Kalinin VVER-IOOO Nu(~learPower Station Unit 1 PRA

Procedure Guides for aProbabilistic Risk Assessment

English Version

Manuscript Completed: May 2005Date Published: December 2005

Sponsored by the Joint Cooperative Program Between the Ge,vernments oftQeUnited States and Russia

The BETA Project

Brookhaven National LaboratoryUpton, NY 11973-5000

Prepared forDivision of Risk Analysis and ApplicationsOffice of Nuclear Regulatory ResearchU.S. Nuclear Regulatory CommissionWashington, DC 20555-0001NRC Job Code R2001

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Nuclear Regulatory CommissionExhibit # - PWA000020-00-BD01Docket # - 05000293Identified: 02/22/2011

Admitted: Withdrawn: Rejected: Stricken: 02/22/2011

3. Technical Activities

3.3.5.5 References

EPRI, "MMP4 - Modular Accident AnalysisProgram for LWR Power Plants,· RP3131-02,Volumes 1-4, Electric Power Research Institute,1994.

Summers, R., M, eta!., "MELCOR Computer CodeManuals - Version 1.8.3," NUREG/CR-6119,SAND93-2185, Volumes 1-2, Sandia NationalLaboratories, 1994.

NRC, ·Severe Accident Risks: An Assessment forFive U.S. Nuclear Power Plants,· NUREG-1150,U.S. Nuclear Regulatory Commission, December1990.

Harper, F. T., et ai, ·Evaluation of Severe AccidentRisks: Quantification of Major Input Parameters,·NUREG/CR-4551, Volume 2, SAND86-1309,Sandia National Laboratories, December 1990.

NRC, "Individual Plant Examination: SubmittalGuidance,· NUREG-1335, U.S. Nucle:uRegulatory Commission, August 1989.

Jow, H. J., et a!., "XSOR Codes User Manual,"NUREGlCR-5360, Sandia National Laboratories,1993.

3.4 Level 3 Analysis(Consequence Analysisand Integrated RiskAssessment)

In this section, the analyses performed as part ofthe Level 3 portion of a probabilistic riskassessment (PRA) are described.

3.4.1 Assumptions and Limitations

In most level 3 (i.e., consequence) codes,abnospheric transport of the released material iscarried out assuming Gaussian plume dispersion.This assumption is generally valid for flat terrain toa distance of a few kilometers from the point ofrelease but is inaccurate both in the immediatevicinity of the reactor building and at fartherdistances. For most PRA applications, however,the inaccuracies introduced by the assumption ofGaussian plumes are much smaller than theuncertainties due to other factors, such as the

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source term. In specific cases of plant location,such as, for example, a mountainous area or avalley, more detailed dispersion models thatincorporate terrain effects may have to beconsidered. There are other physical parametersthat influence downwind concentrations. Drydeposition velocity can vary over a wide range .depending on the particle size distribution of thereleased material, the surface roughness of theterrain, and other factors. An assessmentoftheseuncertainties focused on the factors whichinfluence dispersion and deposition has beencarried out recently (Harper et aI., 1995). Earlierassessments of the assumptions and uncertaintiesin consequence modeling were reported in otherPRA procedures guides (NRC, 1983).

Besides atmospheric transport, dispersion, anddeposition of released material, there are severalother assumptions, limitations, and uncertaintiesembodied in the parameters that impactconsequence estimation. These include: modelsof the weathering and resuspension of materialdeposited on the ground, modeling of the ingestionpathway, i.e., the food chains, ground-crop-manand ground-crop-animal-dairy/meat-man, internaland external dosimetry, and the health effectsmodel parameters. other sources of uncertaintyarise from the assumed values of parameters thatdetermine . the effectiveness of emergencyresponse, such as the shielding provided by thebuilding stock in the area where people areassumed to shelter. the speed of evacuation, etc.Comparison of the results of differentconsequence codes, which embody differentapproaches and values of these parameters, on astandard problem are contained in a studysponsored by the Organization for Economic Co­operation and Development (OECD, 1994). Anuncertainty analysis of the COSYMA code resultsusing the expert elicitation method is currentlybeing carried out (Jones, 1996).

