ML092640172

NUREG/CR-6572, Rev. 1, BNL-NUREG-52534-R1, Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA Procedure Guides for a Probabilistic Risk Assessment

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50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, BNL-NUREG-52534-R1, Job Code R2001, RAS 284 BNL-NUREG-52534-R1, NUREG/CR-6572, Rev 1
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NUREG/CR-6572,. Rev. 1 BNL-NUREG-52534-R1 Kalinin VVER-1000 Nuclear 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 Governments of the 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

ABSTRACT

'In order to facilitate the probabilistic risk assessment (PRA) of a VVER-1000 nuclear power plant, a'set of procedure guides has been written. These procedure guides, along with training supplied by experts and supplementary material from the literature, were used to advance the PRA carried out for theKalinin Nuclear Power Station in the Russian Federation. Although written for a specific project, these guides have general applicability. Guides are procedures for all the technical tasks of a Level 1 (determination of core damage frequency for different accident scenarios), Level 2 (probabilistic accident progression and. source term analysis), and Level 3 (consequence analysis and integrated risk assessment) PRA. In addition, introductory material is provided to explain the rationale and approach for a PRA. Procedure guides are also provided on the documentation requirements.

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FOREWORD During the Lisbon Conference on Assistance to the Nuclear Safety Initiative, held in May 1992, partidpants agreed that efforts should be undertaken to improve the safety of nuclear power plants that were designed and built by the former Soviet Union. That agreement led to a collaborative probabilistic risk assessment (PRA) of the Kalinin Nuclear Power Station (KNPS), Unit 1, in the Russian Federation. The KNPS Unit 1 PRA was intended to demonstrate the benefits obtained from application of risk technology towards understanding and improving reactor safety and, thereby, helping to build a risk-informed framework to help address, reactor safety issues in regulations.

The U..S. Department of State, together with the Agency for International Development (AID),

requested that the U.S. Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation (Gosatomnadzor, or GAN) work together to begin applying PRA technology to Soviet-designed plants.' On the basis of that request, in 1995, the NRC and GAN agreed to work together to perform a PRA of a WER-1 000 PWR reactor. Under that

.agreement, the NRC provided financial support for the PRA with funds from AID and technical support primarily from Brookhaven National Laboratory and its subcontractors. KNPS Unit 1 was chosen for the PRA, and the effort was performed under the direction of GAN with the assistance of KNPS personnel and the following four other Russian organizations:

Science and Engineering Centre for Nuclear and Radiation Safety (GAN's and now Rostechnadzor's technical support organization)

• Gidropress Experimental and Design Office (the VVER designer)

• Nizhny Novgorod Project Institute, "Atomenergoprojekt" (the architect-engineer)

• Rosenergoatom Consortium (the utility owner of KNPS)

One of the overriding accomplishments of the project has been technology transfer. In NRC-sponsored workshops held in Washington, DC, and Moscow from October 1995 through November 2003, training was provided in all facets of PRA practice. In addition, the Russian participants developed expertise using current-generation NRC-developed computer codes, MELCOR, SAPHIRE and MACCS. Towards the completion'of the PRA, senior members of the Kalinin project team began the development of risk-informed, Russian nuclear regulatory guidelines. These guidelines foster the application of risk assessment concepts to promote a better understanding of risk contributors. Efforts such as this have benefited from the expertise obtained, in part, from the training, experience, and insights gained from participation in the KNPS Unit 1 PRA project.

The documentation of the Kalinin PRA comprises two companion NUREG-series reports:

, NUREG/CR-6572, Revision 1, "Kalinin VVER-1000 Nuclear Power Station Unit 1 PRA:

Procedure Guides for a Probabilistic Risk Assessment," was prepared by Brookhaven National Laboratory and the NRC staff. It contains guidance for conducting the Level 1, 2, and 3 PRAs for KNPS with primary focus on internal events. It may also serve as a guide for future PRAs in support of other nuclear power plants..

1 As a result of a governmental decree in May 2004, GAN was subsumed into a new organization, known as the Federal Environmental, Industrial and. Nuclear Supervision Service of Russia (Rostechnadzor).

