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Comprehensive Safety Analysis Code System for Nuclear Fusion Reactors II: Thermal Analyses During Plasma Disruptions for International Thermonuclear Experimental Reactor

T. Honda, K. Maki, T. Okazaki, T. Uda, Y. Seki, I. Aoki, T. Kunugi

Fusion Science and Technology

Volume 26 / Number 4 / December 1994 / Pages 1288-1295


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Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 µm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively.

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