Flibe-Tritium Research for Fission or Fusion Reactors at Kyushu University

There is increasing interest in using ionic molten-salt Flibe not only as self-cooled tritium(T)-breeding material in a fusion reactor blanket but also as fuel solvent of molten-salt fission reactors. Application of Flibe to T-breeding fluid for a stellarator-type fusion reactor operated at a high magnetic field brings large simplification of its blanket structure, allowing continuous operation under high-beta plasma conditions. Using mixed Flibe-ThF4+UF4 fuel in molten salt fission reactors permits stable long-term operation without fuel exchange. When Flibe or Flinak is irradiated by neutrons, however, acid and corrosive TF is generated, and some T permeates through structural walls. In order to solve these problems, chemical conditions of Flibe are changed using the redox-control reaction, Be+2TF=BeF2+T2. In addition, permeation of hydrogen isotopes is lowered by enhancing T recovery rates. Part of Flibe-tritium researches are performed at Idaho National Laboratory (INL) under the Japan-US collaboration work of JUPITER-II. Our own contributions to the topics are shortly introduced in this paper.