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NUREG/IA-0212, "Kalinin WER-1 000 Nuclear Power Station Unit 1 PRA: Volumes 1 and 2,"was written by the Russian* team and,. by- agreement, includes both 'a non-proprietary and proprietary volume. The non-proprietary volume, Volume 1, "Executive Summary Report," discusses the project objectives, summarizes how the project was carried out, and presents a general summaryof the PRA results, The proprietary volume, Volume 2, contains three, parts. Part 1, "Main Report: Level 1 PRA, Internal Initiators," discusses the Level 1 portion of the PRA; Part 2, "Main Report: Level 2 PRA, Internal Initiators," discusses the Level 2 portion; and Part 3, "Main Report: Other Events Analysis," discusses preliminary analysesof fire, internal flooding, and seismic events, which may form the basis for additional risk assessment work at some future time.

Carl J. Paperiello, Director Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission VI

'ACKNOWLEDGMENTS The following organizations and individuals collaborated in performing the PRA for the Kalinin NPS, Unit 1:

U.S. NuclearRegulatory Commission (NRC)

Charles Ader John Lane Mark Cunningham Scott Newberry Mary Drouin Themis Speis Thomas King Andrew Szukiewicz NRC Contractors Mohammed Ali Azarm, Brookhaven National Mark Leonard, Dycoda Laboratory (BNL) Hossein Nourbakhsh, BNL Dennis Bley, Buttonwood Consulting Inc. Robert Kennedy, RPK Structural Mechanics Tsong-Lun Chu, BNL Consulting David Diamond, BNL Robert Campbell, EQE International Inc.

Ted Ginsberg, BNL Yang Park, BNL David Johnson, PLG Inc. Trevor Pratt, BNL John Lehner, BNL Jimin Xu, BNL Federal Nuclear and Radiation Safety Authority of the Russian Federation (GAN), now the Federal Environmental,Industrial and NuclearSupervision.Service of Russia (Rostechnadzor)

Mikhail Mirochnitchenko Alexandr Matveev Alexandr Gutsalov Science and Engineering Center for Nuclear and Radiation Safety Irina Andreeva Dmitri Noskov Tatiana Berg Gennadi Samokhin Valentina Bredova Eugene Shubeiko Boris Gordon, Vyacheslav Soldatov

.Irina loudina Sergei Volkovitskiy Artour Lioubarski Elena Zhukova Kalinin Nuclear Power Station Grigori Aleshin Eugene Mironenko Oleg Bogatov Maxim Robotaev I Experimental and'Design Office "Gidropress" Viatcheslav Kudriavtsev Vladimir Shein Valeri Siriapin Nizhny Novgorod Project Institute "Atom energoprojekt" Ludmila Eltsova Valeri Senoedov Vladimir Kats Alexander Yashkin Svetlana Petrunina Rosenergoatom Consortium.,

Vladimir Khlebtsevich xii,

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, "MAAP4 - 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, et al., "MELCOR Computer Code depending on the particle size distribution -of the Manuals -- Version 1.8.3," NUREG/CR-6119, released material, the surface roughness ofthe SAND93-2185, Volumes 1-2, Sandia National terrain, and other factors.. An assessment of these 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 al., 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 al, "Evaluation of SevereAccident 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. Nuclear Regulatory deposited on the ground, modeling of the ingestion Commission, August 1989: pathway, i.e., the food chains, ground-crop-man and ground-crop-animal-dairy/meat-man, internal Jow,- H. J., et al., "XSOR Codes User Manual," and external dosimetry, and the health effects NUREG/CR-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 3.4 Level 3 Analysis 'building stock in the area where people are assumed to shelter, the speed of evacuation, etc.

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

3. Technical Activities

• identification of the consequence code and the to the environment include:

version used to carry out the analysis, 3

• a description of the site-specific -data and

• land contamination assumptions used in the input to the code,

• specifications of the source terms used to run

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

the code, and contamination.

• dis6ussion and .definition of the emergency response parameters, .Groundwater contamination has yet to be included

• a description of the computational process in a Level 3 analyses, although it may be important used to integrate the entire PRA model to consider it in certain specific cases.

(Level 1 - Level 3),

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

3.4.3 Analytical Tasks These costs can include:

A Level 3 PRA consists of two major tasks: short-term costs incurred in the evacuation and relocation of people during the emergency

1. Consequence analyses conditional on various phase following the accident and in the release mechanisms (source terms) and destruction of contaminated food, and

2. Computation of risk by integrating the results - long-term costs of interdicting contaminated of Levels 1, 2, and 3 analyses. farmland and residential/urban property which J cannot be decontaminated in a cost-effective Task 1 - Consequence Analysis manner,. i.e., where the, cost of decontamination is greater than the value of The consequences of an accidental release of the property.

