TY - GEN
T1 - Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor
AU - Watanabe, Kazuhito
AU - Nakamura, Makoto
AU - Tobita, Kenji
AU - Someya, Youji
AU - Tanigawa, Hisashi
AU - Utoh, Hiroyasu
AU - Sakamoto, Yoshiteru
AU - Araki, Takao
AU - Asano, Shiro
AU - Asano, Kazuhito
N1 - Publisher Copyright:
© 2015 IEEE.
PY - 2016/5/31
Y1 - 2016/5/31
N2 - 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.
AB - 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.
KW - accident scenario analysis
KW - safety study
KW - safety system
KW - water-cooled fusion DEMO
UR - http://www.scopus.com/inward/record.url?scp=84978796210&partnerID=8YFLogxK
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U2 - 10.1109/SOFE.2015.7482333
DO - 10.1109/SOFE.2015.7482333
M3 - Conference contribution
AN - SCOPUS:84978796210
T3 - Proceedings - Symposium on Fusion Engineering
BT - 2015 IEEE 26th Symposium on Fusion Engineering, SOFE 2015
PB - Institute of Electrical and Electronics Engineers Inc.
T2 - 26th IEEE Symposium on Fusion Engineering, SOFE 2015
Y2 - 31 May 2015 through 4 June 2015
ER -