United States Nuclear Regulatory Commission - Protecting People and the Environment

Incorporating Aging Effects into Probabilistic Risk Assessment — A Feasibility Study Utilizing Reliability Physics Models (NUREG/CR-5632)

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Publication Information

Manuscript Completed: November 2000
Date Published: August 2001

Prepared by:
C. L. Smith, V. N. Shah
T. Kao*, G. Apostolakis*

Idaho National Engineering and Environmental Laboratory
Idaho Falls, ID 83415-3850

*ASCA, Inc.
755 Deep Valley Drive
Rolling Hills Estates, CA 90274

A. Buslik, NRC Project Manager

Prepared for:
Division of Risk Analysis and Applications
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code W6329

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Traditional probabilistic risk assessments (PRAs), in general, do not include passive SSCs (systems, structures, and components) since they are much more reliable than the active components. Aging phenomena, however, may make passive SSCs less reliable than assumed. Further, operation of power plants over durations of decades may lead to degradation of material properties and component strength, and eventually, if not mitigated, to aging-related failures. This report documents the feasibility assessment performed by the Idaho National Engineering and Environmental Laboratory for incorporating aging information and models directly into a PRA.

The work in this report focuses on corrosion, specifically flow-accelerated corrosion. We have evaluated the feasibility of modeling aging by incorporating a flowaccelerated corrosion model into PRA. The corrosion model that we used, the KWU model developed by Kastner and Riedle, estimates wall thinning. For this purpose, we have used the PRA model developed for the Surry Individual Plant Examination. In addition, we have employed a load-capacity model based upon a reliability physics model to estimate failure caused by flow-accelerated corrosion. Also, we have shown that a rigorous treatment of both aleatory and epistemic uncertainty is a vital part of the overall methodology. Successful demonstration of the methodology embodied in this report led to the realization of both expected and unexpected results. Expected results include the calculation of core damage frequency as a function of time due to the flow-accelerated corrosion aging mechanism. Unexpected results showed that attempting to model aging-related failures via a failure rate parametric model (e.g., a linear aging failure rate model) was incomplete. We have demonstrated that the results and insights gained from the U.S. Nuclear Regulatory Commission (NRC) Nuclear Power Aging Research Program and other NRC and industry programs related to materials degradation can be integrated into existing PRA models. However, our evaluation represents the feasibility of such integration and should not be construed to represent either relative or absolute magnitude of risk posed by flow-accelerated corrosion in pressurized water reactors.

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