The Study of Loss-of-Coolant Accident Parameter Identification for a BWR-4 Plant

Institute of Nuclear Energy Research (Taiwan)

Abstract

In this study the parameter identification tool (PIT) for the loss-of-coolant accident (LOCA) of a BWR-4 plant with Mark-I containment, Chinshan nuclear power plant (NPP) in north Taiwan, is developed. The PIT is composed of the Simplex search algorithm and the MAAP5 code with the integration approach presented by Tsai et al. The postulated LOCA is assumed to occur after station blackout sequence with the break elevation of 17 m and the break area of 2.1×10^-2 m². Through the developed PIT, the break elevation of 16.93 m and the break area of 2.1×10^-2 m² are identified with the imitated plant data, the RPV pressure and the shroud water level, generated by MAAP5 code. The timing of events and source terms in the LOCA of Chinshan NPP can be predicted with by the MAAP5 code with the identified parameters. It demonstrates that the PIT is helpful and important in severe accident management.