Uncertainty quantification for risk assessment of loss-of-coolant accident frequencies in nuclear power plants

textThis research presents the methodologies used to resolve the Nuclear Regulatory Commission Generic Safety Issue 191. The presented results are specific to South Texas Project Nuclear Operating Company (STPNOC). However, the proposed methodologies may be applicable to other nuclear power plants given the appropriate plant-specific frequencies. This research provides important inputs to CASA Grande, a computer program used to model physical phenomena and quantify uncertainties to obtain estimates of failure probabilities for post-loss-of-coolant accident events at the STPNOC containment. We provide modeling and sampling methods for loss-of-coolant accident (LOCA) frequencies and break sizes. We focus on a study known as NUREG-1829 (Tregoning et al., 2008), which includes an expert elicitation of quantiles governing the (annual) frequency of a LOCA in boiling water reactors and pressurized water reactors. We propose to model LOCA frequencies with bounded Johnson distributions and to sample break sizes using uniform distributions. We then develop a new method to distribute LOCA frequencies to different locations within a plant to account for the location-dependent differences while preserving the NUREG-1829 frequencies. We also propose to linearly interpolate the NUREG-1829 LOCA frequencies to obtain the frequencies for any break sizes other than those from NUREG-1829. In addition, we present a method to obtain the distribution of LOCA frequency within a break-size interval providing important inputs to the probabilistic risk assessment quantification for STPNOC. We review methods of combining the probability distributions of multiple experts to obtain a single probability distribution. More specifically, we describe the relative merits of the arithmetic mean (AM) and geometric mean (GM) as ways of performing this aggregation in the context of probabilities associated with rare events. Examining a set of pressurized water reactor results from NUREG-1829, we conclude that the GM represents a consistently sensible notion of the middle of the opinions expressed by nine experts. We further conclude that the AM is inappropriate for representing the center of the group's opinion for large effective break sizes. Instead, as the break size grows large a single expert's opinion dominates the combination produced by the AM.Mechanical Engineerin