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

Abstract

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 + BeF 2 molten salt mixture called Flibe, judging from their chemical, nuclear and heat-transfer properties. Li 0.17Pb 0.83 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 1.8 MCi/day and a target tritium leak rate of 10 Ci/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 D 2 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.