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Home / Publications / Journals / Fusion Science and Technology / Volume 56 / Number 2 / Pages 632-640

Molten Salt Fuel Version of Laser Inertial Fusion Fission Energy (LIFE)

R. W. Moir, H. F. Shaw, A. Caro, Larry Kaufman, J. F. Latkowski, J. Powers, P. E. A. Turchi

Fusion Science and Technology / Volume 56 / Number 2 / August 2009 / Pages 632-640

Laser Fusion-Fission Hybrid / Eighteenth Topical Meeting on the Technology of Fusion Energy (Part 2)

Molten salt with dissolved uranium is being considered for the Laser Inertial Confinement Fusion Fission Energy (LIFE) fission blanket as a backup in case a solid-fuel version cannot meet the performance objectives, for example because of radiation damage of the solid materials. Molten salt is not damaged by radiation and therefore could likely achieve the desired high burnup (>99%) of heavy atoms of 238U. A perceived disadvantage is the possibility that the circulating molten salt could lend itself to misuse (proliferation) by making separation of fissile material easier than for the solid-fuel case. The molten salt composition being considered is the eutectic mixture of 73 mol% LiF and 27 mol% UF4, whose melting point is 490°C. The use of 232Th as a fuel is also being studied. (232Th does not produce Pu under neutron irradiation.) The temperature of the molten salt would be ∼550°C at the inlet (60°C above the solidus temperature) and ∼650°C at the outlet. Mixtures of U and Th are being considered. To minimize corrosion of structural materials, the molten salt would also contain a small amount (∼1mol%) of UF3. The same beryllium neutron multiplier could be used as in the solid fuel case; alternatively, a liquid lithium or liquid lead multiplier could be used. Insuring that the solubility of Pu3+ in the melt is not exceeded is a design criterion. To mitigate corrosion of the steel, are fractory coating such as tungsten similar to the first wall facing the fusion source is suggested in the high-neutron-flux regions; and in low-neutron-flux regions, including the piping and heat exchangers, a nickel alloy, Hastelloy, would be used. These material choices parallel those made for the Molten Salt Reactor Experiment (MSRE) at ORNL. The nuclear performance is better than the solid fuel case. At the beginning of life, the tritium breeding ratio is unity and the plutonium plus 233U production rate is ∼0.6 atoms per 14.1 MeV neutron.

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