Preliminary assessment of the safety performance of Westinghouse LFR

Abstract

The safety performance of Westinghouse Lead Fast Reactor (LFR) is evaluated using the SAS4A/SASSYS-1 system code coupled to the GOTHIC containment code. The Westinghouse LFR is a 950 MW(th) lead-cooled, pool-type reactor with microchannel primary heat exchangers (PHEs) directly immersed in the primary coolant. The reactor vessel (RV) is surrounded by a guard vessel (GV) to safeguard against the unlikely event of reactor vessel failure. The emergency decay heat removal is performed by the passive heat removal system (PHRS), which consists of a water pool system surrounding the GV filled with enough water to remove decay heat for the first seven days, and stacks to circulate air and remove decay heat indefinitely after the water has boiled off. The responses of the reactor system to Station Blackout (SBO) and Transient Overpower (TOP) are analyzed. During SBO, when the normal decay heat removal is assumed unavailable, the passive water-cooling transitioning to air-cooling outside of GV successfully removes the decay heat indefinitely. During TOP, where inadvertent withdrawal of most reactive control rods is postulated, the negative reactivity feedback of the core limits the power excursion to 140% of the normal operating power before the passive shutdown system is actuated and scrams the reactor. Fuel melting and cladding failure are avoided. The reactor design is evolving, and the preliminary results presented in the paper demonstrate the capability of the coupled SAS4A/SASSYS-1 and GOTHIC codes and characterize the safety performance of the Westinghouse LFR.