Abstract
CFETR (Chinese Fusion Engineering Testing Reactor) is a superconducting Tokamak device. The simulation was performed on the cooling circuit inside water cooled breeder blanket which is one of the breeding blanket candidates for CFETR to compare the thermal hydraulic characteristics based on superheated steam and PWR (pressurized water reactor) water conditions. The work is carried out using MELCOR code. As a result, under the optimal flow distribution, it turns out that the thermal hydraulic characteristics meet the design requirements and the nuclear heat inside blanket module as well as heat flux from plasma could be removed completely by the cooling circuits on both conditions. It's also found that the temperature of first wall (FW) is higher on PWR condition than superheated steam. But due to the generation of superheated steam, the wall temperature would grow significantly in the backward zone of cooling circuit. Based on the steady state, LOFA (loss of flow accident) was simulated. It was predicted that FW will melt anyway due to the high nuclear heat without considering thermal radiation. As for different ways of FPSS, a 100s delay is invalid and a 3s delay could buy us time to take measures.
