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Exhibit Pwa 00020, NUREG/CR-6572, Rev. 1 BNL-NUREG-52534-R1 - Kalinin PRA, Procedure Guides for Probabilistic Risk Assessment
ML110030946
Person / Time
Site: Pilgrim
Issue date: 12/31/2005
From:
Brookhaven National Lab (BNL), Office of Nuclear Regulatory Research
To:
Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML110030939 List:
References
RAS 19374, 50-293-LR, ASLBP 06-848-02-LR, PWA 00020
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~(;/CR-6572,Rev.l B~-~{;-52534-Rl Kalinin VVER-IOOO Nu(~lear Power Station Unit 1 PRA Procedure Guides for a Probabilistic Risk Assessment English Version Manuscript Completed: May 2005 Date Published: December 2005 Sponsored by the Joint Cooperative Program Between the Ge,vernments oftQe United States and Russia The BETA Project Brookhaven National Laboratory Upton, NY 11973-5000 Prepared for Division of Risk Analysis and Applications Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code R2001

3. Technical Activities 3.3.5.5 References source term. In specific cases of plant location, such as, for example, a mountainous area or a EPRI, "MMP4 - Modular Accident Analysis valley, more detailed dispersion models that Program for LWR Power Plants,* RP3131-02, incorporate terrain effects may have to be Volumes 1-4, Electric Power Research Institute, considered. There are other physical parameters 1994. that influence downwind concentrations. Dry deposition velocity can vary over a wide range .

Summers, R., M, eta!., "MELCOR Computer Code depending on the particle size distribution of the Manuals - Version 1.8.3," NUREG/CR-6119, released material, the surface roughness of the SAND93-2185, Volumes 1-2, Sandia National terrain, and other factors. An assessmentofthese Laboratories, 1994. uncertainties focused on the factors which influence dispersion and deposition has been NRC, *Severe Accident Risks: An Assessment for carried out recently (Harper et aI., 1995). Earlier Five U.S. Nuclear Power Plants,* NUREG-1150, assessments of the assumptions and uncertainties U.S. Nuclear Regulatory Commission, December in consequence modeling were reported in other 1990. PRA procedures guides (NRC, 1983).

Harper, F. T., et ai, *Evaluation of Severe Accident Besides atmospheric transport, dispersion, and Risks: Quantification of Major Input Parameters,* deposition of released material, there are several NUREG/CR-4551, Volume 2, SAND86-1309, other assumptions, limitations, and uncertainties Sandia National Laboratories, December 1990. embodied in the parameters that impact consequence estimation. These include: models NRC, "Individual Plant Examination: Submittal of the weathering and resuspension of material Guidance,* NUREG-1335, U.S. Nucle:u deposited on the ground, modeling of the ingestion Regulatory Commission, August 1989. pathway, i.e., the food chains, ground-crop-man and ground-crop-animal-dairy/meat-man, internal Jow, H. J., et a!., "XSOR Codes User Manual," and external dosimetry, and the health effects NUREGlCR-5360, Sandia National Laboratories, model parameters. other sources of uncertainty 1993. arise from the assumed values of parameters that determine . the effectiveness of emergency response, such as the shielding provided by the building stock in the area where people are 3.4 Level 3 Analysis assumed to shelter. the speed of evacuation, etc.

(Consequence Analysis Comparison of the results of different and Integrated Risk consequence codes, which embody different approaches and values of these parameters, on a Assessment) standard problem are contained in a study sponsored by the Organization for Economic Co-In this section, the analyses performed as part of operation and Development (OECD, 1994). An the Level 3 portion of a probabilistic risk uncertainty analysis of the COSYMA code results assessment (PRA) are described. using the expert elicitation method is currently being carried out (Jones, 1996).

