Risk-Informed Assessment of Degraded Containment Vessels (NUREG/CR-6920)
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Publication Information
Manuscript Completed: June 2006
Date Published: November 2006
Prepared by:
B.W. Spencer, J.P. Petti, D.M. Kunsman
Sandia National Laboratories
Operated by Sandia Corporation for the
U.S. Department of Energy
P.O. Box 5800
Albuquerque, NM 87185-0744
H.L. Graves, III, NRC Project Manager
Prepared for:
Division of Fuel, Engineering and Radiological Research
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
Job Code Y6164
Availability Notice
Abstract
Various forms of degradation have been observed in the containment vessels of a number of operating nuclear power plants in the United States. Examples of degradation include corrosion of the steel shell or liner, corrosion of reinforcing bars and prestressing tendons, loss of prestressing, and corrosion of bellows. The containment serves as the ultimate barrier against the release of radioactive material into the environment. Because of this role, compromising the containment could increase the risk of a large release in the unlikely event of an accident.
It is possible to assess the effect that degradation has on the pressure retaining capacity of the containment vessel by performing structural analysis using a model that takes into account the effects of the degradation. While such an analysis provides useful information about the effects of the degradation on the structural capacity of the containment, it does not necessarily provide a perspective on the effect that the degradation could have on the risk associated with operating a nuclear power plant. To understand the risk, one must take into account both the structural capacity of the containment and the probabilities of occurrence of a variety of loading scenarios.
In this study, structural analysis results are integrated with risk models to gain a risk-informed perspective on the issue of containment degradation. Risk models require a probabilistic description of the capacity of the containment to resist a number of different failure modes. Latin hypercube sampling is used to develop sets of inputs for detailed finite element models of the containment in both its original and degraded condition. Probabilistic descriptions of the containment capacity are obtained from the results of the structural analysis models, and used as input for the risk model. By using this approach, both the risk and the change in risk associated with the degradation can be measured.
The approach described above has been applied to case studies of containment degradation at four “typical” U.S. nuclear power plants. These include a Pressurized Water Reactor (PWR) with a large, dry reinforced concrete containment, a Boiling Water Reactor (BWR) with a Mark I steel containment, a PWR with a steel ice condenser containment, and a PWR with a large, dry prestressed concrete containment. In this study, each of these containments is subjected to a number of hypothetical degradation scenarios. While this degradation is typical of that which has been observed in actual plants, it is not implied that such degradation was actually observed in the plants studied here.
The risk analysis results reflect the effects of degradation and are evaluated with respect to the guidelines given in the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.174. Recommendations are made on the appropriateness and limitations of this approach. Integrating structural analysis with risk analysis can give plant operators and regulators improved information to make the best decisions on how to deal with containment degradation.
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