Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

Abstract

The U.S. Nuclear Regulatory Commission recently identified a possible safety concern for pressurized water reactors. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: (1) describes the important aspects of the problem, (2) provides an initial analysis of the consequences, and (3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of Westinghouse design, the concern is greatest for those plants. There is less concern for most plants of Combustion Engineering design, and likely no concern for plants of Babcock and Wilcox design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a Westinghouse pressurized water reactor. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed.