United States Nuclear Regulatory Commission - Protecting People and the Environment

SCDAP/RELAP5 Thermal-Hydraulic Evaluations of the Potential for Containment Bypass During Extended Station Blackout Severe Accident Sequences in a Westinghouse Four-Loop PWR (NUREG/CR-6995)

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Publication Information

Manuscript Completed: February 2010
Date Published: March 2010

Prepared by:
C.D. Fletcher, R.M. Beaton, V. V. Palazov, and D.L. Caraher
R.W. Shumway (Consultant)

Information Systems Laboratories, Inc.
10070 Barnes Canyon Road
San Diego, CA 92121

C. Boyd, NRC Point of Contact

Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code Y6198

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Abstract

The U.S. Nuclear Regulatory Commission has been conducting studies to evaluate the risk associated with steam generator tube failure during low probability severe accidents in pressurized-water reactors (PWRs) employing U-tube-type steam generators as part of the agency's Steam Generator Action Plan (ML 011300073). The evaluations focus on station blackout events that include a series of unlikely events and conditions that result in a "high-dry-low" plant condition. The high-dry-low condition refers to high primary side pressure along with a dried-out steam generator that is at low pressure. Failures of hot leg piping, pressurizer surge-line piping, and the reactor vessel lead to discharge of fission products into the containment. Failure of steam generator tubes prior to the failure of one of these other components leads to discharge of some fission products into the steam generator secondary system from where they may be discharged to the environment through the pressure-relief valves. This latter sequence is potentially more risk-significant since it involves a containment bypass scenario. The relative timing of these structural failures therefore affects the event sequence and whether the containment is bypassed.

This report summarizes thermal-hydraulic evaluations performed using the SCDAP/RELAP5 systems analysis code and a model representing a Westinghouse four-loop PWR. An assessment is made of the SCDAP/RELAP5 capabilities for predicting the phenomena and behavior important for this application. The plant model has benefitted from (1) extensive iterative comparisons with evaluations of natural circulation flows and turbulent mixing using a computational fluid dynamics code and (2) from comparison with experimental data for pertinent fluid-mixing behavior. This report documents the recent history of SCDAP/RELAP5 model improvements for this application, numerous sensitivity evaluations, estimates of the uncertainties in the calculated results, analyses of extended station blackout accident event sequences in a Westinghouse four-loop PWR, and a categorization of those event sequences based on containment bypass outcome.

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