).

2. Direct Irradiation from Cloud Immersion. Twomodelsare commonlyused to calculatedirectgamma orbeta dose from cloud immersion: the finite plumemodel or the semi-infinite plume model. The finiteplume model is preferred for most accident analysiscases where a puff release occurs and the lateraldimension of the cloud is limited by unfavorablemete-orology. If a semi-infinite model is used, a “finiteplume” correction factor should be applied to calcula-tion of close-in doses (<10 km) (Strenge 1980).NRCrecommendationscontained in RegulatoryGuide 3.33(NRC 1977B)state that immersion dose to the wholebody should be assumed at the 5-cm tissue depth(Strenge1980).ICRP 30also usesan acceptablemethodof calculatingimmersion dose by the method of Postonand Snyder(Poston 1974).Calculationofdose to skin israrely necessary because its biological significanceisusuallylow compared with that of other organs receiv-ingdose.

3. Ingestion Dose. Ingestion dose as a controllingconsequence of an accidental release is considered a

lower likelihood than either inhalation or immersiondose. However, a major leak to nearby waterwaysshould be considered in some instances. Although un-likelyto approachthe radiation doselimitsproposedforChapter I of DOE Order 6430.1, ingestiondose mightfall in the categoryof “other consequencesto be cm-sidered”discussedin SectionV.G. of the Guide.

Dispersionof radionuclidesinto the aquatic environ-ment can be estimated by the method of RegulatoryGuide 1.113 NRC (1977C). Doses from ingestion ofcontaminated water have been calculated by severalmodels;RegulatoryGuide 1.109(NRC 1977D),Strenge(1980),Huang (1983), and ICRP 30 (ICRP 1979)areexamples.

G. Other Consequencesto be Considered

Potential consequences of radiological accidentsother than dose to individualsmay influencedecisionson site selectionand major design features.The conse-quences of entionmental contamination, populationdose, and public health effectsmay be significantcon-siderations when a comparison must be made amongseveral alternative sites. In the absence of numericalguidance,the conditionsunder whicheach of the conse-quences could be evaluated are discussed, leaving as-sessment of the result to an authority other than theanalyst. Usefulnessof each result will depend on indi-vidual circumstances.

Environmental Contamination. Contaminationby radionuclides from the postulated accident couldcausean economic impact or production loss in excessof that deemed acceptable. Although numerical guid-ance has not been provided in DOE orders or in NRCregulatoryguides, the cost of cleanup or the impact oflost production at nearby facilities may be significantconsiderationsin the siteevaluation and facilitydesign.

The public health consequences of long-term ex-posure to residual accident debris, afier decontamina-tion and return to originaluse,are small compared withdirect inhalation exposure doses and may usually beneglected(seeAppendixG). A possibleexceptionmightariseifradioiodinewerethe major radionuclidereleasedinto the environment by the accident. The cow-milk-infant pathway for radioiodine may cause morerestrictingdoses than the dose from direct inhalationfrom the passingcloud.

At present, consensus has not been reached amonggovernment agencieson appropriate decontaminationlimits for soiland property.Three approachesare com-monly made toward limits: those based on health ef-fects, food chains, and pathway analysis; those on de-tection levels and ALARA concepts; and those which

I--

.-1

I

26

are expedient in terms of cost and political consider-ations. Preferred limits should be based on levels thatwouldgivelittleadditionalhealth risk to the publiconcethe land and propertyare decontaminatedand returnedto normal use. It is assumed that confiscation and

. .- condemnation of private property in lieu of decon-tamination and renovation to original use are notacceptableapproachesto this problem.

--- Current soil remedial action guidelines(DOE 1984)for actinides and common fission products have beenderived for the DOE FUSRAP and SFMP programsbased on earlier work by Healy and the EPA (Healy1971, 1977, 1979A, 1979B;Napier 1982;ORO 1983;Gilbert 1983;EPA 1977;and EPA 1983).The EPA hasrecently published the final remedial-actionguidelinesfor the natural uranium decay series in 40 CFR 192(EPA 1983).The soil remedial-actionderivation meth-ods (pathway analyses) given in ORO 831 and 832(ORO 1983,Gilbert 1983)have been used by the DOEto be the basisfor developingaction guidelinesfor otherradionuclideswhen they are needed.

Several proposed cleanup levels currently exist foractinides.The EPA has suggesteda soil screeninglevelfor plutonium (EPA 1977).This screening level wascalculated by EPA to meet the EPA proposed doseguidance (1 mrad/yr to lung and 3 mrad/yr to bone)with no remedial action necxxsary.Another suggestedlevel is a limit also based on a maximum dose to anyorgan(Healy 1977,1979A,and 1979B).These proposedcriteria are based upon limiting the amount of pluto-nium that could be inhaled or ingested by the generalpublic living or working in areas contaminated withplutonium.

Decontaminationcostestimatesmay be based on theapproach used in WASH 1400(NRC 1975B)and in thePantexEIS(Wenzel1982).This approach includesthreeland use categories(farm, suburban, and commercial),but doesnot refinethe analysisto the extentof includingsite specificationfeaturesbeyond these three categories.Other potentialcosts,suchas decontaminationof onsitebuildings and loss of operating time at contaminatedonsitebuildings,couldbe consideredseparately.

Decontamination methods, decontamination guide-lines,and costestimatesare discussedin greaterdetail inAppendixG.

2.in Section 111.B.-- Calculationof radiologicaldosesto individualmembers

ofa populationisalsodiscussedelsewhere(SectionV.F.and AppendixE).Populationdosecan be valuableas an--

. additional index of the suitability of a site and as anintermediateresultneededin the estimation ofpotentialhealth effectsof a postulatedaccident.

The population which is subject to exposure is com-monlyconsideredthe total populationwithin an 80-km

radius of the proposed site. However, the accidentcaseusually involves only that part of the population con-taining individuals directly contacted by the accidentcloud. That is, a limited number of people could beexposed potentiallyhigh concentrations of airborneradioactive material. In this limited population, ahigher incremental health risk or a higher risk to theindividualreceivingthe averagedosewouldbe expectedto occxrthan in the totalpopulation.Therefore,dilutionof the population dose over a larger population shouldbe avoided when using the population dose in an esti-mation of incremental health effects due to a majoraccident.

Population dose may be calculated using integratedconcentration values and population densities. Cloudcenterlinedosestimes the total population is consideredoverlyconservativefor this estimate.

Assumptions regarding makeup of the exposedpopulation may require variation to suit the actualpopulationand the radionuclidesreleased.If populationdoseis the desiredendpoint, say for comparison amongseveral alternative sites, a homogeneous populationmade up of adults is an acceptableassumption. Errorsresultingtlom assuminga homogeneouspopulationareconsidered relatively minor in comparison withpossibleerrors in other areas of dose calculation(Etnier1979).However,basinga sitingor major designfwturedecision on a homogeneouspopulation dose does notaccount for differenc..s in health risk factors as in-fluencedby ageat the time of exposureand sexdistrib-utionof the population. More study is needed to assessthe error magnitudes when homogeneous populationdoses are used to estimate health effects.In the mean-time, care should be taken to use conservativeage-andsex-averaged health risk factors with homogeneouspopulationdoses.

Credit for a fraction of the population being indoorsduringcloudpassagemay bejustified i.fasuitablemodelhas been devised. The protection factor afforded bybeing indoors is discussedby Cohen (1979).Credit foremergencyplanningand evacuationshouldnot be takenunless the release is delayed beyond the time in whicheffectiveaction couldbe taken.

3.a at a DOE

nuclearfacility.Thesehealth effectsmight includeacuteeffectsbut would usually involve only delayed effectssuchas cancer mortality and perhaps serioushereditaryeffects.Estimatesshouldbe based on health risk factorswhich have been recommended by recognizedadvisorygroups. As discussed in Section V.G.2, health effectsestimatesshouldbe based on conservativeage-and sex-averaged risk factors when using a homogeneous (all-adult) population dose. Two methods of estimating~ealth risks resulting from exposure to low levels of

27

ionizingradiation have been prepared for use in DOENEPAdocuments(Buhl 1984).The first method is usedwhen demographic data are not available for the ex-posed population; in this case, age- and sex-averagedlifetimecancer(or serioushereditarydefect)risk factorsare provided, based on the makeup of the US popula-tion. The second method provides for a more detailedrisk calculationwhen the exposedpopulation is signifi-cantly different from the US population. Use of riskfactorsaveragedby ageand sexover the US populationwould lead to differencesof a factor of 2 to 3 from theexposed populations with more extreme age and sexdistributions(Buhl 1984).Health risk factorswhich areconsideredappropriate for the purposeare discussedingreaterdetail in AppendixB.

Advice on terminology to be used when reportingestimated health effects in safety analysis documentshas varied among peer reviewers of the Guide. Betterpublic understanding or reduced chance of mis-interpretation of the data by the reporting media or bymembers of the public is the objectivewhen reportingpotential health effects. It is difficult to put any non-voluntary health risk into a more positive or favorablelight. The negative as well as the positive aspects ofDOE activities should be reported objectivelywithoutshadingmeaningor 10SSof accuracy.

A number of semitechnicalterms have been used inthe past to describehealth risk.The followingare exam-plesof terms that can be usedunder appropriatecircum-stances:delayed cancer mortality, incremental risk ofeventual cancer death, expected cancer deaths, pro-jected cancer deaths, potential health effects, and per-hapsothers.The term used shouldacknowledgethe factthat radiation can cause cancer (or potentially serioushereditarydefects)but at low doses to individuals, thislikelihood is quite low. Increased chance of eventualcancer death and incremental risk of eventual cancerdeath are similar terms which accurately describe thesituationofan individualwho receivesa doseas a resultof a postulated accident. This dose representsa risk ofcancer, not a certainty. It is also considered helpful tothe understanding of this increased risk to include in-creased risks from common activities of the public forpurposesof comparison.

Points to be made in providinghealth effectinforma-tion are (1) a finite but small probabilityexists that anaccident might occur which could have offsite conse-quences; (2) should the accident occur, the analysisshowssome members of the publiccould receiveradia-tion doses;(3) it is not clear whether any of these doseswould be large enough to cause observable health ef-fects-it is known that theseeffectswould probablynotbe observablefor many years;(4)thesehealth effectsarepurposely overestimated in the analysis to assure thatany error in the estimate is accounted fo~ and (5) theseestimates are made to promote more informed deci-

sions regardingthe location of the nuclear facilityandthe major design f~tures included to minimize theeffectsofany potentialaccident.

‘- -

ACKNOWLEDGMENTS

This has been a project of many participants. Theauthors are indebted to the 14 peer reviewers and ap-proximately30 other interested peoplewho spent timeand effortreviewingand contributingto the Guide. Wealso wish to thank Raymond Cooperstein of the DOEOfficeof NuclearSafetyfor his contributionsas projectmonitor.

In particular, we acknowledge the extra effort ofJames Denovan ofBattellePacificNorthwestLaborato-ries, William Durant of Savannah River Laboratory,and Douglas Wenzel of WestinghouseIdaho Nuclear,who met with the authors and Raymond Coopersteintoaid in reconciling comments of the reviewers. Theauthors also appreciate special assistance provided byJack Healy and Joel Bennett of Los Alamos, MiltonBullockof EG&G Idaho, and David Coatsand WilliamKingof Lamence LivermoreNational Laboratory.

Finally, our appreciation goes to our faithful wordprocessors:CherylLemons,VickieMartinez,and Rose-mary Guenther, without whosediligencethe many ver-sions of the guide would have been unmanageable.Thanksalsoto technicaleditor VirginiaReicheltfor hertimelyand thorougheffort.

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Mishima 1966:J. Mishima, “Plutonium ReleaseStudies,” Battelle Northwest Laboratories reportBNWL-357(November 1966).

Mishima 1970: J. Mishima and L. C. Schwendiman,“The Amount and Characteristicsof Plutonium MadeAirborne Under Thermal Stress,” Pacific NorthwestLaboratoriesreport BNWL-SA3379(October 1970).

Mishima 1979: J. Mishima, L.C. Schwendiman,and J.E. Ayer, “Estimated Airborne Release of Plutoniumfrom WestinghouseCheswickSite as a Result of Postu-lated Damagefrom SevereWind and SeismicHazard,”BattellePacificNorthwest Laboratory report PNL2965(May 1979).

3

Mishima 1980: J. Mishima,L. C. Schwendiman,and J.E. Ayer, “Estimated Airborne Release of Plutoniumfrom the Exxon Nuclear Mixed Oxide Fuel Plant atRichland,Washington,as a Result of Postulated Dam-age from Severe Wind and Earthquake Hazard,” Bat-tel.lePacific Northwest Laboratory report PNL3340(February 1980).

Mishima 1981: J. Mishima and J. E. Ayer, “EstimatedAirborne Releaseof Plutonium from Atomics Intern-ationalNuclear Material Development Facility in theSanta Susana Site, California,as a Result of PostulatedDamage from Severe Wind and Earthquake Hazard,”BattellePacificNorthwestLaboratoryreport PNL3935(August1981).

Napier 1982: B.A. Napier, “A Method for DeterminingAllowable Residual Contamination Levels of Radio-nuclideMixturesin Soil,” PacificNorthwest Laborato-ries report PNL3852 (May 1982).

Newmark 1973: N. M. Newmark and W. J. Hall,“Proceduresand Criteria for Earthquake Resistant De-sign,” in Building Practicesfor Disaster Mitigations,National Bureau of Standards Building ScienceSeries46 (1973),p. 209.

Newmark 1978: N. M. Newmark and W. J. Hall, “De-velopment of Criteria for Seismic Review of SelectedNuclear Power Plants,” US Nuclear RegulatoryCom-missionreport NUREG/CR-O098(May 1978).

NRC 1974: “On-site Meteorological Programs,” USNuclear Regulatory Commission Regulatory Guide1.23(1974),p. 6.

NRC 1975A:“General Site SuitabilityCriteria for Nu-clear Power Stations,” US Nuclear Regulatory Com-missionRegulatoryGuide 4.7 (November 1975).

NRC 1975B:“Calculationof Reactor AccidentConse-quences,” Reactor Safety Study, US Nuclear Regula-tory Commission report WASH-1400 (NUREG75/014)(October 1975),App. VI.

NRC 1977A:“Site SuitabilityReport in the Matter ofClinchRiverBreederReactorPlant,” US NuclearRegu-latoryCommissionreportNUREG-0786(March 1977).

NRC 1977B:“Assumptions Used for Evaluating thePotentialRadiologicalConsequencesof AccidentalNu-clear Criticality in a Fuel Reprocessing Plant,” USNuclear Regulatory Commission Regulatory Guide3.33(April 1977).

NRC 1977C:“Estimating Aquatic Dispersion of Ef-fluents from Accidentaland Routine Reactor Releasesfor the Purpose of Implementing Appendix I,” USNuclear Regulatory Commission Regulatory Guide1.113(April 1977).

NRC 1977D: “Calculation of Annual Doses to Manfrom Routine Releases of Reactor Etlluents for thePurpose of Evaluating Compliancewith 10.CFR, Part50, Appendix I,” US Nuclear RegulatoryCommissionRegulatoryGuide 1.109(October 1977).

NRC 1978A:“Developmentof Floor DesignResponseSpectra for Seismic Design of Floor-SupportedEquip-ment or Components,” US Nuclear Regulatory Com-missionRegulatoryGuide 1.122(February 1978).

