TRITIUM RECOVERY FROM LIQUID BLANKET SYSTEMS FOR FUSION REACTOR

摘要

Systems to recover tritium (T) from fusion liquid blankets of LiPb (Li_(17)Pb_(83)) and Flibe (Li_2BeF_4) under safety conditions are discussed based on available data of T transport properties. We recently performed experiments on hydrogen isotopes permeating through LiPb in Kyushu University and introduced them in the present paper. Solubilities, diffusivities and permeabilities of H and D through Li_(17)Pb_(83) are determined by means of a transient permeation method. Their isotope effects between H and D in the Li-Pb eutectic alloy are found to be independent ofD/H atomic ratio and a function of only temperature. Tritium recovery apparatuses of a liquid-gas counter-current flowing bed or a T permeation window are designed in the liquid blanket system composed of the primary coolant of Flibe or LiPb and the secondary coolant of He or water. Simultaneous transfer of heat and T needs new configuration in order to satisfy the necessary conditions of low T leak and high heat transfer, which make it possible to operate the blanket effectively and safely. Tritium permeation rates through the primary LiPb or Flibe breeder and secondary water or He coolant loops are estimated, which is the most important factor to operate the blanket safely.