3.4.1 Assumptions and Limitations 3.4.2 Products*

In most level 3 (i.e., consequence) codes, abnospheric transport of the released material is Documentation of the analyses performed to carried out assuming Gaussian plume dispersion. estimate the consequences associated with the This assumption is generally valid for flat terrain to accidental release of radioactivity to the a distance of a few kilometers from the point of environment should contain sufficient information release but is inaccurate both in the immediate to aflow an independent analyst to reproduce the vicinity of the reactor building and at farther results. At a minimum, the following information distances. For most PRA applications, however, should be documented for the Level 3 analysis:

the inaccuracies introduced by the assumption of Gaussian plumes are much smaller than the uncertainties due to other factors, such as the 3-114

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3. Technical Activities identification of the consequence code and the The consequence measures that focus on impacts version used to carry out the analysis, to the environment include:
  • a description of the site-specific data and assumptions used in the input to the code, land contamination
  • specifications of the source terms used to run the code, and
  • surface water body (e.g., lakes, rivers, etc.)
  • discussion and definition of the emergency contamination.

response parameters, a description of the computational process Groundwater contamination has yet to be included used to integrate the entire PRA model in a Level 3 analyses, although it may be important (Level 1 - Level 3), to consider it in certain specific cases.

  • a summary of all calculated results including frequency distributions for each risk measure. The economic impacts are mainly estimated in terms of the costs of countermeasures taken to 3.4.3 Analytical Tasks protect the population in the vicinity of the plant.

These costs can include:

A Level 3 PRA consists of two major tasks:

short-term costs incurred in the evacuation

1. Consequence analyses conditional on various and relocation of people during the emergency release mechanisms (source terms) and phase following the accident and in the destruction of contaminated food, and
2. Computation of risk by integrating the results of Levels 1, 2, and 3 analyses. long-term costs of interdicting contaminated farmland and residentiaVurban property which Task 1 - Consequence Analysis cannot be decontaminated in a cost-effective manner, Le., where the cost of The consequences of an accidental release of decontamination is greater than the value of radioactivity from a nuclear power plant to the the property.

surrounding environment can be expressed in several ways: impact on human health, impact on The costs of medical treatment to potential the environment, and impact on the economy. accident victims are not generally estimated in a The consequence measures of most interest to a Level 3 analysis, although approaches do exist for Level 3 PRA focus on the impact to human health. incorporating these costs (Mubayi, 1995) if They should include: required by the application.

number of early fatalities, The results of the calculations for each consequence measure are usually reported as a number of early injuries, complementary cumulative distribution function.

They can also be reported in terms of a number of latent cancer fatalities, distribution-for example, ones that show the 5th percentile, the 95th percentile, the median, and the mean.

population dose (person-rem or person-sievert) out to various distances from the plant, A probabilistic consequence assessment (PCA) code is needed to perform the Level 3 analysis.

individual early fatality risk defined in the early Such codes normally take as input the characteristics of the release or source term fatality QHO, i.e., the risk of early fatality for the average individual within 1 mile tram the provided by the Level 2 analysis. These plant, and characteristics typically include for each specified source term: the release fractions of the core

  • ... individual latent cancer fatality risk defined in inventory of key radionuclides, the timing and the latent cancer QHO, i.e'., the risk of latent duration of the release, the height of the release cancer fatality for the average individual within (i.e., whether the release is elevated or ground 10 miles of the plant level), and the energy of the release. PCA codes incorporate algorithms for performing weather 3-115
3. Technical Activities sampling on the plume transport in order to obtain the public and provide a more realistic estimate of a distribution of the concentrations and dosimetry , the doses and health effects following an which reflect the uncertainty and/or variability due accidental release. The MACCS code requires to weather. The codes also model variou~ that the analyst make assumptions on the values protective action countermeasures to permit a of parameters related to the implementation of more realistic calculation of doses and health protective actions following an accident The types effects and to assess the efficacy of these different of parameters involved in evaluating these actions actions in reducing consequences. include the following:

Several PCA codes are currently in use for

  • delay time between the declaration of a calculating the consequences of postulated general emergency and the initiation of an radiological releases. The NRC supports the use emergency response action, such as of the MACCS (Jow, 1990 and Chanin, 1993) and evacuation or sheltering; this delay time may MACCS2 (Chanin and Young, 1997) PCA codes be site specific.

for carrying out nuclear power plant Level 3 PRA analyses. A number of countries in Europe fraction of the offsite population which support the use of the COSYMA (KfK and NRPB, participates in the emergency response action, 1991 and Jones, 1996) PCA code for their Level 3 analyses. effective evacuation speed, PCA codes require a substantial amount of degree of radiation shielding provided by the information on the local meteorology, demography, building stock in the area, land use, crops grown in various seasons, foods consumed, and property values. For example, the

  • projected dose limits for long-term relocation input file for the MACes code requires the of the population from contaminated land, and following information:
  • projected 'Ingestion dose limits used to Meteorology - one year of hourly data on: interdict contaminated farmland.

windspeed and direction, atmospheric stability class, precipitation. rate, probability of The selected values assumed for the above (or precipitation occurring at specified distances similar) parameters need to be justified and from the plant site, and height of the documented since they have a significant impact atmospheric inversion layer. on the consequence calculations.