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

should include:

The. results of the calculations for each

• number of early fatalities, consequence measure are usually reported as a complementary cumulative distribution function.

• number of early injuries, They can -also be reported in terms of a distribution--for example, ones that show the 5th

• number of latent cancer fatalities, 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 neede-d to.perform the Level 3 analysis.

individual early fatality riskdefined in the early Such codes normally take as input the fatality QHO, i.e., the riskof early fatality forthe characteristics of the. release or source term ayerage individual within 1 mile from the plant, provided by the Level 2 analysis. These 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 The consequence measures thatfocus on impacts sampling on the, plume transport in order to obtain 3,115.

3. Technical Activities a distribution of the concentrations and dosimetry analyst make assumptions on the values of which reflect the uncertainty and/or variability due parameters related to the implementation of to weather. The codes also model various protective actions following an accident. The types protective action countermeasures to permit a of parameters involved in evaluating these actions more realistic calculation of doses and health include the following:

effects and to assess the efficacy of these different actions in reducing consequences. delay time between the declaration of a general emergency and the initiation of an Several PCA codes are currently in use for emergency response action, such as calculating the consequences of postulated evacuation or sheltering; this delay time may radiological releases. The, NRC supports the use be site specific, of the MACCS (Jow, 1990 and Chanin, 1993) and MACCS2 (Chanin and Young, 1997) PCA codes fraction of the offsite population which for carrying out nuclear power plant Level 3 PRA participate.s in the emergency response action, analyses. A number of countries in Europe support the use of the COSYMA (KfK and NRPB, 1991 and effective evacuation speed, Jones, 1996) PCA code for their Level 3 analyses.

• degree of radiation shielding provided by the PCA codes require a substantial. amount of building stock in the area, information on the local meteorology, demography, land use, crops grown in various seasons, foods

• projected dose limits for long-term relocation.of consumed, and property values. For example, the the population from contaminated land, and input file for the MACCS code requires the following information:* projected ingestion dose limits used to interdict contaminated farmland.

Meteorology - one year of hourly data on:

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

,atmospheric inversion layer.

In summary, the PCA code selected for the Demography - population distribution around calculation of consequences should have the the plant on a polar grid defined by 16 angular following capabi!ities:

sectors and user-specified annular -radial sectors, usually a finer grid close to the plant

• incorporate impact of weather variability on and one that becomes progressively coarser at plume. transport by performing stratified or greater distances. Monte Carlo sampling .on an annual set of.

relevant site meteorological data,

• Land Use - fraction which is land, land which 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 annu'al 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 -employvalidated health effects models based, provisions to model countermeasures to protect the for example, on (ICRP, 1991) or BEIR V public and provide a more realistic~estimate of the (National Research Council, 1990) dose doses and health effects following an accidental factors for converting radiation doses to early release. .The MACCS code requires that the and latent health effects, and 3-116

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

impacts of different levels of earthquake severity on the consequence assessment.. An important attribute by which the rigor of the process is likely to be judged is the ability to To use a consequence code, generally the demonstrate traceability from a specific accident following data elements are required: sequence through the relative likelihood of alternative severe accident progressions and

• reactor radionuclide inventory, measures of associated containment performance (i.e., early versus late failure) and ultimately to the accident source terms defined by the release distribution of fission product source terms and fractions of important radionuclide groups, the consequences. This traceability should be timing and duration of the release, and the demonstrable .in both directions, i.e., from the energy and height of the release, accident sequence to a -distribution of consequences andfrom a specific level of accident hourly meteorological data at the site as consequences back to the fission product source recommended, for example, in Regulatory terms, containment performarnce measures, or Guide 1.23 (NRC, -1986), collected over one or; accident sequences that contribute to that preferably, more years and processed into a consequence level.

form usable by the chosen code, 3.4.4 Task Interfaces

" site population data from census or other reliable sources and processed in conformity The currenttask requires a set of release fractions with the requirements of the code, i.e., to (or source terms) from the Level 2 analysis provide population information for each area (Section 3.3) as input to the consequence analysis.

element on the grid used in the code,

" site economic and land use data, .specifying The consequences are calculated in terms of:

the important crops in the area, value and (1) the acute and chronic radiation doses from all extent.of farm and nonfarm property, pathways to the affected population around the plant, (2) the consequent health effects (such as

" defining the emergency response early fatalities, early injuries, and latent cancer countermeasures, including the possible time fatalities), (3) the integrated population dose to delay in initiating response after declaration of some specified distance (such as 50 miles) from warning and the likely participation in the the point of release, and (4) the contamination of response by the offsite population. land from the deposited material.

