Probabilistic risk assessment scope

The sequences of events that may lead to vessel failure and their frequencies are determined from probabilistic risk assessment (PRA) analyses. The pressure, temperature and heat transfer coefficient time histories at the vessel inner surface are determined from thermal hydraulic analyses for the events identified by the PRA analyses. These time histories are used together with probabilistic fracture mechanics (PFM) analysis to calculate the conditional probability of RPV failure. Discussion of the methodology used to perform the PRA analyses and define the transient events and associated frequencies, and the thermal hydraulic analyses used to define the event pressure and temperature time histories are outside the scope of this chapter. Consequently, the remainder of this chapter focuses on the PFM evaluation assumptions and procedures. 

Assumptions are an integral and inevitable element of any safety case. Assumptions impact on the scope and nature of the arguments and evidence presented. For example, assumptions made concerning the lifetime and maintenance of a system can affect the details of probabilistic risk assessment. Similarly, assumptions made regarding the independence of system functions will determine whether function interactions are explicitly addressed in the safety case. 

The US Nuclear Regulatory Commission (NRC) Safety Evaluation Report (U.S. NRC, 2014) gives the criteria for risk assessment based on core damage frequency and the time scales for the use of safety and nonsafety systems, as derived from a full-scope Probabilistic Risk Assessment (PRA) analysis (Bhatt and Wachowiak, 2006). This NRC approach states the safety guidelines as follows ... 

The principal tool supporting this activity is the full-scope Probabilistic Risk Assessment (PRA) that WSRC is developing for the SRS reactors. The PRA, when completed, will include Level 1, 2, and 3 analyses. The Level 1 analysis identifies core damage sequences and their frequencies. The Level 2 analysis studies each accident sequence from core melt progression, through confinement system response, to the release of radionuclides to the environment. The Level 3 analysis takes the release from each accident sequence and estimates the consequences. Accident initiators Include Internal events, which are associated with component failures in the reactor systems, and external events, which are outside the reactor systems, such as earthquakes, fire, or flooding. 

The Reactor Safety Study was prompted in part by a request from Senator John Pastore for a comprehensive assessment of reactor safety. The AEC s first response to this request was the WASH-1250 report entitled The Reactor Safety Study of Nuclear Power Reactors (Light Water-Cooled) and Related Facilities, which was published in final form in July 1973. However, WASH-1250 did not provide a probabilistic assessment of risk as requested in Senator Pastore s letter. At the time, relevant probabilistic estimates were quite limited in scope and/or highly subjective. For example, in a policy paper dated November 15, 1971, to the commissioners proposing an approach to the preparation of environmental reports, the regulatory staff estimated that the probability of accidents leading to substantial core meltdown was 10 per reactor-year. In retrospect, this was a highly optimistic estimate, but it typifies the degree to which meltdown accidents were considered "not credible."... 

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