Demography - population distribution around In summary, the PCA code selected for the the plant on a polar grid defined by 16 angular calculation of consequences should have the sectors and user-specified annular radial following capabilities:

sectors, usually a finer grid close to the plant and one that becomes progressively coarser incorporate impact of weather variability on at greater distances. plume transport by performing stratified or Monte Carlo sampling on an annual set of

  • Land Use - fraction which is land, land which relevant site meteorological data, is agricultural, major, crops, and growing season. allow for plume depletion due to dry and wet deposition mechanisms,
  • Economic Data - value of farmland, value of nonfarm property, and annual farm sales. allow for buoyancy rise of energetic releases, The MACCS User Manual (Chanin, 1990) and the include all possible dose pathways, external MACCS2 User Guide (Chanin and Young, 1997) and internal (such as cloudshine, groundshine, may be consulted for a complete description of the inhalation, resuspension inhalation, and site input data necessary. ingestion) in the estimation of doses, In addition to site data, a PCA code should have employ validated health effects models based, provisions to model countermeasures to protect for example, on (ICRP, 1991) or BEIR V 3-116

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3. Technical Activities (National Research Council, 1990) dose warning and the likely participation in the factors for converting radiation doses to response by the offsite population.

early and latent health effects, and Task 2 - Computation of Risk

  • allow for the modeling of countermeasures to permit estimation bf a more realistic impact of The final step in a Level 3 PRA is the integration of accidental releases. results from all previous analyses to compute individual measures of risk. The severe acCident The above-cited methods for estimating progression and the radionuclide source term consequences are, in general, adequate for analyses conducted in the Level 2 portion of the accidents caused by internal initiating events PRA, as well as the consequence analysis during both full power operation and shutdown conducted in the Level 3 portion of the PRA, are conditions. However, forextemal initiating events, performed on a* conditional basis. That is, the such as seismic events, certain changes may be evaluations of alternative severe accident needed. For example, the early warning systems progressions, resulting source terms, and and the road network may be disrupted so that consequences are performed without regard to the initiation and execution of emergency response absolute or relative frequency of the postUlated actions may not be possible. Hence, in addition to accidents. The final computation of risk is the changing the potential source terms, a seismic process by which each of these portions of the event could also influence the ability of the close-in accident analysis are linked together in a self-population to cany out an early evacuation. A consistent and statistically rigorous manner.

Level 3 seismic PRA should, therefore, include consideration of the impacts of different levels of An important attribute by which the rigor of the earthquake severity on the consequence process is likely to be judged is the ability to assessment demonstrate traceability from a specific accident sequence through the relative likelihood of To use a consequence code, generally the alternative severe accident progressions and following data elements are required: measures of associated containment performance (Le., early versus late failure) and ultimately to the reactor radionuclide inventory, distribution of fission product source terms and consequences. This traceability should be accident source terms defined by the release demonstrable in both directions, i.e., from the fractions of important radionuclide groups, the accident sequence to a distribution of timing and duration of the release, and the consequences and from a specific level of energy and height of the release, accident consequences back to the fission product source terms, containment performance hourly meteorological data at the site as measures, or accident sequences that contribute recommended, for example, in Regulatory to that consequence level.