NRC 1978B:“Design, Testing and MaintenanceCriteria for Post Accident Engineered-Safety-FeatureAtmosphereCleanupSystemAir Filtration and Absorp-tion Units of Light-Water-Cooled Nuclear PowerPlants,” US Nuclear Regulatory Commission Regula-tory Guide 1.52(March 1978).

NRC 1979A:“Assumptions Used for Evaluating thePotentialRadiologicalConsequencesof AccidentalNu-clear Criticalityin a Uranium Fuel Fabrication Plant,”US Nuclear RegulatoryCommission RegulatoryGuide3.34(July 1979).

NRC 1979B:“Assumptions Used for Evaluating thePotentialRadiologicalConsequenmsof AccidentalNu-clear Criticality in a Plutonium Processing and FuelFabrication Plant,” US Nuclear Regulatory Com-missionRegulatoryGuide 3.35(July 1979).

NRC 1981A:“Standard ReviewPlan for the ReviewofSafetyAnalysisReports for Nuclear Power Plants,” USNuclear RegulatoryCommission report NUREG-0800(Ju1y1981).

NRC 1981B:“Technical Bases for Estimating FissionProductBehaviorDuringLWRAccidents,”US NuclearRegulatory Commission report NUREG-0772 (June1981).

NRC 1983: “AGuideto the Performanceof Probabilis-tic Risk Assessments for Nuclear Power Plants,” USNuclear Regulatory Commission reportNUREG/CR-2300(1983).

b

--

--.

-.-

ORNL 1970: “Siting of Fuel ReprocessingPlants andWaste Management Facilities,” Oak Ridge NationalLaboratoryreport ORNL4451 (July 1970).

32

ORO 1983: “RadiologicalGuidelines for Applicationto DOE’s Formerly Utilized Sites Remedial ActionProgram,” Oak Ridge Operations Office reportORO-831(March 1983).

.&

ORO 1984: “SafetyAnalysisand ReviewSystem,”OakRidge Operations Office Order OR-5481.lB (May 23,

.,’ 1984).

Orth 1980: D.A. Orth, G. H. Sykes,and J. M. McKib-ben, “Performance of Sand Filters for the SeparationAreasat Savannah River Plant,” in “Proceedingsof the16thERDA Nuclear Air CleaningConference,”CONF801038(1980),p. 745.

Pepper 1978: D. W. Pepper,“Calculationof ParticulateDispersionin a Design BasisTomadic Storm from theBabcockand Wilcox Plant, Leechburg,Pennsylvania,”US Departmentof Energyreport DP-1487(E. I. du Pentde Nemours and Co., Aiken, South Carolina, March1978).

Perkins 1980: W. C. Perkins, W. S. Durant, and A. H.Dexter, “Potential Safety-Related Incidents withPossibleApplicabilityto a Nuclear Fuel ReprocessingPlant,” US Department of Energyreport DP-I 558(E. I.du Pent de Nemours and Co., Aiken, South Carolina,December 1980).

Perkins 1981: W. C. Perkins and A. H. Dexter, “APreliminaryHazards Analysisof a BreederFuel Repro-cessing Facility,” US Department of Energy reportDP-1584, (E. I. du Pent de Nemours and Co., Aiken,South Carolina,October 1981).

Poston 1974: J. W. Poston and W. S. Snyder,“A Modelfor Exposure to a Semi-Infinite Cloud of a PhotonEmitter,” HealthPhysics26,287-293 (April 1974).

RFP 1982: “Safety Analysesand Risk Assessment inSupport of the Long Range Rocky Flats UtilizationStudy,”RockwellInternational report SAR-279(232)-3(November 1982).

Selby 1975: J. M. Selby,E. C. Watson, J. P. Corley,etal., “Considerations in the Assessment of the Conse-quences of Eflluents from Mixed Oxide Fuel Fabrica--. tion Plants,” Battelle Pacific Northwest Laboratories,*report BNWL1697 (June 1975).

.--. Slade1968: D. H. Slade,Ed, “Meteorologyand Atomic

Energy,” US Atomic Energy Commission reportTID-24190(July 1968).

Stratton 1967: W. R. Stratton, “Review of CriticalityAccidents,” Los Alamos Scientific Laboratory reportLA-3611(January 1967).

Strenge1980: D. L. Strenge,“Models Selectedfor Cal-culation of Doses, Health Effectsand Economic Costsdue to AccidentalRadionuclideReleasesfrom NuclearPower Plants,” US Nuclear Regulatory CommissionreportNUREG/CR-1021 (PNL3108) (May 1980).

Swain 1983: A. D. Swain and H. E. Guttman, “Hand-book of Human ReliabilityAnalysiswith Emphasis onNuclear Power Plant Applications,”US Nuclear Regu-latory Commission report NUREG/CR-1278 (June1983).

Tera 1984: “Final Report: SeismicHazard Analysisforthe 13endix,Los Alamos, Mound, Pantex, Rocky Flats,Sandia-Albuquerque,Sandia-Livermore, and PinellasSites,”Tera Corporationreport B-82-261(March 1984).

Tuck 1974: G. TuclG“SinlplifiedMethods of Estimat-ing the Results of Accidental Solution Excursions,”NuclearTechnology23 (August1974).

Turner 1970: D. B. Turner, “Workbook of At-mospheric Dispersion Estimates,” US Department ofHealth, Education, and Welfare report PB 191482(1970).

Velhmzi 1973: J. W. Vellozzi and J. J. Healey,“Procedures and Criteria for Wind Resistant Design,”in BuildingPracticesfor DisasterMitigation,NationalBureau of Standards BuildingScienceSeries46 (1973),p. 237.

Walker 1978: E. Walker, ‘“ASummary of ParametersAffecting the Release and Transport of RadioactiveMaterial from an Unplanned Incident,” Bechtel Na-tional report BNFO-81-2(September 1978).

Wenzel 1982: W. J. Wenzeland A. F. Gallegos,“Sup-plementary Documentation for an Environmental Im-pact Statement Regarding the Pantex Plant: Decon-tamination Methods and Cost Estimates for PostulatedAccidents,” Los Alamos National Laboratory reportLA-9445-PNTX-N(1982).

Woodcock 1966: E. R. Woodcock, “Potential Magni-tude ofCriticalityAccidents,”United KingdomAtomicEnergyAuthority report AHSB(RP)R-14(1966).

33

I

A

Absoluterisk. Expressionof excessrisk due to a radia-tion exposure as the arithmetic differencebetween therisk among those exposed and risk in the absence ofexposure.

Absorbeddose.The energyimparted to matter by ioniz-ing radiation per unit mass of irradiated material at theplace of interest. Its unit is the rad or gray (Gy) (seeMetricunits).

Accumulateddose. A dose term coined for use in theGuide to fit the postulated case of dose equivalentaccumulatedover a time interval after a singleacciden-tal intake of long-livedradionuclides.No new sourceofdose is added during the interval. The accumulationinterval should be specified;50 yr is recommended, forthe reasonsdescribedin AppendixE, SectionI.

Activity median aerodynamicdiameter (AMAD).Theaerodynamicequivalentdiameterbelowor abovewhichhay of the activity of an aerosol with lognormallydistributedparticlediameters is associated.

Aerodynamicequivalentdiameter(D.c).The diameterofa unitdensity spherethat wouldhave the same terminalvelocity due to gravity in air as the particle underconsideration.

AnnualLimit onZntake(ALI).An ICRP secondarylimiton occupational exposure, as intake in 1 yr of radio-nuclides resulting in stochastic eflects in all organsequivalentto 0.05Sv (5rem) or nonstochasticeflectsin asingleorganequivalent to 0.5 Sv (50 rem).

AAKZ.American National Standards Institute.

As lowas reasonablyachievable(ALARA).The criterionstated in 10 CFR 20 that exposures of personnel toradiation during routine operation (of LWRS) will be“as lowas reasonablyachievable.”

BEIR. BiologicalE#ects of Ionizing Radiation (Na-tional Academyof SciencesCommittee on. . .).

Breathingrate.The volumetric rate of air exchangebythe respiratorysystem;the product of tidal volumeandrespirationrate.The ICRP has establishedthe followingstandard breathingrates for referenceman:

Resting 1.25X 10-4m3/sLightactivity 3.33X 10-4m3/sHeavywork 7.17X 10-4m3/sHeavy exercise 18.$X 10-4m3/s

Committed dose or dose commitment. The radiationdose calculated for radiation protection purposes toevaluate the dose received during some period of ex-posureplusthe doseaccumulatedover a period ofyears,say50yrofoccupationalexposure,resultingfrom radio-nuclidesdepositedwithin the body during the exposureperiod.

Conj?nement.Usually refers to a system of cladding,containers, piping, glove boxes, other barriers, and aircleaningequipment which prevent release of radioac-tive material into occupied spaces. Primary confine-ment refersto the firstbarrier providedfor this purpose.

Containment.Usually refers to a structure capable ofcontaining (with some nominal leakage) an over-pressure caused by explosion or release of pressurizedcontentsof vessels.

Credibleevent. An event whose probability of occur-rence is above a specifiedthreshold (recommended inthe Guide to be greaterthan 10+ per year).

Criticalfacility. A nuclearfacility for radioactivemate-rial handling processing or storage,and other facilitieshaving vital importance to DOE programs or highdollar value, such as plutonium processing, tritiumprocessing, weapon assembly (Pu/HE), and certainstoragefacilities.

Criticalsystem.A systemwhosecontinued integrityandoperation are essential to assure confinement ormeasurethe releaseof radioactivematerials in the eventof a DBA. Usually the ventilation, fire detection andsuppression,electrical,and utility systems.

Criticalityaccident.The accidental assembly of suffi-cient fissionable material to initiate a self-sustainingneutron chain reaction. The resulting neutron burst, ifunshielded, is a major hazard to nearby workers. Theenergy produced disperses fission products which cancausepotentialhealth effectsonsiteand offsite.

-.

--

Wermsin italicsaredefinedelsewherein thisglossary.

34

Damping. Dissipation of energy by motion within astructure or in underlyinggeologicformations. Usuallyan index of the ability of a structure to withstandvibratorydamage.

DDecontaminationfactor. The quantity of radionuclideper unit area before decontamination divided by the

●quantity remaining after decontamination. Also the-.ratio of upstream to downstream concentrationappliedto removalcapabilityof fluidcleaningsystems.

Design basis accident (DBA). See definition and dis-cussionin Section111.B.1.

Dilutionfactor. The ratio of concentration of radio-nuclides(Ci/cm3)in samplesof standard volume takenbeforeand after dilution of contaminated material by alargervolumeof a medium.

Dispersibleform. The form of radioactive materialwhich makes it subject to airborne or waterborne dis-persion by the dispersive energy at hand; at least afraction of the dispersed material will be directlyrespirable (see respirabiefraction) or subject to con-version to a respirableor ingestibleform by exposureinthe environment.

Dispersion.The processof natural mixing of a materialreleasedto the atmosphere with air, causinga reductionin concentration with distance from the source. Seemedian dispersionfactor and unfavorabledispersionfactor.

Dispersionfactor x/Q (s/m3).Ratio of the air concentra-tion (g/m3or Ci/m3)and the releaserate (g/sor Ci/s) orthe ratio of the time-integratedconcentration (gos/m3or Ci” s/m3) and the total quantity released (g or Ci).Dispersion factor yields dose when multiplied by anamount released,a breathingrate,and a doseconversionfactor.

Dose accumulationtime. A time period over whichexpected dose from a long-livedradionuclide retainedin the body after a singleaccidental intake is estimated(seeaccumulateddose).

.Dose conversionfactor. A factor with units of doseequivalentper unit activity inhaled or ingestedwhich is.%multipliedby other factorsto obtain the doseequivalentreceivedbya specificorgan(seeAppendixE,SectionII)..-

Dose equivalent.A quantity that expressesall kinds ofradiation on a common scale for calculatingthe effec-tive absorbed dose;defined as the product of absorbeddose in rad (or Gy) and modifying factors such asquality factor or an organ distribution factor. The unitof doseequivalentis rem (or Sv).

EJ1.2ctivedose equivalent.See definition and discussionin Section111.B.

E~eBctivereleaseheight.Heightaboveground at whicharelease of airborne radionuclides is assumed to occur.Usually stack height plus adjustments for buoyant ormomentum plume rise and any terrain effect (see Ap-pendixD, SectionHI).

Elevatedrelease.A point sourcereleaseoccurringaboveground level(M-3 m). Usually refers to a releaseat theeflectivereleaseheight.

Engineeredsafetyfeature (ESF).Any feature of a nu-clearfacility, includingstructures,systems,and compo-nents, provided to prevent or mitigate the accidentalreleaseof radioactive materials from the facility.Typi-cal 13SFsare containmentstructures, confinementbar-riers, air cleaning systems (filters, absorbers, traps,scrubbers), devoted emergency cleanup systems, fireprotectionsystems,and safetysystems.

EZS.Environmental Impzct Statement.

EPA.EnvironmentalProtectionAgency.

Event tree analysis. An inductive analysis whichportrays the various paths or scenarios that may resultin a major consequence when some initiating eventdri}’esa systemout of its standard operatingmode.

Facility bounda~’.The boundary, usually a fence orother physicalbarrier, provided for the security of thefacility.Some facilityboundariesprovide a radiologicalaccidentexclusionarea.

Faulttree analysis.A deductive failure analysis whichfocuseson one undesired event and provides a system-atic method for determining causes of this event. Theundesired event constitutes the top event in a fault treediagramand may be a catastrophicfailure.

-.

35

Fumigation. An atmospheric inversion conditionprevalent just after dawn in which newly developedconvective eddies mix the eflluent plume within theshallow unstable layer next to the ground. This con-dition can cause the greatest ground-level concentra-tions observed in the neighborhood of a stack overperiodsof about 30 min to 1h.

Geneticeffect.Serious health effects in future genera-tions resulting from a radiation dose to either parent,usuallyautosomal dominant and x-linked,irregular in-herited, recessive,and chromosomalaberrations.

HEPA falter. A high-efilciency particulate air filter,usuallycapable of 99.97%efficiencyas measured by astandard photometricdynamic equivalent(DOP).

ICRP. InternationalProtection.

test using 0.3-pm droplets (aero-diameter) of dioctylphthalate

Commission on Radiological

Incrementalrisk. A risk added to existingor acceptedrisk by d proposednew activity.

ZNEL.Idaho National EngineeringLaboratory.

Inversion. A meteorologicalwhen ~emperature increasesmosphere. Characteristically,blocksnormal plume rise.

condition which existswith altitude in the at-a layer is formed which

Life table. A statistical determination of age-specificprobabilities of death from all causes among variouspopulation groups. Associated survival and longevitydata are included. Life tables may be used to estimatethe number of radiation-induced cancer deaths that a-wouldresult from accidentalexposureofa population. ‘-

-

~wpopulation zone.The area (immediatelysurround- ‘u-ing an exclusionarea) which contains members of the -public, the total number and density of which are suchthat there is reasonable probability that appropriateprotectivemeasurescouldbe taken in their behalfin theevent of a seriousaccident.

LWR. Light-waterreactor, either pressurized-waterre-actor or boiling-waterreactor.

Mediandispersionfactor.The dispersion factor ((~Q))which is exceededby 50%of the hourly ~Qs observedin the sector and at the distance to the person whosedoseis to be calculated.

Metricton.One thousand kilogramsor 2205lb.

