Preliminary Analysis of In-Vessel First Wall Cooling Pipe Ruptures for ITER

Abstract

Similar to fission reactors, accidents deviated from normal operation may occur in fusion reactors, in which the progressions of the accidents should be analyzed for obtaining the state parameters for further study and suggestions for the mitigation measures. Due to the lack of proper accident analytical code specifically for fusion reactor, an integral safety analytical code developed for fission reactor is adopted for the accident analysis for International Thermonuclear Experimental Reactor (ITER). The accident analytical model for ITER is established, which consists of the vacuum vessel (VV), first wall (FW), blanket and its primary heat transfer system and the pressure suppression system. The loss of coolant accident induced by the in-vessel FW cooling pipe double-ended rupture is selected with two different cases of single and multiple pipes rupture as the initial events to investigate the thermal–hydraulic responses for the systems. The analytical results show that for the accidents of in-vessel FW cooling pipe double-ended rupture, the pressure in VV increases with the coolant releasing, and then decreases with the rupture disk open and remains in the design limitation, which is compared with the results in accident analysis report and suggest that the integral safety analytical code can be used to simulate the accident progression in fusion reactor.