Final Probabilistic Safety Analysis - Replacement Research Reactor

J. BARÓN; J. NÚÑEZ MC LEOD; S. RIVERA

2004-02-01

2004-08-01

Documento Tecnológico para la obtención de la Licencia de Operación en Australia

Energia-Reactores

This report documents the Probabilistic Safety Assessment (PSA) study of the Replacement Research Reactor Facility (Reactor Facility) built at Lucas Heights, NSW. The PSA represents the status of the Reactor Facility consistent with the Safety Analysis Report (SAR) and incorporates all the advances in detailed engineering produced from the time the Preliminary PSA was written to apply for the construction licence. This document is ready to apply for the operation licence. All feedbacks to design were done. This chapter provides an introduction to the study, provides some background to the project and explains the scope of the study, the methods used, and the regulatory safety objectives against which the results are then compared. Chapter 2 discusses the methodology to treat the common cause failures and presents the development of these failures in fault trees, including their quantification with human reliability analysis data. Chapter 3 discusses the initiating events, their derivation, screening, and grouping, and the evaluation of their estimated frequencies. Chapter 4 discusses the event tree headings. These are the safety systems’ responses to an accident sequence, modelled as fault trees to determine the probability of failure of a safety system were an event to occur which required that response. The hypotheses on the development of the headings are discussed and then quantified. Chapter 5 discusses the event trees. These are the symbolic representation of the different possibilities following an initiating event. An interference matrix is presented showing the systems that are required to respond for each initiating event. The event trees are then quantified. Chapter 6 presents the Level I PSA results, that is, the results showing the estimated frequency of each possible sequence and the summed frequency for each kind of end state that contributes to the Core Damage Frequency (CDF). Uncertainty and importance estimations are also included. Chapter 7 presents the analysis of the internal fire event, using IAEA methodology for NPPs.  These analyses are quantified under conservative assumptions and its relative contribution to the CDF is estimated. It also includes analyses relative to the extreme wind events. Chapter 8 presents the results of the Level III PSA considerations. They include the estimation of the frequency and associated consequence (individual doses), for a set of accidents that constitute the risk representative scenarios of the plant. Chapter 9 presents the data used in the PSA, and describes their source. The methodology adopted for this PSA basically follows the NUREG/CR-2300 [[i]] recommendations, except in the treatment of common cause failures which are discussed in section 1.6. The approach for this treatment is considered more appropriate for a new reactor which does not have operational experience. [i]     NUREG/CR-2300, PRA PROCEDURES GUIDE, NRC, USA, 1983, pp. 3-21.