Metricunits.The metric systemor SystemInternational(S1)units are recommended for voluntary adoption inDOE Order 6430.1. The radiological units and con-versionfactorsare as stated below.

Mitigation.Minimizingthe effectof a postulated acci-dent by means of facility siting or its major designfeatures.

R 1 Ckg-l=X R

1 =

1 =

1 =X

--..

36

Nonnucleardetonation or single-pointdetonation. Achemical reaction within the high-explosivecompo-nentsof a nuclearweapon,whichresultsin an explosionthat can disperse radioactive materials in the weapon

0 componentbut with lessnuclear yield than the equiva-lent of 4 lb of TNT (approximately2.5 X 1017fissioninplutonium).a

--Nonstochasticefects. A radiation effectwhose seventyin an individual is a functionof dose.

NRC.NuclearRegulatoryCommission.

Nucleardetonation.An energyreleasethrougha nuclearprocessthat isequivalentto the detonation of more thantheequivalentof4 lb ofTNT withina fewmicroseconds(approximately2.5 X 1017fissionsin plutonium).

Nuclearfacility.Seedefinition in SectionI.C.

Oflsite.Any location beyond the site boundary where amember of the publiccan be legallysituatedbeyond thecontrolof the ownerand operatorof the nuclearfacility.Relateddetailsare discussedin Section111.B.2.

Oflsiteperson.See definition and discussionin Section111.B.2.

Onsite.Any location inside the site boundary but notwithin the facilityunder evaluation.

Populationcenterdistance.Distance from the nuclearfacility structure to the nearest population center ofgreaterthan 25000 inhabitants (seeSection111.B.5).

Populationdose. See definition and discussion in Sec-tion 111.B.4.

PRA.Probabilisticrisk assessment.

Protectiveresponserecommendation(PRR).A projectednumerical radiation dose to individuals in the popula-tion that may triggera protectiveresponseby emergencyresponseagencies.

Recurrencetime or return period. A statistically de-termined time period after whichnatural phenomena or.- other events of a particular seventy would be expected---to be repeated, based on historical records sup-

.- plementedin some casesby expertopinion.--

Reductionfactors.A factor by which the releasedradio-nuclidesare reduced by natural means (plateout,gravi-tationaldeposition,absorption,etc.).SeeSectionV.C.

Relativerisk.Expressionof risk due to an exposuretoradiation as the ratio of the risk amongthoseexposedtothe risk in the absenceof exposure.Relative risk projec-tion method assumes the rate of fiture cancermortalitiesdue to radiation dose is proportional to therate of natural that wouldoccur if noraciation dosehad

Releasefraction.That fractionof total radioactivemate-rial

dispersibleform by apostulatedaccident.SeeSection

Removalfactor. A factor by which the released radio-nuclidesare removed by engineeredmeans (absorptionbeds, falters,sprays,etc.). SeeSectionV.C.

Respirablefraction.aerodynamicequivalentdiameteris less

10pm.

Respirationrate. The rate at which a fill respiratorycycletakesplace(inhalationand exhalation).

Risk, risk assessment.See definitionand discussioninSection111.B.

Safi*tysystem. systemssuchas detectionsystems,isolation valves and dampers, annunciators, and otherautomatic systems required to achieve a high level ofsafety in normal operations or safe shutdown in theevent of an accident.

Site bounda~. Usually the boundary of the propertyover which the owner or operator can exercise strictcontrol without the aid of outside authorities. The siteboundary does not have to be a fenceor other physicalbarrier. Seediscussionin Sections111.B.2and 111.B.5.

Volubilityckzss.One of three classesof biologicalhalf-times (tB)established by the ICRP to indicate rate ofclearanceof radionuclidesfrom the pulmonary regionofthe lungs:ClassD, 1to 10days;ClassW, 10to 100days;and ClassY, more than 100days.

Somatic eflects.Harmful effectsof an agent (such as aradioactive material) within the body of the person oranimal receiving dose, rather than in offspring of thereceptor.

Sourceterm.The amount of radioactive material avail-able for releaseafter the releasefraction from prima~confinementis applied; it is therefore the amount ofradioactivematerial releasedfromprimaryconfinement

dispersib[eform.

37

SRP.Savannah River Plant.

Stochastic eflects. A radiation effect where theprobability of occurrence, rather than severity, is adirect fimction of dose; a radiation effeet occurringwithout threshold, such as hereditary effeets andcarcinogenesis.

Tidal Vohone.each respiratorycycle.

UBC.Uniform BuildingCode.9.

Unfavorabledispersionfactor. The dispersion factor ‘-(x,/Q)whichis exceeded by 0.5% of the hourly ~Qs ~-observedin the sectorand at the distance to the person -.-whosedose is to be calculated.

_--,,

----

38

B

.. I

Radiation health effects data were reviewed withthese objectives:to provide recommended health riskfactorsfor considerationof publichealth effects,and toreviewthe method for calculatingeffectivedoseequiva-lentwheremore than oneorganreceivessignificantdosefrom a singleintake of radioactivematerial.

Appropriate health risk factors are in debate withinthe radiation health and epidemiologicalcommunitiesand major uncertainties exist. However, public healtheffectsfrom routineoperationand potentialaccidentsinnuclearfacilitieshave come under scrutinyand deserveconsideration.

Whetherhealth riskestimatesneed to be reportedwillprobably depend on the magnitude of the dose. If theestimated dose is lower than annual backgroundradia-tion, it wouldcause little additionalhealth risk. Report-ing trivial riskswouldbe of littlevalue. However,dosesto individualsin the rangebetweenbackgroundand the25-rem siting guidelinedose could cause health effectswhichmay warrant consideration.

Portions of this discussion can be found in greaterdetail in a Los Alamos report (Buhl 1984).This reportprovidesmethodsof estimatingradiation health risk foruse in National Environmental Policy Act (NEPA)documentspreparedby DOE.

II. R

Estimates of the increased risk of cancer mortalityresultingfrom exposureto ionizingradiation have beenpublishedby

the Committeeon the BiologicalEffectsof IonizingRadiations (BEIR Committee) of the US NationalAcademy of Sciencesin 1972(the BEIR I report,BEIR 1972)and in 1980(the BEIRHI repoz BEIR1980);the United National ScientificCommittee on theEffectsof Atomic Radiation (the UNSCEARCom-mittee, UNSCEAR 1977,1982);andthe International Commission on RadiologicalProtection (ICRP 1977).

DifYerentapproacheswere taken by these organizationsin presentingtheir risk estimates.Both UNSCEARandthe ICRP publishedage-averagedand sex-averagedriskcoefficients,giving the incremental lifetime risk of anindividual dying of a radiation-induced cancer eitherper unit absorbed dose (UNSCEAR) or per unit doseequivalent(ICRP).The BEIR Committeeson the otherhand, tended to publish age-and sex-specificrisk rates,givingthe annual risk of dyingof cancer in terms of ageat exposureand elapsedtime sincethe exposure.

The first approach has the advantageof simplicity.Ifthe cumulativeorgan dose to a population in an assess-ment area isgiven,the number ofhealth effectsresultingfrom that dose is easily estimated by multiplying thecumulativeorgan dose by the risk factor for that organ. JHowever, if the population at risk were significa~tlydifferentfromthe populationoverwhichthe risk factorshad been averaged(for example, if the population con-sisled of male radiation workers between ages 20-25),the estimate of health effects using an age- and sex-averagedrisk factormay not be representative.

This difficulty is remedied by using the second ap-proach,whichemploysrisk-ratecoefficientsfor each sexand age group. The enhanced flexibility in this ap-proach, however, is offset by an increased complexity.Input data required to perform this health effectscalcu-lation include the population distribution by age andsex, life table for each sex, and if a relative risk projec-tion model is used, cancer mortality rates by age andsex.

The risk estimates from BEIR III age- and sex-averaged lifetime risk factors were calculated from theBEIR III risk-rate factors (when lifetime risk factorswere not given).These estimates are listed in Table B-Ifor the most important organsof concern. In obtainingthe BEIR III lifetime risk factors, we used a life-tablecalculationwith the 1980US populationdistributionbyageand sex(US Bureauof the Census 1982)and the USdecennial life tables (US National Center for HealthStatistics,1975).

The BEIR 111lifetime risk estimates were calculatedusingthe linear (L)dose responsecurve, whichassumesthat the cancerrisk increaseslinearlywith dose (forbothlow-LET and high-LET radiation); the quadratic (Q),which assumes a quadratic model to provide a lowerbound (low-LETradiation only); and the linearquad-ratic (LQ) (low-LETradiation only), which is an inter-mediate estimate. The BEIR III Committee recom-mended that a linearquadratic model be used for low-LET radiation. However, the linear finction may be

39

preferable in view of the recent reassessment of thedoses at Hiroshima-Nagasaki,which has resulted in areduction of the estimated neutron flux, especiallyatHiroshima. The linear dose response function is lessaffected by changes in the neutron relative biologicaleffectivenessand neutron flux and at this time wouldappear to conservatively estimate the risk of cancerinductionby radiation (Buhl 1984).

For completeness,the risk of seriousgeneticdisorderin all subsequentgenerations that may result from ex-posure to ionizing radiation has also been included inTable B-I. This risk was taken as the equilibrium riskfrom the UNSCEAR and BEIR III reports, since aspointed out in BEIRHI, “the total of all serious*geneticeffectsthat willbe expressedover all future generationsas a consequenceof exposurelimited to a singlegenera-tion, is numericallyequal to the total foreachgenerationin the equilibrium situation” (BEIR 1980).The geneticrisk estimator recommended by the ICRP was adopted(ICRP 1977),in which the risk of serioushereditary illhealth in the first two generationswas estimated to be100 X 10+/rem and of the same magnitude in latergenerations.The total risk was taken by the ICRP to be200X 10-6/rem.

Risk estimates applied to postulated accident casesshould be derived from similar exposure modes wherepossible. Similarity in radiation type, exposurepathway, dose rate, population makeup, and clearancetime would be ideal but is seldom available. Whatchoices should be made when similarity is lacking,orwhether differencescan be tolerated, is a major ques-tion.

The major points to be made (each considered ap-plicableto healtheffectsconsiderationsin accidentanal-yses)are as follows:

● Primary reference.The BEIR 111report can berelied upon heavily in arriving at recommendedrisk factors,primarily because the BEIR III reportis the most recen~ thereby benefiting from inputfrom the earlier reports and from experimentaldata reported in the meantime. Extensionsof theBEIR 111report results were made to broaden itsareas of applicability.

● Populationcomparability.Use of age- and sex-averaged cancer risk factors is recommended forestimatinghealth risk in a populationsimilar to the1980US population (or a population of unknownmakeup, but probably similar to the 1980 USpopulation).These factorsare basicallythe sameasthoselistedin Table B-I.Aa more detailedcalcula-

tion might be preferred for estimating health riskfrom an exposure of a population much differentthan the US population (that is, a nearby workforcecomposedprimarilyof men aged25 to 30 yr).Acomputercode isavailablefor this purpose(Buhl1984).Use of risk factorsaveragedby age and sexover the US population may lead to differencesofup to a factor of 2 or 3 from a similar treatment ofexposedpopulationwith more extremeageand sexdistributions. The analyst would evaluate thesignificanceof this uncertaintyand decidewhethera more detailedcalculationis appropriate.

● Reportinghealthrisk. When results of health riskcalculationsare expressed,many optionsare avail-able. As shown in Table B-I, the BEIR 111reportexpressedorgan cancer risks from low-LETradia-tion in terms of linear, linearquadratic, and quad-ratic dose responseequations,each modifiedby anabsolute or relative risk projection model. Otherthan an indicationof the lowerlimit of uncertaintyin risk estimation, the quadratic calculation is notconsidered to acceptably tit dose response data.The linearquadratic form, althoughrecommendedby the BEIR HI Committee as a preferred centralvalue, has recently come into question followingreview of Hiroshima-Nagasakigamma and neu-tron doses.Until the ensuingrecalculationis com-pletedand acceptedby radiation protectionbodies,use of a model which conservativelyestimates riskis advisable.The linear model is unanimouslycon-sidered by UNSCEAR, BEIR HI, and ICRP tooverestimate low-LET radiation risk. For mostcases,risk resultscalculatedby the linear modelareconsideredalso suitablefor useas an upper limit ofuncertainty. The range bounded by the absoluterisk projection and the relative risk projection ofthe linear estimate are considered appropriate forexpressingrisk results of accident analyses (Buhl1984).

. Doserateand levelconsiderations.The NCRP rec-ommendsa reductionof linearlyextrapolatedtotalcancer risk from whole-bodylow-LETradiation, ifthe exposureis at lowdosesand lowdose rate. Thisallowancecompensates for the effect of biologicalrepair mechanisms.Reducingthe risk factor in theall-cancer category for whole-body radiation (inTable B-I)by a factor of two would be appropriateif the dose rate were less than 5 rad/year and thedose levelwerelessthan 20 rad.

●

✎

.--

.

--4

----

*Autosomaldominantand x-linked;irregularinherited,re-cessive,andchromosomalaberrations(BEIR1980).

40

R D A R ACANCER(CancerDen!hs/lObPerson-Rad)

@

.

H R ( r

B

A R A RU i R R U

C 501 (L) — — —(

—( 7 226(LQ) —

R— — —

I 2

B —S

— —5 - 1 - 2

L-

2L

21

8 2<

T—

5 5 —L

—1

—- S

(RedMarrow) ——

— - 2 — — —— — ( — — —

GD 1 6 1

L= l d r = I iQ =

‘The LL modelWISusedin nrskinBthese riskestimstes.bAqnalityfsrtorof 20 h- beenassumed.Wfrebreastcsnccrrisk forwomenhss beenredumdby 93%forthe genersipopulation.‘Thr BEIRIII reportlists ~ dose-squsredexpnnentidfrmtionmrds ikrearfrsctiontoexpressthedosrrespmuc reistionforboneam-ers. FormnvenIen~ onlythellnar frsctionis givenhere.7he RBEofdphsrndistion forirmganmr Is 8-15(BEIR 1980,P.327).‘Glrulstcd fmmtheThorotmstpatients”tits.gNumberofserlousdisordersin sll subsequentgerrcrstionw%leqrmted riskfnctorof149X 104geneticdisordersperr~dis takenfmmUNSCEAR(1982).This vsluewrpersedssthepmvlonsvslneof 185X 104/rsd gl-en inUNSCEAR(1977).

Dosescalculatedfor potential accidentscan be com-pared with guideline doses existing for individual or-gans in a straightforward manner, unless significantdose is calculatedfor more than one organ. In this case,a method is needed to considerthe contribution of eachorgan dose to possible delayed health effects. Thepurpose of this appendix is to review existing healtheffect data (Table B-I) and suggest a method whichcould be used to evaluate a quantity equivalent to asingle-organdose. For the purposes of the Guide, thedesired quantity is called effectivedose equivalent andis related to the whole-bodydose. The effective doseequivalent may be compared with the 25-rem whole-body dose limit proposed for DOE Order 6430.1,Chapter I. The effectivedoseequivalentmaybe derivedby establishing a risk equivalence between multiple-organ dosesand a singlewhole-bodydose. Establishing

-- this riskequivalencewassuggestedearlierby Strom and.-. Watson (Strom 1975)and by the ICRP (ICRP 1977).

