SBLOCA analysis for a 3-loop Westinghouse PWR reactor using TRACE

Abstract EN

There are two licensing approaches for evaluation of Loss of Coolant Accident (LOCA) consequences; conservative methodology and best estimate methodology.

Recently, the trend of nuclear reactor safety analysis reveals an increasing interest to substitute best estimate for conservative methodologies to achieve the safety margins and regulate the licensing and operations of nuclear reactors.

In the current study, the TRACE best estimate code was used to perform small break loss of coolant accident (SBLOCA) analysis for a 3-loop Westinghouse reactor.

Both cold-leg and hot-leg SBLOCAs, with 2 inch break size, were studied.

Also, a break size sensitivity study was performed to determine a limiting break size for the cold-leg SBLOCA.

Three additional cold- leg break cases were run with break sizes from 1 to 4 inches.

The research purposes are to enhance the understanding of thermal hydraulics during the SBLOCA transient and to perform a comparative study between break size and location in order to determine which accident scenario is the most limiting case regarding compliance of plant performance against applicable regulatory acceptance criteria.

The results show that 2 inch cold-leg break is the most limiting SBLOCA situation, since core heatup is encountered, and the highest value of the maximum peak fuel rod surface temperature of 913 K is detected