Tritium recovery system for Li–Pb loop of inertial fusion reactor

The best material for a wet wall and blanket of an inertial fusion reactor is selected among Li, eutectic alloys of Li–Pb, Li–Sn and a 2LiF+BeF2 molten salt mixture called Flibe, judging from their chemical, nuclear and heat-transfer properties. Li0.17Pb0.83 is found to be the most promising one because of low Li vapor pressure, moderate melting temperature, good heat-transfer properties under the condition of a KOYO-fast circulation loop operated between 300 and 500°C. A counter-current extraction tower packed with metallic rashig rings is proposed to extract tritium generated and dissolved in the Li–Pb eutectic alloy. Mass-transfer parameters when He and Li–Pb flow counter-currently through the tower packed with the rings are determined to satisfy the two necessary conditions of a self-sufficient tritium generation rate of 1.8MCi/day and a target tritium leak rate of 10Ci/day. It is found that the height of a tower to achieve the 99.999% recovery is comparatively low because of the promising property of a large equilibrium pressure of tritium. In order to mitigate the disadvantage of its high density, which needs a large pumping power, a porous packing material that keeps good contact between He and Li–Pb should be developed in the future. It is found experimentally that D2 addition in He purge gas is effective to achieve a faster rate of tritium recovery from the Li–Pb flow. The rate-determining step of tritium permeation through a steam generator is determined as a function of a Li–Pb flow rate in a stainless-steel heat-transfer tube.