The ICRP states in ICRP 26 that the risk of delayed-. mortality shouldbe treated the same whether the whole-. body is irradiated uniformly or whether several organs

receivethe dose. Derivation of organ weightingfactors

from health effectsdata availablein 1976-1977allowedICRPto calculatean effectivedoseequivalent.Asimilarmethod for calculating weighting factors is recom-mended, although not necessarilyusing the ICRP 26organ risk factors.The finalized risk factors calculatedfrom BEIR 111 (Table B-I) after review byepidemiologistsand health scientists will probably beadopted. In the meantime use of weightingfactors inICRP 26 (ICRP 1977)maybe appropriate.

BEIR 1972:National Academy of Sciences, National

“The Effectson Populations ofExposure to Low Levels of Ionizing Radiation” (Na-tionalAcademyPress,Washington,DC, 1980).

41

BUHL 1984:T. E. Buhland W. R. Hansen, “Estimatingthe Risks of Cancer Mortality and Genetic DefectsResulting from Exposures to Low Levels of IonizingRadiation,” Los Alamos National Laboratory reportLA-9893-MS(May 1984).

ICRP 1977:“Recommendationsof the ICRP,” Intern-ationalCommissionon RadiologicalProtection Publica-tion 26 (1977).

STROM 1975:P. O. Strom and E. C. Watson, “Calcu-lated Dosesfrom Inhaled Transuranium Radionuclidesand Potential Risk Equivalenceto Whole Body Radia-tion,” International Atomic Energy Agency reportIAEA-sM-199/l14(1975).

UNSCEAR 1977:United Nations ScientificCommitteeon the Effectsof Atomic Radiation, “Sources and Ef-fects of Ionizing Radiation” (United Nations, NewYork, 1977).

UNSCEAR 1982:United Nations ScientificCommittee Ion the Effects of Atomic Radiation, “Ionizing IRadiation:Sourcesand BiologicalEffects”(United Na-tions,New York, 1982).

R

US Bureauof the Census 1982:“Preliminary Estimates --of the Population of the United Statesby Age,Sex,and .-Race: 1970to 1981,”US Bureauof the Census,Current ‘..Population Reports, SeriesP-25, No. 917 (US Govern- -ment PrintingOffke, Washington,DC, 1982).

US National Center for Health Statistics 1975:“UnitedStatesLifeTables: 1969-71,”US Department of Health,Education, and Welfare Publication (HRA) 75-1150(US Government Printing Office, Washington, DC,1975). .

---..

-...-

42

, I. I-.

Risk is the combination of probability of the occur-rence of a seriousevent and the possibleconsequencesof the event, either to persons,facilities,or the environ-ment. Various procedureshave been employed for thedetermination of estimated risks, including de-terministic, probabilistic,judgmental, and cost benefit(Rhyne 1983).

In deterministicassessment,specificreleasesare pos-tulated without an identified mechanism causing therelease and without any known probability of occur-rence. The primary exampleof this is the large“loss ofcoolant accident” assumed for LWRS. A calculatedconsequencesuch as a radiologicaldose is comparedwith some upper limi~ with the facilitysite consideredacceptable if no consequence exceeds this limit. Noinference of the safety of the reactor is made in thisassessment,only of the legalacceptabilityof the site.

Probabilisticcriteria provide an upper limit on theprobabilitythat a consequencewillexceeda givenvalueand thus require that all events which could contributesignificantlyto perfomlance criteria be considered.Al-though inherently more complex and requiring moredata, this method providesgreaterinsightinto potentialsystemfailures.

Judgmentalproceduresare usefulwhen acceptabilitycriteria are not explicitly defined and are left to thejudgmentof the analystor the regulator.This procedureis not considereda viable basis for many safety reviewcriteria.

The cost-benefitcriteria are attempts to compare theexpectedcosts with the expectedbenefits using a com-mon value system; however, there is a difficulty inplacing a common value on human life and injury.While used in some NRC regulatory processes, deci-sionsbased upon this method are usuallyof a relativelynarrowscoperather than the basisfor the entire process.

Current regulatoryprocessesconsistlargelyof overallsafety criteria developed over several decades. Theanalyst reviews the design to assure that the criteria

-- have been met. Current DOE and NRC criteria are#. largely deterministic and are based on scenarios and

valuesthat have been used in the past. For a number of.-reasons,probabilisticmethods are presentlyused more-.to supplement deterministic methods than to replacethem.

Of the three approaches considered acceptable foruse, the deterministic approach has the advantages ofgreater familiarity and wider acceptance by reviewersincludingthe lay public, less complexity,and generallytaking less time to perform the analysis. Its majordisadvantage is the lack of formal structure which al-lowsgreater variety of results, depending on the judg-ment of the analysts.NRC is encouraginggreateruse ofprobabilisticmethods(NRC 1983).Althoughtheyresultin a more comprehensive analysis, the probabilisticmethods require extensive training of the analyst andextensivesupportingdata,

In performing a deterministic risk analysis, theanalyst starts with a known or conservativelyassumedsequenceof events leading to a major release. Conse-quencesof the releaseare calculatedand then comparedwith a givenupper limit. If the postulatedconsequencesdo not exceedthisupper limit, the facilitysiteand majordesign features are acceptable under this event. Thecalculations are repeated for each event and can berepeated for the contributions between interrelated orpossible contributing events which might mitigate orenhancea givenaccident.

The analyst starts a probabilistic analysis with theconsiderationof all events which can contribute to theperformance(failure)of all of the given systems,assign-inga probabilityvalueto eachcomponentof the system.Usuallythis probabilityis the possibilityof a faiIureperunit of time, obtained from historical data and failurerecords and also best estimations. One then considersthe magnitudeof both the probabilityand the potentialconsequences to arrive at an assessment of the riskpresentedby the particular ftilure. An estimate is madeof the consequencespossiblethrough a stated chain ofevents. While more complex and requiring a greateramount of time and data, this method provides greaterinsightinto potentialsystemfailuresand potential inter-actions of the set of systems.For example, fkilureof asecondsystemmay produceamplificationof the conse-quencespostulatedfrom a prior failure.

One approach is to use a combination of the de-terministicand the probabilisticmethods, starting withthe deterministicmethod and followingwith probabilis-tic risk analysisas a confirmatory process.Usefid toolsfor these techniques are available, including fault treeanalysis (NRC 1981)and the Management OversightRisk Tree (MORT) diagrams (MORT 1975, Briscoe1979).These must be coupledwith adequate databases.The safetyanalystuses these to

. identi~ sources of energy within a system whichare largeenoughto causea major releaseof radioaetive or otherwisehazardousmaterial,

. identify the conditions under which this materialcould be releasedas wellas the factorswhichcouldampli~ or mitigatethe release,and

● estimate the probabilityand magnitudeof the pos-tulated release and estimate the consequencesofthe postulatedrelease.

This same procedure is followedfor each step from theinitiating event, through the entire system, to the esti-mation of the consequencesto the facility, the peopleexposed, and the environment. For whatever methodand toolsare usedby the safetyanalystin the assessmentof risks, it is extremely important that the methods,estimations, and calculationsbe adequately supportedby documentation and examples, to allow reviewerstoapproximate the steps in the analysisand validate theconclusionsof the analyst.

Q

A qualitative or relative risk evaluation approachcurrently in use (Lucas 1981,ANSI 1976,Brynda 1981,DOD 1977,ORO 1984)requiressubjectivejudgment inassigningan accidentto a probabilityclass,forexample,anticipated, unlikely,extremely unlikely,or incredible.These classes may be assigned ranges of numericalprobabilities.A dose guidelinemaybe selecteddepend-ing on the analyst’sdetermination of probabilityclass;the more probable the accident, the lower the doseguideline selected. This method leaves more to theanalyst’sjudgment but provides a systematicapproachto risk assessment.

Q

Quantitative methods are currently in use at DOEnuclearfacilities(Durant 1980,1981;Lucas 1981).Acci-dent probabilities are based on incident frequenciesrecorded in extensive data bases. Because the rela-tionship between frequency of incidents (componentfailures,operatorerrors, etc.)and accidentprobabilityisnot well known, subjectivejudgment cannot be com-pletely eliminated from the process. However, for-malizedriskassessmentmethodssuchas PRA (probabi-listicrisk assessment)(NRC 1983)and fault tree analy-sis (NRC 1981),coupledwith an adequate incidentdatabase, can be valuable tools in evaluating the risks as-sociated with a facility.A conservative risk limit mayalso be applied in conjunction with this risk evaluationmethod.

For many DOE facilities,an adequate incident database may not exist, preventing an estimate of theprobability of occurrence of a certain event. Becausebroadly accepted risk guidelinesdo not presentlyexistand presentexperienceis not adequate to allowcalcula-tion of the probabilitiesof most accidents,adoption byDOE of a risk assessmentmethod is not recommended.The probabilistic methods certainly provide a usefulsupplementto the deterministicmethod currentlyin useand shouldbe continued by localoption.

ANSI 1976:“Criteria for the Design of Plants for theManufacture of Mixed Oxide (U-PU) Fuels,” ANSIN287-1976 (American National Standards Institute,Inc., New York, December30, 1976).

Briscoe 1979:G. J. Briscoe,“Application of Manage-ment OversightRisk Tree (MORT) to Reviewof SafetyAnalysis,”EG&G Idaho reportSSDC-17(1979).

Brynda 1981:W.J. Brynda,L.Junker, R. C. Karol, P. R.Lobner, and L. A. Goldman, “Nonreactor Nuclear Fa-cilities:Standardsand CriteriaGuide,” US Departmentof Energy report DOE/TIC-l 1603 (BNL 51444) (Sep-tember 1981).

DOD 1977:“Military Standard SystemSafetyProgramRequirements,” US Department of DefensedocumentMIL-STD-882A(June 1977).

Durant 1980:W. S. Durant, “The Applicationof Proba-bilisticRisk Assessmentto Nuclear Fuel Reprocessingat the Savannah River Plant,” E. I. du Pent de Nemoursand Co., Savannah River Laborato~, US Departmentof Energyreport DP-MS-80-59(November 1980).

Durant 1981:W. S. Durant, “Savannah River Labora-tory Data Banksfor RiskAssessmentof FuelReprocess-ing Plants,” US Department of Energy report DP-MS-81-90(E. I. du Pent de Nemours and Co., Aiken,South Carolina,October 1981).

Lucas 1981:D. E. Lucas, “Safety Analysis Guide forNonreactor Nuclear Facilities,” Hanford EngineeringDevelopment Laboratory report HEDL-MG-153(1981).

MORT 1975:“Management Oversightand Risk Tree”(SystemSafety Development Center, Aerojet General,Idaho Falls,Idaho, June 16,1975).

--

-.4-

-..-

44

NRC 1981:“Fault Tree Handbook,” US Nuclear Regu- ORO 1984:“SafetyAnalysisand ReviewSystem,”DOElatory Commission report NUREG-0492 (January Oak Ridge Operations Oflice Order OR-5481.lB (May1981). 23, 1984).

0 NRC 1983:“PRA Procedures Guide. A Guide to the Rhjne 1983:W. R. Rhyne, “A Proposed Approach for--Perfiormanceof ProbabilisticRisk Assessmentsfor Nu- the SafetyReviewof the New Production Reactor Pro-

. clearPowerPlants,” Vols.I and II, US Nuclear Regula- ject” (H & R Technical Associates, Inc., Oak Ridge,tory Commission report NUREG/CR-2300 (January Tennessee,January 1983).~-1983).

---

. .

-

45

APPENDIX D

DISPERSION CALCULATION METHODS

I. BASICDISPERSION EQUATION

To calculate downwind doses from an accidentalreleaseof radioactivematerial, air concentrationsof thegasor airborneparticulatematerialmust be determined.The straight-lineGaussian dispersion equation is typi-cally used.The Gaussian formula for the air concentra-tion, ~, at a downwind location (x,Y,z)may be found inSlade(1968).

Air concentrations should be calculated for 16 com-pass directions (22.5”sectors centered on north-northnortheast etc.).

The Gaussian dispersion model is estimated to beaccurate to within a factor of 2 for distances of 0.1 to10-20km when onsite meteorologicaltower data areavailable and conditions are reasonably steady andhorizontallyhomogeneous(AMS 1978).Beyond20 km,Gaussiandispxsion calculationscan onlybe consideredto be order-of-magnitudeestimates. Conditions whichwill reduce the accuracy of the Gaussian dispersioncalculationsincludeaerodynamicwake flows,rough orurban terrain, very buoyant or dense gases, or dis-persionunder very stableor unstableconditions.

The use of the Gaussian model may not be ap-propriate for reactive gases and particulate matter.Alternative methods accounting for possible trans-formations should be investigated. The dispersion ofdensegasesis discussedby Britter(1979, 1980,1983).

II. DISPERSION PARAMETERS/STABILITYCLASSIFICATIONS

The horizontaland verticaldispersioncoefficients,CYand c- required for the Gaussian dispersion equationcan be estimated from measurements of the standarddeviation of the wind angles or by estimating the at-mosphericstabilityand usinga setof curveswhichareafunction of downwind distance and stability. Ifmeasurements of co, the standard deviation of thehorizontalcrosswindcomponent of the wind, are avail-able, creis estimated from the formula Cy= c%xf7x),where x is the downwind distance from the source.Several sets of fhnctions have been presented for fix)(Cramer 1976,Draxler 1976,Horst 1979,Irwin 1979A,Pasquill 1976);some of these functions have been re-viewedby Irwin (1983).

If measurementsof C9are not available, the stabilityclassmust be determined. Severalmethods for estimat-ing stability class are currently in use (Turner 1970,

●

✍✎

NRC 1974, NRC 1983). The NRC has also issued .-regulatoryguideswhich apply to specificfacilities;Reg- .-rulatory Guides 1.4(NRC 1974A)and 3.33 (1977A)aretypical. One of the more widely used methods, the -vertical temperature gradient (AT) method (Hanna1977,NRC 1983),has the disadvan~ge of not assessing

the mechanicalcomponent of turbulence, that is, windshear and surface roughness. Thus, one of the othermethods, including both the effects of buoyancy andmechanicalmixing,may be a better selection.The Rich-ardson number, bulk Richardson number, and Monin-Obukhov length (Golder 1972) can also be used todeterminethe stabilityclass.

Functions relating measurementsof a,, the standarddeviation of the vertical wind angle, to q have beenpresented (Draxler 1976,Cramer 1976,Pasquill 1976,Irwin 1979A).However,determiningaZdirectlyfrom a.has not been recommended by the American Meteoro-logicalSociety(Hanna 1977)as the standard method fordetermining a, because of the difficultyin making ac-curate measurementsof q. Thus the atmospheric stab-ilityclasscan be determined in a manner similar to thatdescribedfor CZ,and o. is then determined from a set ofcurves.Many sets of curves are availablefor determin-ingayand a,, dependingon the sourceheigh}averagingtime, etc. For OY,the AMS (Hanna 1977)has suggestedthat Pasquill-Gifford (Gifford 1961) curves and theMcElroy-Pooler(1968)urban curves are acceptable ifadjustments are made for sampling duration and sur-faceroughness.Curvespresentedby Briggs(1973)for Cyand a, have combined data from the Pasquill-Gifford,Brookhaven, and TVA curves for dispersion fromelevatedsources.Lamb (1979)has presented equationsfor o, (and the effectivereleaseheighth) for nonbuoyantreleasesinto an unstableatmosphere.Somesite-specificdispersion curves have also been developed (Yanskey1966, Fuquay 1964)which are the most appropriatechoicesfor thosesites.

