Loss of Coolant Accident in Pressurized Water Reactor. Prediction of a 6-inch Cold Leg Break with Relap5 and Cathare 2

Abstract

This paper describes the approach to model Loss of Coolant Accident (LOCA) in the 900MWe Nuclear Power Plant Westinghouse PWR reactor.Two thermal-hydraulic system codes were applied: RELAP5 and CATHARE 2 to simulate steady state conditions and a transient evolution. The purpose of this paper is to analyse a 6-inch cold leg break in both codes and compare calculations performed in frame of the benchmark procedure between RELAP5 and CATHARE 2 codes. Obtained steady state and transient results for both codes are comparable and predict similar accident sequences. The results are not final, and the benchmark procedure is ongoing.