Analysis of the behavior of the Atucha-I nuclear power plant in a pressurized thermal shock scenario

Because of the results obtained with the thermal hydraulic code RELAP5/MOD3, in a LOCA scenario in a hot leg of the core cooling primary system of the nuclear power plant Atucha I, a study is made, based on lumped parameters in the local core flow temperature evolution in the pressure vessel downcomer. The main object is to know how is the stress intensity factors variation, respect to the injected emergency water temperature, during thermal hydraulic transients, which have a place under the imposed conditions in the postulated scenario. Particularly, it is considered a 50 cm2 break (the most critical), near the connection between the pressurized and the hot leg. From this study, it is possible to see, how the emergency water temperature can mitigates the thermal shock effects on the pressure vessel walls. (author)