Probabilistic Safety Assessment (PSA) for European Pressure Reactor (EPR)

Tentative Project Title: Development of Probabilistic Safety Assessment to the Conceptual Design Stage for Gen IV Sodium-Cooled Fast Reactors

Summary
A probabilistic safety assessment (PSA) methodology will be developed for risk evaluation of sodium-cooled fast reactors (SFR). As part of this study, the characteristic features of the SFR design

will be described and the design requirements will be examined. The scope of the PSA performed at the development stage of the technical design is defined. Initiating events are determined and

then event trees will be developed to quantify the probability of event sequences in the PSA. According to PSA results and sensitivity analysis of several SFR design alternatives, the best design

will be optimised. The important problems which must be solved in a PSA for the designs of new NPPs will be discussed.

Introduction and Justification for the Research
The first reactors of the Generation IV nuclear energy systems are foreseen to operate in the period 2020 – 2030. In order to get a favourable public perception -especially after the disaster of

Fukushima (2011)- they will have to show satisfactory level of safety. Because of that, the Generation IV systems will likely launch considerable innovative technological changes compared to

current plants and these changes will have to be at the higher possible level of safety,.

One of the most popular safety tools is the Probabilistic Safety Assessment (PSA), which is an analytical method used to examine the risk from a variety of possible initiating events (e.g.

transients, LOCAs, support system failures, etc.) and to determine what the most likely contributors are to that failure. Use of PSA techniques in the conceptual design stage allows comparison

of design alternatives and consideration of the effects of these alternatives for safety improvements and system optimisation.

In the past, the design of NPPs was mainly based on deterministic methods. PSA is recently used to support deterministic criteria and analyses in the design process for new reactor concepts in

several projects. Since that the Generation IV reactors are new and alive subject of research, it is proposed to conduct a research on the development of PSA to the conceptual design stage for

Gen IV sodium-cooled fast reactors.

Hypothesis, Aims and Objectives
Hypothesis
Generation IV viability evaluations will be conducted with incomplete design information. For these evaluations, the deterministic concept of defence in depth needs to be integrated with

simplified probabilistic considerations to provide metrics for acceptability and a basis for additional requirements, and to ensure a well-balanced design.
Aims
PSA is conducted during the design process to support and optimise the design of systems and processes. This allows a well-balanced system and process design to be attained. It also provides

a reasonable assurance that plant will comply with the general safety objectives.

Objectives
The objectives of this study are to:
1) Identify initiating events potentially significant relative to reactor safety.
2) Estimate the occurrence frequency of these events.
3) Assess the event consequences.
4) Quantify statistical uncertainties in the frequency and consequence estimates.
5) Select the best design of the reactor based on the PSA.

Methodology and Approach

The method begins by selecting initiating events and then is continued by constructing event trees for accident sequences, analysing the sequences of events to obtain the probabilities and to

evaluate the consequences of the escape of radioactivity, and finally utilising the results to support the plant design.

Initiating events are selected that have the potential to lead to the release of radioactivity from the plant. Once the initiating events are defined, a systematic presentation of the progression

of the accident sequences from initiation to termination is provided in an event tree for each initiating event.

To anticipate and understand these sequences, systems analysis is needed to show the transient response, such as core temperatures, and to show the response of active systems such as the

ability of cooling systems to remove the decay heat under the conditions specified in the accident sequence. Intersystem dependencies may also be important.

The probability of occurrence of each event along each of the accident sequences within the event tree will be obtained from fault tree analysis. A fault tree is a logic diagram which gives the

probability of an undesired state of a . system (e.g., loss of cooling) when the various component failure modes, probabilities, and dependencies are known.

The component failure probabilities come from data banks containing standardised reliability values and raw experience data. In the evaluation of fault trees, it is important to consider common

mode failures which can lead to simultaneous failure of redundant components or systems. Uncertainty analysis allows the generation of mean values for probabilities of accident sequences.

The analysis of consequences and physical phenomena for the accident sequences is simplified by grouping the sequences into a smaller number of release categories such that the system

responses of sequences within a given category are very similar and therefore result in about the same consequences, given a release category, the transient analysis is done first to determine

the condition of key components – such as the temperatures in the reactor core. Based on these results, the time-dependent radionuclide transport is calculated with the end result being

radiation dose or health effects to the public and the uncertainty distributions for these accident consequences.