Curvesdescribingplume dispersionare often used todescribethe dispersionof a puff release.However, puff ~releases initially have a faster rate of growth than -plumes (Hanna 1982).Gifford (1977)has summarized --22 experiments of relative (puff) diffusion showing +cloud width proportional to (time)312for travel times -between 1000and 3000s. Beyond 10000s (2.78h), the ‘:puff growth rate slows to approximately a linear de- -pendence on time. A method for calculatingpuff dis-persion coefficientsis presented by Hanna (1982).Toaccount for the size of the puff at the releasepoint, thedispersion coefficientsshould be initially set equal to(puffradius)/2.15(Turner 1970).

46

m.

--

.-.

.+

--l

.

--

111.RELEASE EFFECTS

The dispersionequations may require modificationsdue to plume rise or buildingwakeeffects.The specificmodificationsare presentedin the sectionsdealingwiththosetopics.

Equationsappropriate for the descriptionof buoyantand momentum plume rise have been presented byBriggs (1969, 1975). The reader is referred to thesesourcesfor a completepresentation of these plume riseequations.An expressionfor the cloud rise from a high-explosive detonation has been presented by Church(1969).The verticaldispersioncoefficient,a., may alsorequire modificationdue to plume rise. Pasquill (1976)has suggestedthat a, be enhanced for a buoyant plume.

If the exit velocityof a stack release is less than 1.5times the wind speed at stack height, stack tip down-washis considered.Briggs(1973)presentsa method forcalculation.

If the release height is less than two and one-halftimes the heightof the buildingor adjacent solid struc-ture, building wake effects are considered. Hosker(1981)and NRC (1982)present guidance for the selec-tion of appropriate equations to assess building wakeeffects.

As the cloudof releasedmaterial movesdownwind,anumber of processesmay affect the air concentrations.These include radioactivedecay, plume trapping by aninversion,dry deposition, etc. The necessarymodifica-tions to the dispersion equation are presented in thefollowingsections.

R

The amount of radioactive material will be affectedby radioactive decay and daughter product buildup asthe release travels downwind. These effects are ac-counted for in the decay and buildup models cited inSectionV.A., SourceTerms.

As the release travels downwind, its vertical spreadcan be limited by the presenceof an inversion or by themixing height. The Gaussian dispersion equation ismodifiedto considerreflectionfrom this elevatedstablelayer. The Gaussian equation is adjusted to includeretlcctionfrom a stablelayerby adding terms accordingto Turner (1970). Climatologicalestimates of mixirigheights can be obtained from Holzworth (Holzworth1972).Hourly estimatesof mixingheightscan be madeusingradiosondedata, acousticsoundermeasurements,or a parametric relationship which specifies mixingheightas a fi.mctionof other boundary-layerparameters(Arya 1981,Venkatram 1980,Benkley1979).The valueof the vertical dispersioncoefficient,o,, should be lim-ited to the depth of the mixing layer or the inversionheight.Alternatively,the Gaussian dispersionequationcan be usedwithout modificationif aZis limited to 0.8P,where J?is the depth of the mixing layer. Beyond thedistance where CZ=f!,the material is spread uniformlybetween the ground and the lid. The air concentrationcan then be expressedin a simpler form of the Gaussianequation (Turner 1970).

High ground-levelconcentrations can be producedduringfumigationconditions(Hanna 1982).These con-ditions can occur in the proximity of large bodies ofwater(NRC 1982)or fora shortperiod followingsunrisewhen a surface-based inversion is present. Modifica-tions to the Gaussian equation to account for fumiga-tion are presentedby Turner (1970)and Hanna (1982).Guidance on the length of time fumigation conditionsmay be observed is provided by NRC (1982). IdahoNational EngineeringLaboratory uses l-h duration ifrelease height exceeds 75 m and 15-30min for lowerreleaseheights.

Under unstableor neutral atmosphericconditions,anairborne releasewill tend to rise over downwind terrainobstacles.However, the oliginal effectivereleaseheightabove the terrain will not be maintained. The effectivereleaseheightshouldbe reducedby the terrain heightorthe release height divided by 2, whichever is smaller(Briggs1973).Under stableatmosphericconditions, thereleasewillnot risewith the terrain and so may impingeon the ground if the obstacleis sufficientlytall.

If a releaseimpactselevatedterrain, an artificialjumpin the concentrationat the terrain featurecan occurdueto the reflection term in the Gaussian equation (Egan

47

1979).This effectcan be corrected by the requirementthat along the axis of maximum concentration, theconcentrationcannot increasewith distance.

Under stableconditions,diffusion in a valley is lim-ited by the valleywalls.When the width of the valley,w,is equal to 2CY,the highestconcentrations occur alongthe valley wall. These may be calculated as shown byHanna (1982).

Another approach to estimating concentrations incomplexterrain under neutral and stable conditions isto define an amplification factor, the ratio of the max-imum concentrationoccurringin the presenceof terrainto the maximum concentration from the same sourcelocated in level terrain. Snyder (1983)has summarizedseveral wind-tunnel and towing-tank studies whichhave been performed to defineamplificationfactorsforsources located in various positions with respect toterrain features.

Complexterrain can also increasethe horizontal andvertical dispersion of a plume or puff as it travelsdownwind. If the Pasquill-Gifford stability typingschemeis used for a location in complex terrain, it hasbeen suggested(Stnmaitis 1981)that during nighttimestable conditions, the Pasquill-Giffordstability shouldbe changed by one class toward unstable to reflect theincreased dispersion. Where possible, onsite &tashouldbe evaluated to determineappropriate modifica-tions of dispersionparameters in complexterrain.

Ground deposition can be estimated by multiplyingthe radionuclideconcentration in air at ground levelbya deposition velocity representative of the particles inthe cloud. Deposition velocities for unitdensitysphericalparticles,0.1- to 100-ymD.., range from 10–4to 25 cm/s. A common assumption is that depositionvelocityvaries from 0.1 to 10cm/s with an averageof 1cm/s (Sehmel 1980).

Dry deposition for gasesis treated in the same man-ner as particulate material. Deposition velocities forgases range from 10-4 to 10 cm/s; 1 cm/s is oftenassumed for deposition calculations (Sehmel 1980).Noble gases should be treated as having a zero depo-sitionvelocity.

If ground deposition is calculated, downwind airconcentrationsare modified to reflect the effectivede-creasein the source term. This is done by replacingthetotal emission,Q, in the dispersionequation with Q(x),the source remaining at a downwind distance, x (NRC1983) OverCamp (1976) has presented a modifiedGaussianplume model for calculatingdry depositionoffineand heavy particlesand gases.It combinesa down-ward slopingplume to account for settling and a con-stant depositionvelocity.

Plume depletion from wet deposition may be con-sidered for determining the total amount of materialdeposited on the ground followingthe release. In gen- .eral, wet deposition should not be included in the u-calculationof air concentrationsunless it can be shown .that the release has a strong probability of occurrence “.duringa rainstorm or snowstorm.

--

The amount ofmatenal depositedon the ground can -be calculated as a tlmction of the air concentration(NRC 1983).The air concentration may be reduced bywashout as it travels downwind (Hanna 1982). Ascavengingcoefficientcan be determined as a functionof rainfall rate and a stabilitydependent coefilcientaccordingto Ritchie(1978).

The meteorologicalvariablesrequired for evaluatingdownwind air concentrations are typically developedfrom 2-3yr ofdata collectedat a locationrepresentativeof the site of the release (Strimaitis 1981)or are con-servativelyspecifiedas PasquillType F and wind speedof 1 m/s. Guidance for onsite meteorologicalmeasurement programs, including instrument location andmeasurement techniques, is presented by Strimaitis(1981).The estimation ofwind speedat releaseheightisdiscussed by Irwin (1979B) and Hanna (1982). Thetreatment of calm winds is discussed in NRC (1977B)and Hanna (1982).Two meteorologicalcategories,me-dian and unfavorable,are suggestedfor accident-releasedispersioncalculations.Becauseaccidentreleasesare ofshort duration, the median and unfavorabledispersionfactorsare assumed to be constant during the durationof the release.Descriptionsof unfavorableand mediancategoriesare providedbelow.

To calculatethe concentration to which the exposedperson is exposedunder unfavorableconditions,hourlydispersion factors (~/Q) should be calculated at thedistanm of the offsiteperson. For each of 16 sectors, a _cumulative probability distribution of X/Qs should beconstructed.The z/Q value that is exceededby 0.5%of ~-the total number of hourly ~/Qs in the data set is the ~-“unfavorable” dispersion factor for that sector. For -example, if the data set comprises 8760 observations, ‘-the 0.5%X/Qforthe sectoris the~/Qexceededby O.005 “-x 8760or 44 observations.The sectorhavingthe highestunfavorabledispersionfactordefinesthe wind directionassumed to occur during the release, and the un-

48

favorable dispersion factor in this sector is used tocalculatethe desireddose.

The 0.5%~/Q wasselectedfor consistencywith NRCRegulatoryGuide 1.145(NRC 1982).The 0.5%sector.

.-, x/Q waschosenbythe NRC as beingconsistentwith the5%direction-independent~/Q (NRC 1981),while al-.

> lowingthe considerationof the directionaldependence-. ofatmosphericdiffusionconditions.The NRC retained

a requirementin RegulatoryGuide 1.145for comparingthe highest 0.5% sector x/Q with the 5% direction-independent x/Q and selecting the highest for dosecalculations.However, from a parametric study (NRC1981),it wasjudged that for most sitesthe 0.5%z/Q willbe the most conservativez/Q and the comparison withthe 5%directionindependent~/Q need not be included.

To calculate the concentration to which the max-imallyexposedperson is exposedunder median ccmdi-tions,a similarprocedureis followed.For each sector inwhich the person is located, hourly dispersion factorsare calculated for the distance to the person from thepoint of release.A cumulative probability distributionof ~Qs is developed,and the ~Q which is exceededby50%of the hourlydispersionfactorsin that one sectorisdefined as the “median” dispersion factor. The sectorhaving the highestmedian dispersion factor definesthewind directionassumed to occurduring the release,andthe dispersion factor in this sector is used to calculatethe desireddose.

To calculate the population dose, the release is as-sumed to travel in the direction having the largestnearby population. The meteorological conditionsproducing the unfavorable dispersion factor for thissectorare used to calculatethe dose to the peoplein thatsector. If there are population centers in several direc-tions from the release, several sectors may requireevaluationto determine the largestpopulation dose.

It is not considered necessary to expose the entirepopulationto the concentration at the cloud centerline.Integratedactivityconcentrationsover the area contain---ing people and the actual population density in those.-areas wouldbe used to calculatepopulationdose.

--

AMS 1978: American Meteorological Society, “Ac-curacy of Dispersion Models—APosition Paper of theAMS 1977Committeeon AtmosphericTurbulenceandDiffusion,” Bulletin of the American MeteorologicalSociety9 (8) (August1978).

Arya 1981:S.P.S. Arya, “Parameterizing the Height ofthe Stable Atmospheric Boundary Layer,” JournalofApp/iedMeteoro/ogy20, 1192-1202(1981).

Benkley1979:C. W. Benldeyand L. L. Shulman, “Esti-matingHourlyMixingDepths from HistoricalMeteoro-logical Data,” Journal of Applied Meteorology

Briggs 1969:G. A. Briggs,Plume Rise, AEC CriticalReview SeriesTID-25075 (ClearingHouse for FederalScientificand Technical Information, Springfield,Vir-ginia, 1969).

Briggs 1973:G. A. Briggs,“Diffusion Estimation forSmall Emissions,” ATDL Contribution File No. 79(Atmospheric Turbulence and Diffusion Laboratory,Oak Ridge,Tennessee, 1973).

Briggs1975:G. A. Briggs,“Plume Rise Predictions,” inLectureson Air Pollutionand EnvironmentalImpactAnalyses, WorkshopProceedings,Boston, Massachu-setts, September 2—October 3, 1975(American Mete-orologicalSociety,Boston,Massachusetts,1975).

Brittw 1979:R. E. Britter, “The Spread of a NegativelyBuoyantPlume in a Calm Environment,” AtmosphericEnvironment

Brittcr 1980:R. E. Bntter, “The Ground Extent of aNegatively Buoyant Plume in a Turbulent BoundaryLayer,”AtmosphericEnvironment

Britter 1983: R. E. Britter and R. F. Griffiths, Eds.,Dense Gas Dispersion (Elsevier Scientific, 152 Van-derbiltAve.,New York, 1983).

Church 1969:H. W. Church, “Cloud Rise from HighExplosiveDetonations,” Sandia National Laboratoriesreport SC-RR-68-903(May 1969).

Cramer 1976:H. E. Cramer, “Improved TechniquesforModelling the Dispersion of Tall Stack Plumes,” in“Proceedingsof the 7th Int. Technical Meeting on AirPollution Modelling and its Application,” NationalTechnical Information Service report PB 270 799(1976),pp. 731-780.

49

Draxler 1976: R. R. Draxler, “Determination of At-mospheric Difision Parameters,” Atmospheric En-vironment 99-105(1976).

Egan 1979: B. A. Egan, R. D’Errico, and C. Vaudo,“Estimating Air Quality Levels in Regions of HighTerrain Under Stable Atmospheric Conditions,” inFourthSymposium on Turbulence,Dijkion, and AirPollution (American MeteorologicalSociety, Boston,Massachusetts,January 1979).

Fuquay 1964:J. J. Fuquay, C. L. Simpson, and W. T.Hinds, “Prediction of Environmental Exposures fromSources Near the Ground Based on Hanford Ex-perimental Data,” Journalof Applied Meteorology3,761-770(1964).

Gifford 1961:F. A. Gifford,“Uses of Routine Meteoro-logicalObservations for Estimating Atmospheric Dis-persion,”NuclearSafety2 (4),47-51(1961).

Gifford 1977:F. A. Gifford, “Tropospheric RelativeDifision Observations,”JournalofAppliedMeteorol-ogy

D. F. Golder, “RelationsAmongStabilityParameters in the SurfaceLayer,” Bounda~ LayerMe-teorology3,47-58 (1972).

Hanna 1977:S. R. Hanna, G. A. Briggs,J. DeardorfT,B.A. Egan, F. A. Gifford, and F. Pasquill, “AMS Work-shop on Stability Classification Schemes and SigmaCurves-Summary of Recommendations,” Bulletinofthe American MeteorologicalSociety 58, 1305-1309(1977).

Hanna 1982: S. R. Hanna, G. A. Briggs,and R. P.Hosker, “Handbook on Atmospheric Diffusion,” USDepartment of Energyreport DOE/1lC-1 1223(1982).

Holzworth 1972:G. C. Holzworth, “Mixing Heights,Wind Speeds, and Potential for Urban Air PollutionThroughout the ContiguousUnited States,” US Envi-ronmental ProtectionAgencyreport AP-1OI(1972).

Horst 1979: T. W. Horst, J. C. Doran, and P. W.Nickola, “Evaluation of Empirical Atmospheric Dif-fusion Data,” US Nuclear RegulatoryCommission re-pOltNUREG/CR-0798,PNL2599 (1979).

Hosker 1981:R. P. Hosker, “Methods for EstimatingWakeFlowand EtlluentDispersionNear SimpleBlock-Like Buildings,” NDAA Tech Memorandum ERLARL108 (AtmosphericTurbulenceand DiffusionLab-oratory,Oak Ridge,Tennessee, 1981).