3.4.2 Products·

Documentation of the analyses performed toestimate the consequences associated with theaccidental release of radioactivity to theenvironment should contain sufficient informationto aflow an independent analyst to reproduce theresults. At a minimum, the following informationshould be documented for the Level 3 analysis:

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3. Technical Activities

3.4.3 Analytical Tasks

The consequence measures that focus on impactsto the environment include:

Groundwater contamination has yet to be includedin a Level 3 analyses, although it may be importantto consider it in certain specific cases.

The economic impacts are mainly estimated interms of the costs of countermeasures taken toprotect the population in the vicinity of the plant.These costs can include:

•

•

•

•

identification of the consequence code and theversion used to carry out the analysis,a description of the site-specific data andassumptions used in the input to the code,specifications of the source terms used to runthe code, anddiscussion and definition of the emergencyresponse parameters,a description of the computational processused to integrate the entire PRA model(Level 1 - Level 3),a summary of all calculated results includingfrequency distributions for each risk measure.

•

land contamination

surface water body (e.g., lakes, rivers, etc.)contamination.

A Level 3 PRA consists of two major tasks:

1. Consequence analyses conditional on variousrelease mechanisms (source terms) and

2. Computation of risk by integrating the resultsof Levels 1, 2, and 3 analyses.

Task 1 - Consequence Analysis

The consequences of an accidental release ofradioactivity from a nuclear power plant to thesurrounding environment can be expressed inseveral ways: impact on human health, impact onthe environment, and impact on the economy.The consequence measures of most interest to aLevel 3 PRA focus on the impact to human health.They should include:

number of early fatalities,

number of early injuries,

number of latent cancer fatalities,

population dose (person-rem or person­sievert) out to various distances from theplant,

individual early fatality risk defined in the earlyfatality QHO, i.e., the risk of early fatality forthe average individual within 1 mile tram theplant, and

• ... individual latent cancer fatality risk defined inthe latent cancer QHO, i.e'., the risk of latentcancer fatality for the average individual within10 miles of the plant

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short-term costs incurred in the evacuationand relocation of people during the emergencyphase following the accident and in thedestruction of contaminated food, and

long-term costs of interdicting contaminatedfarmland and residentiaVurban property whichcannot be decontaminated in a cost-effectivemanner, Le., where the cost ofdecontamination is greater than the value ofthe property.

The costs of medical treatment to potentialaccident victims are not generally estimated in aLevel 3 analysis, although approaches do exist forincorporating these costs (Mubayi, 1995) ifrequired by the application.

The results of the calculations for eachconsequence measure are usually reported as acomplementary cumulative distribution function.They can also be reported in terms of adistribution-for example, ones that show the 5thpercentile, the 95th percentile, the median, and themean.

A probabilistic consequence assessment (PCA)code is needed to perform the Level 3 analysis.Such codes normally take as input thecharacteristics of the release or source termprovided by the Level 2 analysis. Thesecharacteristics typically include for each specifiedsource term: the release fractions of the coreinventory of key radionuclides, the timing andduration of the release, the height of the release(i.e., whether the release is elevated or groundlevel), and the energy of the release. PCA codesincorporate algorithms for performing weather

3. Technical Activities

sampling on the plume transport in order to obtaina distribution of the concentrations and dosimetry ,which reflect the uncertainty and/or variability dueto weather. The codes also model variou~protective action countermeasures to permit amore realistic calculation of doses and healtheffects and to assess the efficacy of these differentactions in reducing consequences.

Several PCA codes are currently in use forcalculating the consequences of postulatedradiological releases. The NRC supports the useof the MACCS (Jow, 1990 and Chanin, 1993) andMACCS2 (Chanin and Young, 1997) PCA codesfor carrying out nuclear power plant Level 3 PRAanalyses. A number of countries in Europesupport the use of the COSYMA (KfK and NRPB,1991 and Jones, 1996) PCA code for their Level 3analyses.

PCA codes require a substantial amount ofinformation on the local meteorology, demography,land use, crops grown in various seasons, foodsconsumed, and property values. For example, theinput file for the MACes code requires thefollowing information:

Meteorology - one year of hourly data on:windspeed and direction, atmospheric stabilityclass, precipitation. rate, probability ofprecipitation occurring at specified distancesfrom the plant site, and height of theatmospheric inversion layer.

Demography - population distribution aroundthe plant on a polar grid defined by 16 angularsectors and user-specified annular radialsectors, usually a finer grid close to the plantand one that becomes progressively coarserat greater distances.

• Land Use - fraction which is land, land whichis agricultural, major, crops, and growingseason.

