Behavior of PWR Reactor Coolant System Components, Other than Steam Generator Tubes, Under Severe Accident Conditions, Phase I Final Report (NUREG/CR-6792)

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Publication Information

Manuscript Completed: June 2002
Date Published: May 2003

Prepared by:
S. Majumdar, W. Shack

Argonne National Laboratory
9700 South Cass Avenue
Argonne, IL 60439

J. D. Page, NRC Project Manager

NRC Job Code Y6421

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

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A critical step in the assessment of risk of containment bypass attributable to pressure and temperature-induced failures of steam generator (SG) tubes during severe accidents is the prediction of the sequence in which the SG tubes fail relative to other reactor coolant system (RCS) components. This report summarizes our current understanding of the behavior of coolant system components other than SG tubes during severe accidents in pressurized water reactors (PWRs). A detailed analysis of RCS components during severe accidents reported In NUREG-1570 predicted that the failure times of steam generator tubes, the pressurizer surge line, and hot leg piping are very close. However, the analyses conducted for predicting failure of RCS components were less rigorous and detailed than those for SG tubes. This report reviews the methods used in NUREG-1570 to analyze the behavior of RCS piping, power operated relief valves, safety relief valves, and manway bolted connections during severe accidents and recommends future research and analyses that should be conducted to bring the failure prediction methodology on a par with that followed for SG tubes. This will make possible a more balanced assessment of the potential for containment bypass attributable to SG tube failures.

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