Irwin 1979A: J. S. Irwin, “Estimating plume Di.+persion—A Recommended Generalized Scheme,”Proceedingsof 4th Symposium on Turbulence,Dif&ion, and Air Pollution (American MeteorologicalSociety,Boston,Massachusetts,1979),pp. 62-69.

Irwin 1979B:J. S. Irwin,“A TheoreticalVariationof theWind Profile Power Law Exponent as a Function ofSurface Roughness and Stability,” Atmospheric En-vironment13, 191-194(1979).

Irwin 1983: J. S. Irwin, “Estimating plume Dis-persion-A Comparison of Several Sigma Schemes,”Journalof ClimateandAppliedMeteorology22,92-114(1983).

Lamb 1979:R. G. Lamb,“The Effectsof ReleaseHeighton Material Dispersion in the Convective PlanetaryBounda~ Layer,” FourthSymposium on Turbulence,Di&ion, and Air Pollution(American MeteorologicalSociety,Boston,Massachusetts,Janua~ 1979).

McElroy 1968:J. L. McElroyand F. Pooler, “St. LouisDispersion Study,” US Department of Health, Educa-tion, and Welfkrereport AP-53(1968),Vol. 2.

NRC 1974:“Onsite MeteorologicalPrograms,”US Nu-clear Regulatory Commission Regulatory Guide 1.23(1974).

NRC 1974A:“Assumptions Used for Evaluating thePotential Radiological Consequences of a Loss ofCoolantAccidentfor PressurizedWater Reactors,” USNuclear RegulatoryCommission RegulatoryGuide 1.4(June 1974).

NRC 1977A:“Assumptions Used for Evaluating thePotentialRadiologicalConsequencesof AccidentalNu-clear Criticality in a Fuel Reprocessing Plant,” USNuclear Regulatory Commission Regulatory Guide3.33(April 1977).

NRC 1977B:“Methods for Estimating AtmosphericTransport and Dispersionof GaseousEflluentsin Rou-tine Releasesfrom Light-Water-CooledReactors,” Nu-clear RegulatoryCommission RegulatoryGuide 1.111(July 1977).

NRC 1981: “Technical Basis for Regulatory Guide1.145:Atmospheric Dispersion Models for PotentialAccident ConsequenceAssessmentsat Nuclear PowerPlants,” US Nuclear Regulatory Commission reportNUREG/CR-2260(1981).

I

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.

50

NRC 1982:“Atmospheric Dispersion Models for Po-tential Accident ConsequenceAssessmentsat NuclearPower Plants,” US Nuclear Regulatory CommissionRegulatoryGuide 1.145(1982),p. 14.

●--. NRC 1983:“A Guide to the Performanceof Probabilis-

. tic Risk Assessmentsfor Nuclear Power Plants,” US..3 Nuclear Regulatory Commission report.

NUREG/CR-2300(1983).

Overcamp 1976:T. J. OverCamp,“A General GaussianDiffusion-Deposition Model for Elevated PointSources,”JournalofApp/iedMeteoro/ofl

Pasquill 1976: F. Pasquill, “Atmospheric DispersionParameters in Gaussian Plume Modeling: Part 2,PossibleRequirementsfor Changein the Turner Work-book Values,” US Environmental Protection Agencyreport EPA-600/4-76-0306(1976).

Ritchie 1978:L. T. Ritchie, W. D. Brown, and J. R.Wayland,“Effectsof Rainstorms and Runoffon Conse-quencesofAtmosphericReleasesfrom Nuclear ReactorAccidents,”NuclearSafety19,220-238(1978).

Sehmel 1980: G. A. Sehmel, “Particle and Dry GasDeposition:A Review,” AtmosphericEnvironment14,983-1011(1980).

Slad.e1968:D. H. Slade,Ed., “Meteorologyand AtomicEnergy,” US Atomic Energy Commission reportTID-24190(Jdy 1968).

Snyder1983:W. H. Snyder,“Fluid Modelingof TerrainAerodynamicsand Plume Dispersion-A PerspectiveView,” Sixth Symposiumon Turbulenceand Dijiuion(American MeteorologicalSociety,Boston, Massachu-setts,March 1983).

Strimaitis 1981: D. Strimaitis, G. Hoffnagle, and A.Bass,“Onsite MeteorologicalInstrumentation Require-ments to Characterize Diffusion horn Point Sources,Workshop Report,” US Environmental ProtectionAgencyreport EPA-600/9-81-020(April 1981).

Turner 1970:D. B.Turner, “Workbookof AtmosphericDispersion Estimates,” US Department of Health,Education,and Welfmereport PB 191482(1970).

Venkatram 1980: A. Venkatram, “Estimating theMonin-ObukhovLength in the Stable Boundary Layerfor Dispersion Calculations,”BoundaryLayer Meteor-ology19,481-485(1980).

Yanskey 1966:G. R. Yanskey,E. H. Markee, and A. P.Richter, “Climatographyof the National Reactor Test-ing Station,” US Environmental Science Service Ad-ministration report IDO-12048(1966).

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L I

Doses to whole body and to specificorgans from apostulatedaccident are calculated for comparison withthe site criteria doses proposed for Chapter I of DOEOrder 6430.1.Contributions to dose from the accidentcase come primarily from radioactive material inhaledduring cloud passageand direct radiation to the wholebody received while immersed in the cloud. Otherpossible sources of dose (ingestion by the cow-milk-human or crop-food-humanpathways,direct radiationfrom the facility, or delayed exposures from con-taminated water or land) are secondary contributorswhichmay require evaluation under specialconditions.

Inhalation dose depends on the time-integratedradionuclide activity concentration (Ci ”s/m3 orBqs s/m3), the breathing rate of the subject (m3/s),thefraction of the inhaled radionuclide reaching the organof interest, and the organ dose conversion factor(rem/Ci or Sv/Bq). Recommendationsare made in thechoice of parameters to calculate these major dose-related factors.

Dose to an organ depends on how soon the radio-nuclideclearsthe lungs,how much of it is transferred tothe organ through the bloodstream,and how much andhow long it is retained by the organ tissue. Severalmodels are available which also account for dose re-ceived by an organ from nearby organs containingradioactivematerial.

The primary methods used to calculatedose or doseconversion factorshave been publishedby the Intern-ationalCommissionon RadiologicalProtection (ICRP),an authoritative radiation protection organization.Thevarious contractorsand offkes in the DOE are in vary-ing stages of transition from the ICRP Publication 2modelsfor lung bone, and other organs(ICRP 1959)tothe recent models in ICRP Publication 30 (ICRP 1979,1980).Morecomplexthan ICRP 2, the ICRP 30modelsaccountfordoseto (target)organsfrom beta-or gamma-emitting nuclidesdeposited in neighboring(source)or-gans.This added complexityaccounts for no changein

. .-s

dose if the nuclideis an alpha emitter but maybe quite .-Iargefor some organsif the nuclideis a gamma emitter. .~..For example, lung dose directly from depositedlJTCs-lJTBa(ClassD) only0.0025rem/Ci (6.8x. -10-10Sv/Bq),whiledosefrom neighboringorganswouldcontribute an additional 0.014 rem/Ci (3.8 x 10-9Sv/Bq). Dose calculations for mixed fission productsgenerallyyield higher results with the ICRP 30 modelsthan with earlier models. The major differencesin theICRP 2 and ICRP 30 models have been discussed inseveralpublications(Bair 1979,Runkle 1981).

The apparent movement towardadoption of ICRP 30models is considered favorablebased on the followingobservations:

The methods are recommended by a recognizedinternational commission whose earlier recom-mendationshave been broadly accepted.The methods are believed to be made more ac-curate by the source-target dose refinement de-scribed above and by updated transfer fractionsincorporatedin it.Progressis being made toward adoption of ICRP30modelsor variationsof them by NRC, EPA,andother US agencies,such as the current revision of10CFR 20 occupationaldose standardsby NRC.DOE is currently revising ChatXer XI of DOEOrder 5480.1Ato-provide~CRP30-basedexposurelimits.

Computercodesbasedon modelsother than ICRP 30generallyyield comparable results for alpha emitters ifcommon input parameters are used. The major sourcesof variation come from choice of quality factor, organmass, transfer fraction, compound volubility(clearancerate), and dose model for some organs.The organ dosemodels and major parameters used in the major dosecodesare describedin later sections.

Since ICRP has converted to the International Sys- -tern of Units (S1),usingsievert (Sv) for dose equivalent -and bccquerel(Bq)for activity,similar usagewithin the --DOEadds conveniencefor those ofilcesusingthe ICRP -30 models.Accordingto DOE Order 6430.1,increased -useofSIunits is encouragedon a voluntarybasis.In this -;transition period, it is suggestedthat quantities in other ‘units be accompanied by converted quantities in S1units.

The ICRP Task Group on Lung Dynamics (TGLD)model (ICRP 1966)has been incorporated into most of

-- the computer codes presently in use within DOE. The-- 1966model has been modified slightlyin ICRP 19and

r. ICRP 30 (ICRP 1979).The form recommended for useris found in ICRP 30. Regionaldeposition fractionsare<provided in ICRP 30 and its supplements for only one.particle size (1.O-~mactivity median aerodynamic di-ameter), although a formula for calculatingdepositionfractionsof particlesof other sizesis provided.

The ICRP lung model is based on these assumptionsand considerations(Bair 1979):

Irradiation of the lung is more limiting than ir-radiation of lymphatic tissues, although for someradionuclide compounds, the average doses tolymph nodes may be many times greater than theaveragelungdose.The risk of radionuclidesas particlesin the lungsislikelyto be lessthan if the sameamount of materialis distributed more uniformly.Dose equivalents to specific regions of therespiratory tract can be estimated; however, be-cause of the many uncertainties regardingcells atrisk, loerdizationof depositedradionuclides,etc., itwas concluded that such estimates are unwar-ranted.In adults, the tracheobronchialregion,pulmonaryregion, and the pulmonary lymph nodes are con-sideredas one organ of 1000g.The dose to the nasopharyngeal region wasneglectedsince it is usually small compared withother regions.Radioactivedaughtersremain withand behavelikethe parent radionuclide.

The gastrointestinal tract has been partitioned intofour sections (stomach, small intestine, upper large in-testine, and lower large intestine) accordingto the bio-logicalmodel developed by Eve (Eve 1966).The Eve

4 model has been at least temporarily adopted by ICRP- (ICRP 1977,1979).The Eve model calculatesthe dose

-- at the surface of the bolus (contents), not through theA mucous layer or at the site of the most sensitive cells;

however, application of a 1/100 factor corrects this.- defectfor alpha emitters (Bair 1979).--The transferof radioactivematerials to body fluidsis

estimatedfrom the fractionofa stableelementabsorbedinto the blood followingingestion(fl values).Valuesoff, for a number ofclassesof compoundsare includedforeach element (ICRP 1979). For radioactive decay

daughtersformed in the GI tract, the value of f, used isthat appropriate for the parent nuclide. Also, themetabolicbeha~iorof the daughter is assumedto be thesameas that of the parent.

ICRP 30departs from the ICRP 2 practice,whichhasbeen to adjust volume bone dose for distribution ofsurfaceseekers(Pu and Sr) relative to radium by apply-ing a distribution factor n of 5. Doses are estimated bythe ICRP 30 model for two regions of bone, both ofwhich are considered to be at risk for cancer inductionby radiation (Bair 1979):

Marrow-Since hematopoietic stem cells are as-sumed to be randomly distributed in the marrowwithin trabecular bone of adults, the dose equiva-lent to the hematopoietic cells is calculated as theaverage over the marrow filling the cavities. Themass ofactive red marrow in the trabecularbone istaken to be 1500g.Bone Surfaces-For the osteogenic cells on en-dosteal surfacesand epitheliumson bone sufiaces,the dose equivalent is calculated as the averageover tissue up to a distance of 10 pm from therelevant bone surfaces.The total endosteal area istaken to be 12 m2. The mass of the ltl+m-thicklayeris taken to be 120g.

Severalassumptionsare made in the absenceof dataon the distribution of radionuclidesin bone:

●

●

Radionuclides of the alkaline earths (Mg, Ca, Sr,Ba, Ra) with radioactive half-livesof greater than15days are considered to be uniformly distributedthroughout the bone volume.Radionuclideswith half-livesof less than 15 davsare consideredto be distributed on bone surfaces.

Bonedose from actinidescan be affectedby burial ofdeposited material by new bone mineral, which isvariable with age of the person and rate of exposure;however, this effect was considered too complex to beaccountedfor in the ICRP 30 model.

As noted earlier, the particle size distribution in anewlyformed cloudchangeswith time as smallparticlesagglomerateinto largerpa]liclesand the largerparticlessettle out of the cloud. An estimate should be made ofthe particlesizecharacteristicdescribedby activity me-dian aerodynamic diameter (AMAD)that exists where

53

the exposedperson receivesthe exposure.This AMADbecomesinput to the TGLD model, from which comesregional deposition in the respiratory system. TheAMAD of the respirablefraction [massassociatedwithparticleswhoseaerodynamicequivalent diameter (D=)is less than 10 ~m] is the desired input to the TGLDmodel, rather than AMAD of all particlesremaining inthe cloud.One-micronAMAD has been used widelytoprovide a conservative estimate of inhalation dose forplutonium and umnium aerosols;however, the actualAMAD may lead to largerdoses from accidentsinvolv-ingparticlesizessmallerthan 1~m.

Particle size information for plutonium undergoingsimulated accidentconditions is extensive,particularlythe plutonium release studies performed by Mishimaand his colleagues(1965, 1966,1968,1973,and others).Selby(1975)has summarized particle size and concen-tration data for accident cases involving plutonium.Kirchner (1966)and Elder (1974)characterized pluto-nium aerosolsin gloveboxesand ducts under nonacci-dent conditions. More recent plutonium aerosolcharacterizationshave been reported by Raabe (1978).Information on size and amounts of particles poten-tially released by accidents in fiel reprocessing orprocessesinvolvingmaterialsother than plutonium hasnot been locatedduring preparation of the Guide.

Selection of compound volubilitycan cause broadvariability in results of dose calculations, as can beobserved in the ICRP lung model (ICRP 1966).Clearanceof inhaled materials in the ICRP lung modelis describedby D, W, or Y classification:D for thosewith a pulmonary clearance half-time of less than 10&ys, W for 10to 100days, and Y for greater than 100days. Radionuclidecompounds have been assigned toone of these classificationsby ICRP (ICRP 1980).TheICRP assignmentsto volubilityclassificationsmay beappropriate for use in dose calculations unless ex-perimentaldata to the contrary can be provided.

ExistingDOEordersor NRC regulatoryguidesdo notspecifythat the exposedpersonbe anyoneother than anadult receiving the highestdose as a result of an acci-dent. Past practice has been to assume ICRP referenceman as the receptor of accident-induced dose (ICRP1974). A change in this practice is not considerednecessary,although each case should be reviewed onindividual merits. Exposure to radioiodine causeshigherdose in thyroids of infmts and children than in

adultsand may be a specialcasewhichdeservesdescrip-tion and evaluation.

Breathingrate of the exposedperson maybe basedonthe standard rates established by ICRP for referenceman (ICRP 1974).The standard rate of 3.5 X 10-4m3/scorresponding to light activity is recommended fordoses received as a result of an accident under 8-hduration.

H. Quality

A singlequality factor (QF) for alpha radiation is notconsistentlyused throughoutthe DOE, allowinga factorof 2 differencein dose.The followinglist from ICRP 26(ICRP 1977)contains the recommended quality factorsfor all radiation types. The quality factor for neutronsmay be vaned if the averageenergyor energyspectrumis known.

