Abstract
A concept is offered for tokamak and molten-salt thorium blanket with liquid-metal mass-exchanger which allows an uninterrupted extraction of protactinium from the blanket and its accumulation in a cascade salt trap separately from fission products. At uninterrupted extraction of protactinium from neutron field with the same rate, such a facility can become attractive for industrial production of nuclear fuel ( 233 U) from thorium. For this, it is offered to use a reduction extraction of radio-nuclides into a liquid-metal carrier (directly contacting with the molten salt) by managing RedOx potential (Fermi level) of the salt composition. Establishing Fermi level in the first cascade of molten-salt trap only for oxidizing the lanthanides allows extracting only them from the liquid-metal carrier. In the second cascade of this trap, one can extract protactinium by shifting down Fermi level at higher oxidation potential. For correct operation of the trap cascades, the lanthanides portion in the second cascade will be less than 0.01% of the first one and the portion of protactinium in the first cascade will be four orders less than in the second one.
Highlights
The hybrid of tokamak, molten-salt thorium blanket, and liquid-metal mass-exchanger can become effective for producing 233U from thorium as a nuclear fuel for thermal reactors
Numerical calculations have shown that the power of neutron flux of neutrons may be obtained up to ~15 MW at deuterons injection of 23 MW and the density of this flux on blanket surface will be ~0.2 MW/m2
Numerical calculations have shown that the power of neutron flux of neutrons may be obtained up to ~15 MW at deuterons injection of 23 MW and the density of this flux on blanket surface will be ~0.2 MW/m2 that can provide producing protactinium in the salt composition of 0.7FLiNaK–0.3ThF4 up to 4 kg/m2 a year
Summary
The hybrid of tokamak (as a thermonuclear source of neutrons), molten-salt thorium blanket, and liquid-metal mass-exchanger can become effective for producing 233U from thorium as a nuclear fuel for thermal reactors. In this system, one can exclude uncontrollable reactor runaway and feasible radiation of the environment at coolant loss due to continuous and uninterrupted removing radio-nuclides from a reactor core. Numerical calculations have shown that one can obtain a neutron flux of ~15 MW and its power density in the blanket surface of ~0.2 MW/m2 at the injection of double-weight hydrogen by power of 23 MW It can provide a rate of protactinium production in the salt composition of 0.7FLiNaK–0.3ThF4 up to ∼4·10-7 mol/s
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