The consequence measures to be calculated depends on the application as defined in PRA 3-117

3: Technical Activities Scope. Generally, in a Level 3 analysis, a Jones, J. A., et al., "Uncertainty Analysis on distribution of consequences is obtained by COSYMA," Proceedings of the Combined 3rd statistical sampling of the weather conditions at the COSYMA Users Group and 2nd International site. Each set of consequences, however, is MACCS Users Group Meeting, Portoroz, Slovenia, conditional on the characteristics of the release (or 41228-NUC 96-9238, KEMA, Arnhem, the source term) which are evaluated, in the Level 2 Netherlands, September 16-19, 1996.

analysis.

Jow, H.. N., et al., "MELCOR Accident An *integrated risk assessment combines' the Consequence Code System (MACCS), Volume II, results of the Levels 1, 2, and 3 analyses to Model Description," NUREG/CR-4691, Sandia compute the selected measures of risk in a self- National Laboratories, February 1990.

consistent and statistically rigorous manner. The risk measures usually selected are: early fatalities, .KfK and NRPB, "COSYMA - A New Program latent cancer fatalities, population dose, and Package forAccident Consequence Assessment,"

quantitative health objectives (QHOs) of the U.S. CEC Brussels, EUR 13028, Nuclear Regulatory Commission (NRC) Safety Kernforschungszentrum (Karlsruhe) and National Goals (NRC, 1986). Again, the actual risk Radiological Protection Board, 1991.

measures calculated will depend on the PRA Scope. Mubayi, V., et al., "Cost-Benefit Considerations in Regulatory Analysis," NUR.EG/CR.-6395, 3.4.5 References Brookhaven National Laboratory, 1995.

Chanin, D.I., and M. L. Young, "Code Manual for National Research. Council, "Health Effects of MACCS2: Volume 1, User's Guide," SAND97- Exposure to Low Levels of Ionizing Radiation,"

0594, Sandia National Laboratories, March 1997. BEIR V, Washington, DC, 1990.

Chanin, D.I., et al., "MACCS Version 1.5.11.1: A NRC, "Severe Accident Risks: An Assessment for Maintenance Release of the Code," NUREG/CR- Five U.S. Nuclear Power Plants," NUREG-1150, 6059, Sandia National Laboratories, October1993. Vol. 1, Main Report, U.S. Nuclear Regulatory Commission, 1990.

Chanin, D.I., et al., "MELCOR Accident Consequence Code System (MACCS), Volume 1, NRC, "Safety Goals for the Operation of Nuclear User's Guide," NUREG/CR-4691, Sandia National Power Plants, Policy Statement," Federal Register, Laboratories, February 1990. Vol. 51, No. 149, U.S. Nuclear Regulatory Commission, August 4, 1986.

Harper, F. T., et al., "Probabilistic Accident Consequence Uncertainty Analysis, Dispersion, NRC, "Onsite Meteorological Programs,"

and Deposition Uncertainty Assessment," Regulatory Guide 1.23, U.S. Nuclear Regulatory NUREG/CR-6244, Sandia National Laboratories, Commission, April 1986.

1995.

NRC, "PRA Procedures Guide - A Guide to the ICRP, "1990 Recommendations of the ICRP," Performance of Probabilistic Risk Assessments for Annals of the ICRP, Vol. 21, No. 1-3, ICRP Nuclear Power Plants," NUREG/CR-2300, Vol, 2Z Publication 60, International Commission on U.S. Nuclear Regulatory Commission, 1983.

Radiological Protection, Pergamon Press, Oxford, England, 1991. OECD, "Probabilistic Accident Consequence Assessment Codes, Second International Comparison", Organisation for Economic Cooperation and Development, Nuclear Energy Agency, Paris, France, 1994.

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