20Alpha-recoil 20Betaor electron 1Gamma 1Fissionfragment 20Fissionneutron 10

Decay scheme data presently in use with ICRP 30modeldosecalculationsare contained in ICRP Publica-tion 38 (ICRP 1983)as calculated by an ORNL com-puter code (Dillman 1980).Other sources of updateddata to consider are the ENDF/B fission product data(Rose 1976)and the ORNL radioactivedecayhandbook(Kocher 1981).The importance of this updated data inimprovingaccuracyof dose calculationsis unknown.

Calculationof an accident dose conversion factor isusuallybased on a short-termintake of the nuclide(lessthan 8 h). ICRP 30doseconversionfactorsare based on50-yrdose receivedafter intake of a unit amount of theradionuclide and are therefore compatible with acuteinhalationexposure.

Dose accumulation times should be selected to belong enough to account for most of the dose. For thepurposesofthis Guide, the 50-yrperiodcommonlyusedas a dose accumulation time, roughly equivalent toaverageoccupationallifetime, is used. Fitly years mayalso be assumed to equal or exceed the remaining life-

54

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time of the individual of average age in an exposedoffsitepopulation.Intermediatetimes maybe usefulforillustrativepurposes,but fordosecalculatedto comparewith proposed DOE Order 6430.1A dose criteria, 50 yrhasbeen used.

Effectivedose equivalent is used when dose to mul-tiple organs from a singleexposure is postulated. Themethodologyfor calculating this dose for comparisonwith whole-bodydose limits is simple:

D.= ~ (individual organdose, Di)X (individualorganweightingfactor,Wi)

Individual organ weightingfactors may be the ICRPweightingfactors(ICRP 1977).

I

Two modelsare commonlyusedfor calculatingdirectdose from cloud immersion: the finite plume and thesemi-infiniteplume models (Slade 1968).Although thefiniteplume model is more complexand requires morecomputer time to run, it has the advantage of greateraccuracyat distancesnearer to the accidentsite (Wenzel1982).The results of both models converge when theplume is relativelylarge (compared with the mean freepath ofgamma photons)and has diffised to the ground.At relatively short downwind distances, the semi-in-finite model overestimates the dose for ground-levelreleases during stable meteorological conditions andunderestimates the dose from elevated releases. It issuggestedthat use of the semi-infinitecloud model belimited to calculations of dose beyond 10 km if therelease is elevated (Wenzel 1982) or that a “finite”plume correction factor be applied to calculations ofclose-indoses(Strenge1980).The finiteplume model ispreferred under most conditions and is available inseveral computer codes (RSAC-3,Wenzel 1982;SUB-DOSA, Strenge 1975). Codes using the semi-infinitemodel are also available (EXREM HI, Trubey 1973;RSAC-3,Wenzel 1982).

Immersiondosemodelsusuallyprovidea doserate atthe body surface. For the purpose of calculating animmersiondoseadditiveto whole-bodydose from othersources, a dose at 5-cm tissue depth has been recom-mended (NBS 1954,NRC 1977).Total body dose at 5-cm tissuedepth maybe calculatedfrom the surfacedoseas described by Strenge (1980). ICRP 30 calculates

whole-bodydosefrom immersionby a differentmethod(Poston 1974),whichis also appropriate.Gamma dosesto other organs from cloud immersion do not varygreatlyfrom the whole-body(5-cm)dose,as observedbyStrenge(1980).If skin dose is the major contributor todose (airbornebeta or weak gamma emitters), the skindose shouldbe the sum of surfacegamma dose and thebeta doseat a depth of 7 mg/cm2(NRC 1977).

Ingestionofwatercontaminatedby accidentalreleaseof radioactivematerial to nearby lakesand streams is apossiblesource of dose. However, neither ingestionbythe drinkingwater pathway or by the cow-milk-humanor crop-food-humanpathwayshave been confirmed tobe major contributors to dose from a credible DBA.Althoughit appears unlikelythat ingestiondose wouldapproach the limits on radiation dose proposed forChapter I of DOE Order 6430.1,guidanceon ingestiondose is discussed in this appendix on the assumptionthat ingestion dose may fall in the category of “otherconsequencesto be considered” (Section V.G. of theGuide).

The accident leadingto ingestiondose is assumed tobe a grossleakofa liquid contaminated with radioactivematerial, such as tritium, into nearby lakes or streams.An internaldosimetrymodel such as the GI tract modelof the ICRP (1979)is suitable for calculatingdose ftomdirect radiation of the GI tract and metabolic transferfrom the small intestine to other organs. INREM II(Killough1978)is a suitablecomputercode for calculat-ing ingestiondose conversionfactors.Dispersionof theradionuclidescan be estimated accordingto RegulatoryGuide 1.113 (NRC 1977A);doses ftom ingestion ofcontaminated water can be estimated accordingto Reg-ula.to~ Guide 1.109 (NRC 1977B).An acute liquidrelease model for tntium and other radionuclides hasbeen prepared for the Savannah River Plant by Huang(1983).

Bair 1979:W.J. Bair,“Reviewof Report of the Intern-ationalCommission on Radiological Protection Com-mittee 2: Limits for Intakes of Radionuclidesby Work-ers,” BattellePacificNorthwest Laboratories(1979).

Dillman 1980:L. T. Dillman, “EDISTR-A ComputerProgram to Obtain a Nuclear Decay Data Base forRadiation Dosimetry,”Oak RidgeNational Laboratoryreport ORNL/Th4-6689(1980).

5s

Elder 1974:J. C. Elder,M. Gonzales,and H. J. Ettinger,“Plutonium AerosolSizeCharacteristics,”HealthPhys-ics27,45-53 (July 1974).

Eve 1966:I. S. Eve, “A Reviewof the Physiologyof theGastrointestinal Tract in Relation to Radiation Dosesfrom Radioactive Materials,” Health Physics

Huang 1983:J. C. Huang, “Analysis Aids for Conse-quence Calculationsof Postulated SRP Surface WaterReleases,” E. I. DuPont de Nemours memorandumDPST-83-870(1983).

ICRP 1959:“Report of Committee 2 on PermissibleDose for Internal Radiation,” International Com-mission on Radiological Protection, Publication 2(1959).

ICRP 1966: “Deposition and Retention Models forInternal Dosimetry of the Human Respiratory Tract,”International Commission on RadiologicalProtectionreport in HealthPhysics12 (2), 173-207(1966).

ICRP 1974:“Report of the Task Group on ReferenceMan,” International Commission on RadiologicalProtection,Publication23 (October 1974).

ICRP 1977:“Recommendationsof the ICRP,” Intern-ational Commission on Radiological Protection,Publication26 (1977).

ICRP 1979:“Limits for Intake of Radionuclides byWorkers,” International Commission on RadiologicalProtection,Publication30,Part 1(July 1979).

ICRP 1980: “Limits on Intake of Radionuclides byWorkers,” International Commission on RadiologicalProtection,Publication30,Part 2(1 980).

ICRP 1983:“Radionuclide Transformations-Energyand Intensityof Emissions,”InternationalCommissionon RadiologicalProtection,Publication38 (1983).

Killough 1978:G. G. Killough,D. E. Dunning, Jr., andJ. C. Pleasant, “INREM II: A Computer Implementa-tion of Recent Modelsfor Estimatingthe Dose Equiva-lent to Organs of Man from an Inhaled or IngestedRadionuclide,” US Nuclear Regulatory Commissionreport NUREG/CR-01 14, ORNL/NUREG/TM-84(1978).

Kirchner 1966:R. A. Kirchner, “A Plutonium ParticleSize Study in Production Areas at Rocky Flats,”AmericanIndustrialHygieneAssociationJournal(July-August1966).

Kocher 1981:D. C. Kocher, “Radioactive Data DecayTables,” Oak Ridge National Laboratory reportDOE/TIC-l 1026(1981).

Mishima 1965: J. Mishima, “Plutonium ReleaseStudies.I. Releasefrom Ignited Metal,” BattellePacificNorthwest Laboratories report BNWL-205(December1965).

Mishima 1966: J. Mishima, “Plutonium ReleaseStudies.11.Releasefrom Ignited Bulk MetallicPieces,”Battelle Pacific Northwest Laboratories reportBNWL-357(November 1966).

Mishima 1968:J. Mishima,L. C. Schwendiman,and C.W. Radasch, “Plutonium ReleaseStudies. III. Releasefrom Heated Plutonium BearingPowders,”BattellePa-cific Northwest Laboratories report BNWL786 (July1968).

Mishima 1973:J. Mishima, “The Fractional AirborneReleaseof DissolvedRadioactiveMaterials During theCombustion of 30 Percent Tri-N-ButylPhosphate in aKerosene-Type Diluent,” Battelle Pacific NorthwestLaboratoriesreport BNWL-B-274(June 1973).

NBS 1954:“PermissibleDose from ExternalSourcesofIonizing Radiation,” National Bureau of StandardsHandbook 59 (September1954).

NRC 1977:“AssumptionsUsed for Evaluatingthe Po-tential RadiologicalConsequenceof AccidentalNuclearCriticality in a Fuel ReprocessingPlant,” US NuclearRegulatoryCommission Regulato~ Guide 3.33 (April1977).

NRC 1977A:“Estimating Aquatic Dispersion of Ef-fluents from Accidentaland Routine Reactor Releasesfor the Purpose of Implementing Appendix I,” USNuclear Regulatory Commission Regulatory Guide1.113(April 1977).

NRC 1977B:“Calculation of Annual Doses to Manfrom Routine Releases of Reactor Etlluents for thePurpose of Evaluating Compliance with 10 CFR 50,”US Nuclear RegulatoryCommissionRegulatoryGuide1.109(October 1977),App. I.

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56

Poston 1974:J. W. Poston and W. S. Snyder,“A Modelfor Exposure to a Semi-Infinite Cloud of a PhotonEmitter,” HealthPhysics26,287-293(April 1974).

* Raabe 1978:O. G. Raabe, S. V. Teague, N. L. Rich-2 ardson, and L. S. Nelson, “Aerodynamic and Dissolu-

tion Behavior of Fume AerosolsProduced During the

*’ Combustion of Laser-Ignited Plutonium Droplets inAir,” HealthPhysics35,663-674 (November 1978).

Rose 1976:P. F. Rose and T. W. Bumows,“ENDF/BFission Product Decay Data,” Brookhaven NationalLaboratoryreport BNLNCS-50545(August1976).

Runkle 1981:G. E. Runkle and J. K.. Soldat, “Com-parisonof ICRP 2 and ICRP 30 for Estimatingthe Doseand Adverse Health Effects from Potential Radio-nuclide Releases from a GeologicWaste Repository,”Sandia National Laboratories report SAND 81-2163C(1981).

Selby1975:J. M. Selby,E. C. Watson,J. P. Corley,et al.,“Considerationsin the Assessmentof the Consequencesof Eflluents flom Mixed Oxide Fuel FabricationPlants,” BattellePacificNorthwest Laboratories reportBNWL-1697(June 1975).

Slade 1968:D. H. Slade,Ed., “Meteorologyand AtomicEnergy,” US Atomic Energy Commission reportTID-24190(July 1968).

Strenge 1975:D. L. Strenge, E. C. Watson, and J. R.Houston, “SUBDOSA-A Computer Program for Cal-culating External Doses from Accidental AtmosphericReleasesof Radionuclides,”BattellePacificNorthwestLaboratoriesreport BNWL-B-351(June 1975).

Strenge1980:D. L.Strenge,“ModelsSelectedfor Calcu-lation of Doses,Health Effectsand EconomicCostsdueto Accidental Radionuclide Releases from NuclearPower Plants,” US Nuclear Regulatory Commissionreport NUREG/CR-1021 (PNL-3108)(May 1980).

Trubey 1973: D. K. Trubey and S. V. Kaye, “TheEXREM III Computer Code for Estimating ExternalRadiation Doses to Population from EnvironmentalReleases,” Oak Ridge National Laboratory reportORNL/TM-4322(Deeember 1973).

Wenzel 1982:D. R. Wenzel, “RSAC-3-RadiologicalSafety Analysis Computer Program,” Exxon NuclearIdaho Companyreport ENICO-1OO2(April 1982).

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57

F

I

Several computer codes containing the primary in-halation dose models have been used to calculate in-halation dose from radiologicalaccidents.These codesand their major featuresare summarized in Table F-I.The objectivesof this effortwere to

● @in experiencewith the major codes now in useelsewhere;

● calculate doses with codes of similar capabilitiesusinga semistandard set of input representativeofmajor postulatedaccidents;

. intercompare results from these codes to detectmajor differenceswhich might lead to inconsisten-ciesin the sitingand major designfeatureprocess;

● identifi code5considered most use~ for a~identanalysis;and

. identifi the codes written for chronic dose a55e55-ment whichare adaptable to the accidentcase.

Results of this effort are preliminary and are sum-marized in the followingsections.

Standard input for the codes listed in Table F-I wascompiled for two postulated releases:an instantaneouselevated (p.@ release of 239Pu02and an 8-h ground-level release of mixed fission products. Those codescapable of dispersion calculation were provided com-mon meteorologicaldata and depletiondata. Other datasuch as particle size, uptake time, and dose accumula-tion time werealso common.

* .&

Results from dispersion and dose calculations are ,-presented in Table F-II. Those codes only capable ofdose calculationwere supplied with a common ~Q (at .*10000 m) from which to calculateorgan doses. Theseresultscan be summarizedas follows:

● The cause of a problem with the elevated dis-persion calculation by the HADOC code was notreadily located;RSAC-3and DACRIN ~Q valueswere in good agreement;and CRAC-2~Q valueswere low by a factor of 4 for the ground-levelreleasebecausean expansionfactorproportional tothe time of releasewas includedin CR4C-2 but notin the other codes.

. Lung doses were in good agreement amongHADOC, RSAC-3, DACRIN, and ICRP afterquality factor differences were accounted foqCRAC-2and INREM II were unaccountablylowerin PU02dose.

. Bonedoseswerecalculatedwithin factorsof2-4 byall codes except CRAC-2, which was unaccoun-tably lower.

. Liver doseswere in good agreement(a factor of 2)among HADOC, DACRIN, ICRP, and INREM II.

. Thyroid doses were not valid in this comparisonbecauseof poor choiceof radioiodineamounts.

Experience with these codes (HADOC, DACRIN,ICRP, and CRAC-2at Los Alamos and RSAC-3by D.Wenzelat INEL) showedall wereadaptable to input forthis simpleaccidentcasewith the exceptionof CRAC-2.CR4C-2 waswritten for analysisof health and environ-mental consequencesand mitigationeffectsrelated to anuclearreactoraccidentand is too largeand complextoallow the dose discrepanciesnoted above to be readilyreconciled, nor is the code amenable to running asimplifiedproblem.