Computer Codes:

Assigning the correct probabilities to the components, combining these numbers according to system logic and solving the system model, can be done practically with some verified computer

codes, such as: SAPHIRE, PSAPACK, RISKMAN, FaultTree++, RiskSpectrum, … etc. In this project, some of the PSA computer packages will be evaluated to select the most appropriate code(s) which

will be utilised.

Literature Review

As a first stage of the research, the following sources are some of literature that will be reviewed:

A. Papazoglou et al, “Probabilistic Safety Analysis Procedures Guide”, NUREG/CR-2815, BNL-NUREG-51559, U.S. Nuclear Regulatory Commission, 1984.

Amico, P. G., “Probabilistic analysis in the design of advanced NPPs”. Paper presented at ANP 92 Conference, Tokyo 1992.

Andre, G. R., Iacovino, J. M. & Schulz, T. L., “Application of probabilistic risk assessment in the design of Westinghouse advanced reactors”. Proceedings PSA’91, IAEA, Wien 1991, pp 694.

Breding, R. J. et al, “Summary Description of the Methods Used in the Probabilistic Risk Assessments for NUREG 1150”, Nuclear Engineering and Design, 135 (1992) 1-27.

F. A. Felder, “Probabilistic Risk Analysis of Restructured Electric Power Systems: Implications for Reliability Analysis and Policies, PhD. Dissertation Department of Technology, Management and

Policy, M.I.T. Cambridge, MA 2001.

GA TECHNOLOGIES INC., “RQBABILISTIC RISK ASSESSMENT QF THE MODULAR HTGR PLANT”, HTGR-86-011 Revision 1, June 1986.

General Atomic Technologies Inc, “Probabilistic Risk Assessment of the Modular HTGR Plant”, HTGR-86-011, Revision 1. General Atomic Technologies Inc, San Diego, USA, 1986.

H. Kumamoto and E. J. Henley, “Probabilistic Risk Assessment and Management for Engineers and Scientists”, 2nd Edition, New York, IEEE Press, 1996.

H. Yamano, I. Sato, Y. Tobita, “Development of technical basis in the initiating and transition phases of unprotected events for Level-2 PSA methodology in sodium-cooled fast reactors”,
Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311-1393, Japan.

INTERNATIONAL ATOMIC ENERGY AGENCY, “Applications of Probabilistic Safety Assessment (PSA) for Nuclear Power Plants”, IAEA-TECDOC-1200, 2001.

INTERNATIONAL ATOMIC ENERGY AGENCY, “Case Studies in the Application of Probabilistic Safety Assessment Techniques to Radiation Sources”, IAEA-TECDOC-1494, April 2006.

K. KURISAKA, “Probabilistic Safety Assessment of Japanese Sodium-cooled Fast Reactor in Conceptual Design Stage”, 15th Pacific Basin Nuclear Conf. Sydney, Australia, 15-20 Oct. 2006.

M. Fujii, S. Morooka, and H. Heki, “Application of Probabilistic Safety Analysis in Design and Maintenance of the ABWR”, Advances in Light Water Reactor Technologies, 2011.

N. McCORMICK, “Reliability and Risk Analysis: Methods and Nuclear Power Applications”, Academic Press, 1981.

OECD NUCLEAR ENERGY AGENCY, “Basis for the Safety Approach for Design & Assessment of Generation IV Nuclear Systems”, GIF/RSWG/2007/002, November 24, 2008.

P. Kafka, “Important Issues Using PSA Technology for Design of New Systems and Plants”, Reliability Engineering and System Safety 45 (1994) 205-213.

R. FULLWOOD, “Probabilistic Safety Assessment in the Chemical and Nuclear Industries”, Butterworth-Heinemann, 2000.

Sato T, Akinaga M, Kojima Y, “Safety Design Philosophy of the ABWR for the Next Generation LWRs, Proceedings of ICAPP’09, Paper 9447, Tokyo, Japan, 2009.

US NRC, “Guidelines on Modeling Common-Cause Failures in Probabilistic Risk Assessment”, Washington, NUREG/CR-5485, Nov. 1998.

Yu. V. Shvyryaev, et al, “Use of Probabilistic Analysis in Safety Validation of AES-2006 Designed for the Novovoronezh Nuclear Power Plant Site”, Atomic Energy, Vol. 106, No. 3, 2009.
Project Plan

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