Uncertainty calculation in small break LOCA in the emergency core cooling system connected to the hot leg of Angra 2 nuclear power plant

Abstract

Owing to the occurrence of nuclear accidents, worldwide nuclear regulatory organisations included the analysis of accidents considered as design basis accidents - Loss of Coolant Accident (large and small-break, LBLOCA or SBLOCA) - in the safety analysis reports of nuclear facilities. In Brazil, the tool selected by the licensing authority, Comissao Nacional de Energia Nuclear (CNEN), is RELAP5 Code. The aim of this paper is the evaluation of the performance of the Emergency Core Cooling System (ECCS) of Angra 2 nuclear reactor during SBLOCA. In this study, the RELAP5 code and the Code Internal Assessment of Uncertainty (CIAU) were used to simulate and analyse the uncertainties of the results. The postulated accident is the SBLOCA in the hot leg connected to the ECCS described in the Final Safety Analysis Report of Angra 2 (FSAR/A2). The results from this study were satisfactory when compared with the FSAR/A2.