*

D

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58

TABLEF-I.D1SPERS1ONANDDOSECODESE n

m CI L C R R

CRAC-2 A A A A A A ~Nb c R SR R R R R R 1 m

H A A A A 7600 S Dm

D A A H mR 2 m

I A A Flv K 1 O S

R R I mI(ORNL) A m W O I

mR A A A A A A A FIV W I m

R R R R R R R

A = a cr R = r r

% NN L aA lP P N LO R N L I N E L

% Vp rl% r ee c 7 c A cd r nT G L D m

x/Q (dm3) Whole

x 10-1’J 4.0 x 10-17 1.9x 10+(2.7X 10-7)b (9.OX10-14) (4.2X 10+)

RSAC-3n 2.8 X 10-7 3.1 x 10+ 4.4 x 10-sDACRINS 2.7 X 10-7 2.6 X 10+ 4.6 X 10+CRAC-2’ 1.9x 10-7 ... 1.4x 10+

(2.7X 10-7)b – (2.0x 10-s)ICRP (2.7X 10-7)b – (1.0x 10-’)INREMII*’ (2.7X 10-7)b -- (2.8X 10+)

G

6.5 X 10-8(1.5x 10-’)1.3x 10-’7.9 x 10+

—

(3.1 ;10+)(8.7X 10+)

3.9 x 10-8(8.8x 10--s)1.7Xlo %3.7 x 10+

—-

(6.9 =10+)(3.8X 10+)

Xx 10+ 0.19 4.5

DACRIN 4.6 X 10+ 0.40 3.3cIuc-2 1.2x 10+ 1.6X 10-2 0.62

(4.6X 10+)b (6.1X 10-2) (2.4)Imp (4.5x lo+)b – (6.3)INREMIIC (4.5 x 10-y – (6.0)

4.3 —-3.1 —2.3

4.1 x 10-2 :(0.16) –(3.7) –— —

———.———-.—

7.3 x 10+2.7 X 10+1.4x 10-’2.9 X 10-3(1.1x 10-2)(1.8X 10-2)’(2.5X 10-2)d

‘ c u QF= 10 =

INREM11dosefactorst KUnrdkticdh low).

REFERENCES

Houston 1974:J. R. Houston, D. L. Strenge,and E. C.Watson, “DACRIN-A Computer Program for Calcu-latingOrgan Dose from Acuteor ChronicRadionuclideInhalation,” Pacific Northwest Laboratories reportBNWLB-389 (December 1974).

Killough1978A:G. G. Killough,D. E, Dunning,Jr., andJ. C. Pleasant, “INREM II: A Computer Implementa-tion of Recent Modelsfor Estimatingthe Dose Equiva-lent to Organs of Man from an Inhaled or IngestedRadionuclide,” US Nuclear Regulatory Commissionreport NUREG/CR-01 14, ORNL/NUREG/TM-84(1978).

Killough 1978B:G. G. Killough,D. E. Dunning, Jr., S.R. Bernard, and J. C. Pleasant, “Estimates of InternalDose Equivalentto 22 TargetOrgans for RadionuclidesOccurringin Routine Releasesfrom Nuclear Fuel CycleFacilities,”US Nuclear RegulatoryCommission reportNUREG/CR-0150, ORNL/NUREG/TM-190 (June1978).

,.

NRC 1975:“Calculation of Reactor Accident Conse-quences,”ReactorSafetyStudy,US Nuclear RegulatoryCommission report WASH-1400 (NUREG 75/014)(October 1975),App.VI.

Ritchie 1982:L. T. Ritchie, J.Blond, “CRAC2—ComputerSandia National LaboratoriesNUREG/CR-2326(1982).

D. Johnson, and R. M.Code Users’ Guide,”report SAND 81-1994,

Strenge 1981: D. L. Strenge and R. A. Peloquin,“HADOC-A Computer Code for Calculation of Ex-ternal and Inhalation Doses from Acute RadionuclidesReleases,” Pacific Northwest Laboratories reportPNL3503 (1980). s

Watson 1980:S. B. Watson and M. R. Ford, “A User’s —Manual to the ICRP Code—A Series of Computer -+pro9am8 to pefiorm Dosimetric Calculations for the --ICRP Committee2 Report,” Oak RidgeNational Labo- =ratory report ORNL/TM-6980(February 1980).

Wenzel 1982:D. R. Wenzel, “RSAC3—RadiologicalSafety Analysis Computer Program,” Exxon NuclearIdaho report ENICO-1002(April 1982).

60

G

P E.:*

INTRODUCTION

Potential for dispersingradioactive material into theenvironment could affect both the site selection andmajor design features of a nuclear facility. Long-termradiation dose to the population in the region, majorcleanupcosts,and lossof production at nearby facilitiesare potential consequences. For the purposes of theGuide, it is assumed that contamination would becleanedup to acceptablelimitsand long-termdose fromresidualcontamination would not be a major concern.Lossof production at a nearby facilityis a site-specificmatter not easily handled in a generalized guidancedocument. Therefore, discussion has been limited todecontaminationcosts,which have receivedonly mini-mal attention in most accidentanalyses.This appendix,althoughreflectinga generalshortageof specificdecon-tamination cost information, may be of help to theanalyst when environmental contamination concernsare considered.

II. DECONTAMINATIONPARAMETERS

A. R

-%.

Determination of which radionuclideswould be im-portant if the environment wereseriouslycontaminatedby an accidental release depends on original amountradioactive half-life,and dose conversion factor of theradionuclide.Methods of determination are discussedin SectionV.A.1.of the Guide.

The originalsourcestrengthprovidesan indicationofthe severity of the spill or release. Decontaminationcosts are generallyproportional to the contaminationsurface concentration, which is proportional to theamount releasedat the source(Finley 1979).

The chemical form is important because decon-tamination methods depend on the decontaminationfactors. Ability to remove the radionuclide affects thecost of decontamination; that is, a surface could be

decontaminated by physical or chemical cleaning ortotally removed, such as removing a section of pave-ment withjackhammers.

D.

The dispersalmechanismdeterminesthe distributionof the concentration levels to be expectedand the uni-fomlity over the contaminated areas.

E. Evacuation

Short-term decisions on which areas to evacuate orrestrict usage affect decontamination costs. Securityneedsand costsbefore,during,and perhapsafter decon-tamination depend on the land use pattern and con-tamination levelsinvolved.Even after extensivedecon-tamination, some contamination will remain in theenvironment. Surveillance to follow long-term move-ment through the food chains may be needed. Land usebefore decontamination may be restricted. Certaincrops, livestock,and produce (milk, for example)couldbe purchased and disposed of. Livestock foragingmaybe restricted from some areas.

D

The decontamination factor is the amount of radio-nuclide per unit area before decontamination dividedby the amount remaining after decontamination. Alargevariety of decontamination methods is available,rangingfromcarefulhand scrubbingto wet sandblasting(Fore 1982,Chester 1981).Usually the simpler, inex-pensive methods work for large areas. Removal anddisposalof a surfacemaybe more cost effectivethan itsdecontamination, as is the case for soil (Menzel 1971,James 1973).Decontamination data for radionuclidechemicalforms encounteredare needed for a variety ofsurfaces and liquids. Table G-I provides examples ofdecontamination factors developed for PUOZ(Wenzel1982).

G. Economic PopulationInformationfortheArea

The building replacement,land, and crop valuescanbe obtained from US Department of Agriculture

61

TABLE G-I. DECONTAMINATIONFACTORSFORPuOZ’

GeneralDescriptionof Method DecontaminationFactor

2

4d

2100

2300

.

100305010

W 1 F 1 M 1 J 1 M 1 S 1 1 W1

economists.Dependingon the depth of the analysis,aneconomic activity assessment of the area would beneeded to estimate evacuation and land use denial costestimates. The accuracy of the analysis in general de-pends on the quality of the population density data foreach contamination leveland land use typechosen andthe detailofmonitoringperformedbeforethe operation.

D E

Wastedisposalis a major cost. Whethersolidwasteisretrievable (>100 nCi/g transuranic) or nonretrievabledetermines the waste disposal method. Liquid wasteswill also be disposed of differently, depending on thelevel of contamination. Worker radiation dose isdirectly proportional to radionuciide concentration inthe waste. Packagingand attendant cost is related towaste radionuclide concentration. The transport dis-tance for offsitedisposalshouldbe considered.

The basicdecontaminationtasksand costsdependonland uses of the contaminated area. Table G-II is ageneralized listing of typioai land use types foragriculturalareas, town, and central city. Tables G-IIIand G-IV give building parameter and land use frac-tions helpful for cost estimation. Detailed analysiswould require site-specificdata rather than the gen-eralizeddata in TablesG-II through G-IV.

D

Decontamination cost/unit area or per capita tablesfor each land use pattern (agricultural, town, centralcity) and decontamination level should consider thefollowing:

●

●

●

●

●

●

●

initial radiologicalsurvey of entire contaminatedarea;land usedenial and producepurchase;perimeter,urban, and long-termsecurity;radiation surveys-precleanup, during cleanup,finalcertification,surveillance;decontamination tasks and costs for each land usearea;restoration of topsoil and fertility, lawns, streets,buildings,property,and recreationalpotential;andonsitedecontaminationand restoration.

For illustration,Table G-V givescostsof three land useareasand three decontaminationlevelsas DFs for PU02contamination.

d A-

d -

MC

A

+a

TownorRural Satellite Central

PopulationandLandUse Agriculture City City

Populationdensity(people/acre) 0.01 3.0 16.0

Landuse(fraction/acre)Single-familyresidencesResidences<6 floorsResidences>6 floorsCommercialbuildingsPublicbuildingsParksandcemeteriesUndevelopedlandAgriculturalland

b

--0.050.10.8

0.30.10.050.10.050.20.10.1

0.20.10.10.20.20.10.50.05

b

TABLE G-III. BUILDINGAND DWELLINGPARAMETERESTIMATES

S S

story) Building

Buildings/acreFamilies/acreLotsize(ftz)Streetarea(ftz)Driveway/parkinglot area(ft’)Floorareaperfloor(ft’)Numberof families/floorOpenareaandlawn(ft’)Interiorhorizontalarea(ft’)

11

400003560

1000

136500

2500

55

72601450

3002000

14960

2000

530

72601450

26002

2000

39000

560

72601450

32602600

21000

78000

50

72601450

32602600

01000

78000

from particulate deposition. Some assumptions are0 needed for the nonhomogeneityof close-indeposition.

m can be c

Y— ck c

B. Isopleth/LandUse Category

The contamination isoplethsand the 80-kmgrid canbe scaled and overlaid on topographical or Landsatmaps (see Stephen 1979) to estimate areas of un-developed,agricultural,suburban,commercial,or otherland use patterns. Once the fraction of each land usetype is estimated for each isopleth, then populationcensus data can be used to estimate the population ineach land use fraction.

6

TABLE G-fV. BUILDINGTYPEAND LAND USE FRACTIONESTIMATES

*AreaPlus Occupied .-%

LandUse Type (m2) Driveway Lawn byBuilding ~.

Ruralsingle-familyresidence’ 4047 0.11 0.84 0.05 *Suburbansingle-familyresidence 809 0.20 0.57 0.23<6-Storysuburbanapartment

-809 0.47 0.23 0.30

>6-Storysurburbanapartment 809 0.54 0.12 0.34Commercialorindustrialarea — 0.50 0.05 0.45Publicbuildingarea — 0.50 0.20 0.30Parksandcemeteries — 0.05 0.90 0.05Undevelopedandagricultural” —- 0.025 0.95 0.025

C. AccidentDecontaminationCost

The total cost for the accident in dollars can becalculated by multiplying the area within an isoplethland use category(mz)times the decontamination costper unit area and contamination level for that land usecategory(dollars/m2)and summing these for each landuse category and isopleth. The total cost can then beadjusted for inflation for the year of concern.

REFERENCES

Chester 1981:C. V. Chesterand E. A. Cristy,“Environ-mental Decontamination-Proceedings of the Work-shop,” December 4-5, 1979, Oak Ridge, Tennessee,CONF-791234(February 1981).

Cobb 1973:F. C. Cobb and R. L. Van Hemert, “SourceBookon Plutonium and its Decontamination,”DefenseNuclearAgency,Kirtland Air Force Base,New Mexico,report DNA-3272T(September 1973).

Finley 1979:N. C. Finley,D. C. Aldrich,S.L. Daniel,D.M. Encson, C. Henning-Sacks,P. C. Kaestner, N. R.Ortiz, D. D. Sheldon, J. M. Taylor, and S. F. Herrei~“Transportation of Radionuclides in Urban Environ-ment: Dratl Environmental Assessment,” Sandia Na-tional Laboratories report SAND 79-0369(NUREG/CR-0743)(July 1979).

Fore 1982:C. S. Fore, R. A. Faust, and R. H. Brewster,“Cleanup and Treatment of Radioactively Con-taminated Land Including Areas Near Nuclear Facili-ties—A Selected Bibliography,”Oak Ridge NationalLaboratoryreport ORNL/EIS-199(September 1982).

James 1973:P. E. James and R. G. Menzel, “Researchon Removing Radioactive Fallout from Farmland,”Technical Bulletin No. 1464 (USDA, AgriculturalRe-searchService,May 1973).

Luna 1969:R. E. Luna and H. W. Church,“DIFOUT: AModel for Computation of AerosolTransport and Dif-fusion in the Atmosphere,” Sandia National Laborato-ries report SC-RR-68-555(January 1969).

McGrath 1975: P. E. McGrath, “Consequences of aMajor Nuclear Reactor Accident After Initial Depo-sitionofReleasedRadioactivity,”SandiaNational Lab-oratoriesdraft report (Division 1721)(1975).

Menzel 1971:R. G. Menzel and P. E. James, “Treat-ments for Farmland Contaminated with RadioactiveMaterial,” Agriculture Handbook No. 395 (USDA,AgriculturalResearchService,June 1971). ●

aNRC 1975:“Calculation of Reactor Accident Conse- -quences,” ReactorSafety Study,” Nuclear Regula- ‘yt o r y Commission report WASH 1400 ~-(NUREG-75/014)(October 1975),App. VI. Q.

64

T D

D e

s1200

— —

s s

n

r2

nr

z

2

r

r

300-500780-1600

269-480500 500 500400

1150

1 24

“ i c b

6

D

( 1975) (Smith1978) (Finley1979) (Wenzel1982) aAgricultureLand -b

DFs -900-4 900/acre — -600-720/acre -t10s s — — *

= — — —0

s - 1 — —s s - 8 —

— — —

s - —s - —

a — — —

Smith 1978:C. B. Smith and J. A. Lambert, “Tech- Wenzel 1982:W. J. Wenzeland A. G. Gallegos,“Sup-nologyand Costs for CleaningUp Land Contaminated elementary Documentation for an Environmental Im-with Plutonium,” in Selected Topics:TransuranicEf pact Statement Regarding the Pantex Plant: Decon-jluents in the General Environment,Technical Note lamination Methodsand Cost Estimatesfor PostulatedCSD-78-1(US Environmental Protection Agency,Of- Accidents,” Los Alamos National Laboratory reportfice of Radiation Programs, Washington, DC, April LA-9445-PNTX-N(1982).1978).

Stephen 1979:J. Stephen, H. Foote, and V. Coburn,“Well Location and Land-Use Mapping in the Colum-bia PlateauMea,” RockwellHanfbrd Operations reportRHO-BWI-C-61,PNL3295 (October 1979).

ATTACHMENTA

COMMENTSHEETA GUIDETO RADIOLOGICALACCIDENTCONSIDEIL4TI:ONSFOR

SITINGAND DESIGNOF DOE NONREACTORNUCLEARFACILITIES

%“ImsAlamosNationalLaboratoryreportLA-10294-MS

CommentTopic: Date:*% Commenter: Address:

: -T .UIWIUII: rllullc:

;e/Section D

.

●’

— IAddressee:GroupLeader,GroupHSE-1,IAMAlamosNationalLaboratory,PO Box 1663,IAMAlamos,NM

6

4m

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