Best estimate plus uncertainty analysis of a small‐break LOCA on an advanced Generation‐III pressurized water reactor

To improve the economy of nuclear power plants, a passively pressurized water reactor with additional 40% power output was developed based on AP1000 technologies. In this study, the RELAP5 code was used to model the reactor system components involved in the accident transient, and a postulated 2-in. small break loss of coolant accident (SBLOCA) was simulated and analyzed. The evaluation of the code applicability was conducted for the development models and special process models. With the base input deck, a best estimate plus uncertainty (BEPU) analysis was performed. The significant contributing parameters to reactor safety criteria were identified according to the phenomenon identification and ranking table. The DAKOTA software was used for simple random sampling, and a set of 93 calculations were performed based on the Wilks' formula. Important input uncertainty contributors for the predicted minimum core collapsed level were identified through the spearman rank correlation coefficient and partial rank correlation coefficient.