• Economic Data - value of farmland, value ofnonfarm property, and annual farm sales.

The MACCS User Manual (Chanin, 1990) and theMACCS2 User Guide (Chanin and Young, 1997)may be consulted for a complete description of thesite input data necessary.

In addition to site data, a PCA code should haveprovisions to model countermeasures to protect

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the public and provide a more realistic estimate ofthe doses and health effects following anaccidental release. The MACCS code requiresthat the analyst make assumptions on the valuesof parameters related to the implementation ofprotective actions following an accident The typesof parameters involved in evaluating these actionsinclude the following:

• delay time between the declaration of ageneral emergency and the initiation of anemergency response action, such asevacuation or sheltering; this delay time maybe site specific.

fraction of the offsite population whichparticipates in the emergency response action,

effective evacuation speed,

degree of radiation shielding provided by thebuilding stock in the area,

• projected dose limits for long-term relocationof the population from contaminated land, and

• projected 'Ingestion dose limits used tointerdict contaminated farmland.

The selected values assumed for the above (orsimilar) parameters need to be justified anddocumented since they have a significant impacton the consequence calculations.

In summary, the PCA code selected for thecalculation of consequences should have thefollowing capabilities:

incorporate impact of weather variability onplume transport by performing stratified orMonte Carlo sampling on an annual set ofrelevant site meteorological data,

allow for plume depletion due to dry and wetdeposition mechanisms,

allow for buoyancy rise of energetic releases,

include all possible dose pathways, externaland internal (such as cloudshine, groundshine,inhalation, resuspension inhalation, andingestion) in the estimation of doses,

employ validated health effects models based,for example, on (ICRP, 1991) or BEIR V

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3. Technical Activities

(National Research Council, 1990) dosefactors for converting radiation doses toearly and latent health effects, and

• allow for the modeling of countermeasures topermit estimation bf a more realistic impact ofaccidental releases.

The above-cited methods for estimatingconsequences are, in general, adequate foraccidents caused by internal initiating eventsduring both full power operation and shutdownconditions. However, forextemal initiating events,such as seismic events, certain changes may beneeded. For example, the early warning systemsand the road network may be disrupted so thatinitiation and execution of emergency responseactions may not be possible. Hence, in addition tochanging the potential source terms, a seismicevent could also influence the ability of the close-inpopulation to cany out an early evacuation. ALevel 3 seismic PRA should, therefore, includeconsideration of the impacts of different levels ofearthquake severity on the consequenceassessment

To use a consequence code, generally thefollowing data elements are required:

reactor radionuclide inventory,

accident source terms defined by the releasefractions of important radionuclide groups, thetiming and duration of the release, and theenergy and height of the release,

hourly meteorological data at the site asrecommended, for example, in RegulatoryGuide 1.23 (NRC, 1986), collected over oneor, preferably,' more years and processed intoa form usable by the chosen code,

site population data from census or otherreliable sources and processed in conformitywith the requirements of the code, i.e., toprovide population information for each areaelement on the grid used in the code,

site economic and land use data, specifying'. 'the important crops in the area, value and

extent of farm and nonfarm property,

defining the emergency responsecountermeasures, including the possible timedelay in initiating response after declaration of

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warning and the likely participation in theresponse by the offsite population.

Task 2 - Computation of Risk

The final step in a Level 3 PRA is the integration ofresults from all previous analyses to computeindividual measures of risk. The severe acCidentprogression and the radionuclide source termanalyses conducted in the Level 2 portion of thePRA, as well as the consequence analysisconducted in the Level 3 portion of the PRA, areperformed on a· conditional basis. That is, theevaluations of alternative severe accidentprogressions, resulting source terms, andconsequences are performed without regard to theabsolute or relative frequency of the postUlatedaccidents. The final computation of risk is theprocess by which each of these portions of theaccident analysis are linked together in a self­consistent and statistically rigorous manner.

An important attribute by which the rigor of theprocess is likely to be judged is the ability todemonstrate traceability from a specific accidentsequence through the relative likelihood ofalternative severe accident progressions andmeasures of associated containment performance(Le., early versus late failure) and ultimately to thedistribution of fission product source terms andconsequences. This traceability should bedemonstrable in both directions, i.e., from theaccident sequence to a distribution ofconsequences and from a specific level ofaccident consequences back to the fission productsource terms, containment performancemeasures, or accident sequences that contributeto that consequence level.

