Preliminary Probabilistic Safety Analysis

J. BARÓN; J. NÚÑEZ MC LEOD; S. RIVERA

2000-08-01

2001-06-01

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

Energia-Reactores

This report documents the Preliminary Probabilistic Safety Assessment (PSA) study of the Replacement Research Reactor Facility (Reactor Facility) to be constructed at Lucas Heights, NSW. 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 safety objectives against which the results will be 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 system 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 they are 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. Uncertainty and importance estimations are also included. Chapter 7 presents the results of the Level III PSA considerations. They include the estimation of the frequency and associated consequence (individual doses), for a few accidents, which constitute the risk representative scenarios of the plant. Chapter 8 presents the data used in the PSA. The methodology adopted for this PSA generally follows the NUREG/CR-2300 recommendations, except in the treatment of common cause failures which are discussed in section 1.6 [[i]]. This is considered more appropriate for a new reactor under construction. [i]     NUREG/CR-2300, PRA PROCEDURES GUIDE, NRC, USA, 1983, pp. 3-21.