The large-break LOCA uncertainty analysis in a VVER-1000 reactor using TRACE and DAKOTA

Abstract

An uncertainty analysis is inevitable in simulating reactor performance, particularly in the accident analysis for the estimation of safety borders. In this paper, the application of the uncertainty analysis of a Large Break Loss of Coolant Accident (LBLOCA) from main coolant pipeline at the reactor inlet and outlet, and also pressurizer (PRZ) connecting pipe are presented. First, the model developed in TRACE is verified in steady-state conditions and LBLOCA using the Final Safety Analysis Report (FSAR) of Bushehr nuclear power plant. By the verified model, the uncertainty analysis is made respecting uncertainty in boundary conditions. Finally, the analysis is achieved using the surrogate models and Wilks’ method in DAKOTA. The 95/95 uncertainty band of the peak cladding temperature (PCT) and main reactor parameters are obtained and analyzed. The results indicate that the PCT uncertainty in LBLOCA at the reactor inlet is large, the PCT varies in the range of 910 to 970 °C. The drop temperature time of fuel cladding is in the range of 270 to 600 s for LBLOCA at the reactor inlet. The PCT uncertainty in LBLOCA at the reactor outlet and PRZ connecting pipe is low.