3.4.4 Task Interfaces

The current task requires a set of release fractions(or source terms) from the Level 2 analysis(Section 3.3) as input to the consequenceanalysis.

The consequences are calculated in terms of:(1) the acute and chronic radiation doses from allpathways to the affected population around theplant, (2) the consequent health effects (such asearly fatalities, early injuries, and latent cancerfatalities), (3) the integrated population dose to

3. Technical Activities

some specified distance (such as 50 miles) fromthe point of release, and (4) the contamination ofland from the deposited material.

The consequence measures to be calculateddepends on the application as defined in PRA

"Scope. Generally, in a level 3 analysis,' adistribution of consequences is obtained bystatistical sampling of the weather conditions atthe site. Each set of consequences, however, isconditional on the characteristics of the release (orsource term) which are evaluated in the level 2analysis.

An integrated risk assessment combines theresults of the levels 1, 2, and 3 analyses tocompute the selected measures of risk in a self­consistent and statistically rigorous manner. Therisk measures usually selected are: early fatalities,latent cancer fatalities, population dose, andquantitative health objectives (QHOs) of the U.S.Nuclear Regulatory Commission (NRC) SafetyGoals (NRC, 1986). Again, the actual riskmeasures calculated will depend on the PRAScope.

3.4.5 References

Chanin, 0.1., and M. l. Young, "Code Manual forMACCS2: Volume 1, User's Guide," SAND97­0594, Sandia National laboratories, March 1997.

Chanin, 0.1., etal., "MACCSVersion 1.5.11.1: AMaintenance Release of the Code," NUREG/CR­6059, Sandia National laboratories, October 1993.

Chanin, 0.1., et aI., "MELCOR AccidentConsequence Code System (MACCS), Volume 1,User's Guide," NUREGlCR-4691, Sandia Nationallaboratories, February 1990.

Harper, F. T., et aI., "Probabilistic AccidentConsequence Uncertainty Analysis, Dispersion,and Deposition Uncertainty Assessment,"NUREG/CR--6244, Sandia National Laboratories,1995.

ICRP, "1990 Recommendations of the "ICRP:Annals of the ICRP, Vol." 21, No. 1-3, ICRPPublication 60, Intemational Commission onRadiological Protection, Pergamon Press, Oxford,England, 1991.

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Jones, J. A, et aI., "Uncertainty Ailalysis onCOSYMA," Proceedings of the Combined 3rdCOSYMA Users Group and 2nd IntemationalMACCS Users Group Meeting, Portoroz, Slovenia, .41228-NUC 96-9238, KEMA, Amhem, theNetherlands, September 16-19,1996.

Jow, H. N., et aI., "MElCOR AccidentConsequence Code System (MACCS), Volume II,Model Description," NUREG/CR-4691, SandiaNational "laboratories, February 1990.

KfK and NRPB, "COSYMA - A New ProgramPackage for Accident Consequence Assessment, "CEC Brussels, EUR 13028,Kemforschungszentrum (Karlsruhe) and NationalRadiological Protection Board, 1991.

Mubayi, V., et al., "Cost-Benefit Considerations inRegulatory Analysis," NUREG/CR-6395,Brookhaven National laboratory, 1995.

National Research Council, "Health Effects ofExposure to low levels of Ionizing Radiation,"BEIR V, Washington, DC, 1990.

NRC, "Severe Accident Risks' An Assessment forFive U.S. Nuclear Power Plal'ts," NUREG-1150,Vol. 1, Main Report, U.S. Nuclear RegulatoryCommission, 1990. " .

NRC, "Safety Goals for the Operation of NuclearPower Plants, Policy Statement," Federal Register,Vol. 51, No. 149, U.S. Nuclear RegulatoryCommission, August 4, 1986.

NRC, ·Onsite Meteorological Programs,"Regulatory Guide 1.23, U.S. Nuclear RegulatoryCommission, April 1986.

NRC, ·PRA Procedures Guide - A Guide to thePerformance of Probabilistic Risk Assessments forNuclear Power Plants: NUREG/CR-2300, Vol. 2,U.S. Nuclear Regulatory Commission, 1983.

OECD, "Probabilistic Accident ConsequenceAssessment Codes, Second InternationalComparison", Organisation for EconomicCooperation and Development, Nuclear EnergyAgency, Paris, France, 1994~