Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

• Published in: 2015 IEEE 26th Symposium on Fusion Engineering (SOFE) pp. 1 - 6
• Main Authors: Watanabe, Kazuhito, Nakamura, Makoto, Tobita, Kenji, Someya, Youji, Tanigawa, Hisashi, Utoh, Hiroyasu, Sakamoto, Yoshiteru, Araki, Takao, Asano, Shiro, Asano, Kazuhito
• Format: Conference Proceeding
• Language: English
• Published: IEEE 01.05.2015
• Subjects: accident scenario analysis; Accidents; Coolants; Heating; Inductors; safety study; safety system; Tokamaks; water-cooled fusion DEMO
• ISSN: 2155-9953
• DOI: 10.1109/SOFE.2015.7482333

Safety studies of a water-cooled fusion DEMO reactor have been performed. In the DEMO design, the blanket primary cooling system involves a large amount of energy due to pressurized water coolant (290-325 °C, 15.5 MPa). Moreover, it contains radioactive materials such as tritium and activated corrosion products. Therefore, in the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three options of confinement strategies. In each option, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to the environment were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.