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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

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