Assessment of FBR MONJU accident management reliability in causing reactor trips

Sotsu, Masutake ; Kurisaka, Kenichi 

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor which can supply 280 MW of electricity. The Accident Management (AM) in MONJU is based on three functions: the reactor trip function, the reactor liquid level retaining function, and the decay heat removal function. These are basic safety features, and it is necessary to evaluate the AM capability of these features quantitatively using a PSA technique. This paper describes the AM reactor trip evaluation method comprising plant transient response analysis using the Super-COPD code developed for a best estimate of the plant dynamics of MONJU, the results of this evaluation, and the results of simulator training of plant operators.