FOREWORD

Estimated frequencies of loss-of-coolant accidents (LOCAs; i. e., pipe ruptures as a function of break size) are used in a variety of regulatory applications, including probabilistic risk assessment (PRA) of nuclear power plants. Currently, the U. S. Nuclear Regulatory Commission (NRC) is using such information to establish a risk-informed alternative to the emergency core cooling system (ECCS) requirements in Title 10, Section 50.46, of the Code of Federal Regulations (10 CFR 50.46). Current requirements consider pipe breaks in the reactor coolant pressure boundary, up to and including breaks equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. One aspect of this risk-informing activity is to evaluate the technical adequacy of redefining the design-basis break size (the largest pipe break to which 10 CFR 50.46 applies) to a smaller size that is consistent with updated estimates of pipe break frequencies.

To provide the technical basis for a risk-informed definition of the design-basis break size, this study developed LOCA frequency estimates using an expert elicitation process. This process consolidated operating experience and insights from probabilistic fracture mechanics studies with knowledge of plant design, operation, and material performance. Expert elicitation is a well-recognized technique for quantifying phenomenological knowledge when modeling approaches or data are insufficient.

The results from the expert elicitation provide LOCA frequency estimates for piping and non-piping passive systems, as a function of effective break size and operating time through the end of the plant license — renewal period, for both boiling — and pressurized-water reactors (BWRs and PWRs, respectively). The panelists generally agreed on the important technical issues and LOCA-contributing factors. However, as expected, the panelists’ estimates exhibit both significant uncertainty and diversity. The uncertainty is reflected in the estimated parameters (mean, median, 5th and 95th percentiles) of the individual LOCA frequency distributions, and the diversity is captured by the confidence bounds on the group estimates. In addition, this study considered the sensitivity of the results to various analysis approaches. The results are most sensitive to the method used to aggregate the individual panelists’ estimates to obtain group estimates. In this study, geometric-mean aggregation produces group frequency estimates that approximate the medians of the panelists’ estimates and are also generally consistent with both operating experience and prior LOCA frequency estimates except where increases are supported by specific material aging-related concerns. However, arithmetic-mean and mixture-distribution aggregation are alternative methods that lead to significantly higher mean and 95th percentile group estimates.

Because the alternative aggregation methods can lead to significantly different results, a particular set of LOCA frequency estimates is not recommended for all risk-informed applications. The purposes and context of the application must be considered when determining the appropriateness of any set of elicitation results. In particular, during the selection of the BWR and PWR transition break sizes for the proposed 10 CFR50.46a rulemaking, the NRC staff considered the totality of the results from the sensitivity studies, rather than only the summary frequency estimates from this study. The NRC anticipates that a similar approach will be used in selecting appropriate replacement frequencies for the estimates provided in NUREG/CR-5750, “Rates of Initiating Events at U. S. Nuclear Power Plants: 1987 — 1995,” and for other applications that require frequencies for break sizes other than those in NUREG/CR-5750.

Jennifer L. Uhle, Director

Division of Engineering

Office of Nuclear Regulatory Research

U. S. Nuclear Regulatory Commission