Progress in thermohydraulic analysis of accident scenarios of a water-cooled fusion DEMO reactor

Nakamura, Makoto; Watanabe, Kazuhito; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*; Kunugi, Tomoaki*; Yokomine, Takehiko*; Gulden, W.*

We report recent progress in thermohydraulic analysis of several types of accidents of a water-cooled fusion DEMO reactor. We particularly studied (1) in-vessel (in-VV) loss-of-coolant accident (LOCA) of the first wall (FW) cooling pipes, (2) ex-vessel (ex-VV) LOCA of the primary cooling system, (3) LOCA in blanket modules (in-box LOCA) and (4) loss-of-vacuum accident (LOVA). The analysis identified transient responses of safety-class or safety-important reactor components and structures of the DEMO to these accidents. The pressure loads to the barriers confining radioactive materials were also evaluated. On a basis of these analysis results, strategies to confine the radioactive materials, i.e. tritium and activated tungsten dust, against these accidents were assessed.