Guide 1.23 (NRC, 1986), collected over one or, preferably,' more years and processed into 3.4.4 Task Interfaces a form usable by the chosen code, The current task requires a set of release fractions site population data from census or other (or source terms) from the Level 2 analysis reliable sources and processed in conformity (Section 3.3) as input to the consequence with the requirements of the code, i.e., to analysis.

provide population information for each area element on the grid used in the code, The consequences are calculated in terms of:

(1) the acute and chronic radiation doses from all site economic and land use data, specifying pathways to the affected population around the

'. 'the important crops in the area, value and plant, (2) the consequent health effects (such as extent of farm and nonfarm property, early fatalities, early injuries, and latent cancer fatalities), (3) the integrated population dose to defining the emergency response countermeasures, including the possible time delay in initiating response after declaration of 3-117

3. Technical Activities some specified distance (such as 50 miles) from Jones, J. A, et aI., "Uncertainty Ailalysis on the point of release, and (4) the contamination of COSYMA," Proceedings of the Combined 3rd land from the deposited material. COSYMA Users Group and 2nd Intemational MACCS Users Group Meeting, Portoroz, Slovenia, .

The consequence measures to be calculated 41228-NUC 96-9238, KEMA, Amhem, the depends on the application as defined in PRA Netherlands, September 16-19,1996.

"Scope. Generally, in a level 3 analysis,' a distribution of consequences is obtained by Jow, H. N., et aI., "MElCOR Accident statistical sampling of the weather conditions at Consequence Code System (MACCS), Volume II, the site. Each set of consequences, however, is Model Description," NUREG/CR-4691, Sandia conditional on the characteristics of the release (or National "laboratories, February 1990.

source term) which are evaluated in the level 2 analysis. KfK and NRPB, "COSYMA - A New Program Package for Accident Consequence Assessment, "

An integrated risk assessment combines the CEC Brussels, EUR 13028, results of the levels 1, 2, and 3 analyses to Kemforschungszentrum (Karlsruhe) and National compute the selected measures of risk in a self- Radiological Protection Board, 1991.

consistent and statistically rigorous manner. The risk measures usually selected are: early fatalities, Mubayi, V., et al., "Cost-Benefit Considerations in latent cancer fatalities, population dose, and Regulatory Analysis," NUREG/CR-6395, quantitative health objectives (QHOs) of the U.S. Brookhaven National laboratory, 1995.

Nuclear Regulatory Commission (NRC) Safety Goals (NRC, 1986). Again, the actual risk National Research Council, "Health Effects of measures calculated will depend on the PRA Exposure to low levels of Ionizing Radiation,"

Scope. BEIR V, Washington, DC, 1990.

3.4.5 References NRC, "Severe Accident Risks' An Assessment for Five U.S. Nuclear Power Plal'ts," NUREG-1150, Chanin, 0.1., and M. l. Young, "Code Manual for Vol. 1, Main Report, U.S. Nuclear Regulatory MACCS2: Volume 1, User's Guide," SAND97- Commission, 1990. " .

0594, Sandia National laboratories, March 1997.

NRC, "Safety Goals for the Operation of Nuclear Chanin, 0.1., etal., "MACCSVersion 1.5.11.1: A Power Plants, Policy Statement," Federal Register, Maintenance Release of the Code," NUREG/CR- Vol. 51, No. 149, U.S. Nuclear Regulatory 6059, Sandia National laboratories, October 1993. Commission, August 4, 1986.

Chanin, 0.1., et aI., "MELCOR Accident NRC, *Onsite Meteorological Programs,"

Consequence Code System (MACCS), Volume 1, Regulatory Guide 1.23, U.S. Nuclear Regulatory User's Guide," NUREGlCR-4691, Sandia National Commission, April 1986.

laboratories, February 1990.

NRC, *PRA Procedures Guide - A Guide to the Performance of Probabilistic Risk Assessments for Harper, F. T., et aI., "Probabilistic Accident Nuclear Power Plants: NUREG/CR-2300, Vol. 2, Consequence Uncertainty Analysis, Dispersion, and Deposition Uncertainty Assessment," U.S. Nuclear Regulatory Commission, 1983.

NUREG/CR--6244, Sandia National Laboratories, 1995. OECD, "Probabilistic Accident Consequence Assessment Codes, Second International ICRP, "1990 Recommendations of the "ICRP: Comparison", Organisation for Economic Annals of the ICRP, Vol." 21, No. 1-3, ICRP Cooperation and Development, Nuclear Energy Publication 60, Intemational Commission on Agency, Paris, France, 1994~

Radiological Protection, Pergamon Press, Oxford, England, 1991.

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