Current status of thermal hydraulics for technological development of In-vessel components of fusion reactor

Ezato, Koichiro; Akiba, Masato

In-vessel components such as Blanket and Divertor in a fusion reactor has a function of exhausting high heat and particle loads in order to maintain the structural soundness of the reactor. In the International Thermonuclear Test Reactor called ITER, build by ITER Organization under the framework of collaboration of seven parties including Japan, first wall of blanket and diverter receive as high heat flux as 0.5 MW/m and 20 MW/m, respectively. This paper intends to present current status of technological development of in-vessel components from the viewpoint of thermal hydraulics.