| Summary: | Background238Pu is widely used in the aviation field as the heat source of radioisotope thermal generators (RTGS). It can be produced in a new type molten salt reactor with chloride molten salt target, which has the advantages in terms of economy.PurposeThis study aims to analyze the production of 238Pu in the thermal neutron irradiation channel of the new type molten salt fast reactor.MethodsAccording to the core physical calculation model of new type molten salt fast reactor, standardized computer analyses for licensing evaluation version 6.1 (SCALE6.1) program was employed to analyze the effects of different target matrix and target parameters on the 238Pu production performance, and calculate the 237Np reaction cross sections, target insertion reactivity effect and the concentration of 236Pu in different irradiation channels. The purity and yield of 238Pu with the irradiation time were calculated in details.ResultsThe results show that the 237Np conversion rate of NpCl4 molten salt target matrix is high, and the 237Np utilization rate can be improved by reducing the target radius. The 236Pu concentration generated by the target in the irradiation channel can be reduced to less than 1×10-7 whilst the purity of 238Pu exceeds 98%. When the irradiation cycle is 40 d, the annual yield of 238Pu is 0.8 kg, and the conversion rate of 237Np reaches 99.4%.ConclusionsTherefore, using chloride molten salt target to produce 238Pu in the new type molten salt fast reactor can obtain the high purity of 238Pu and 237Np conversion rate and low concentration of 236Pu.
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