Probabilistic Safety Assessment


The original licensing of reactors in the United States was based on the premise that the Large Break Loss of Coolant Accident LOCA) was likely to be the most severe accident a reactor could experience. The Large Break LOCA assumes that the ~ 30 inch diameter pipe used in the reactor cooling system undergoes a guillotine break and the high pressure steam-water mixture discharges from the pipe.

In 1975, a study entitled WASH 1400 - Reactor Safety Study (also called the Rasmussen Report after Professor Norman Rasmussen of MIT) evaluated the probability of a number of accident sequences that might lead to melting of the fuel in the reactor (also referred to as Core Melt). This study was expected to provide a more realistic assessment of the risks associated with commercial plants. The Surry (PWR) and Peach Bottom (BWR) reactors were the base plants investigated. The study considered meteorological conditions and health effects. WASH 1400 found that transients, small break LOCAs, and human error could be important contributors to risk. Four years later, the Three Mile Island accident confirmed that conclusion.

This risk evaluation methodology was improved upon. In most countries the method is referred to as Probabilistic Safety Assessment (PSA). In the United States, the method is referred to as Probabilistic Risk Assessment (PRA). Same technique-different name. In the early 1990's, all US nuclear plant licensees submitted plant-specific Individual Plant Examinations (IPE) for NRC review. The IPE considers realistic equipment failure rates and may include some human action and human error considerations. Even then, the analysis is still considered to be conservative since credit is not taken for all plant equipment or human actions that could mitigate accidents considered.

The NRC normally considers an upper acceptable risk to be 1 reactor accident resulting in core melt per 10000 reactor years of operation. The 100 reactors operating for 40 to 60 years would not be expected to sustain a core melt accident affecting the public.

The NRC subsequently imposed a regulation, 10 CFR 50.65, also referred to as the Maintenance Rule, that requires that the licensees establish performance criteria for risk significant equipment. That law became effective in July 1996. In the event that the equipment does not meet the performance criteria, maintenance improvement (or get well) programs must be initiated until the equipment again meets expectations.

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Copyright 1996-2006.  The Virtual Nuclear Tourist. All rights reserved. Revised: December 21, 2005.