Neutronic study of ThF4-UF4-LiF fuel mixture in the molten salt hybrid reactor for 233U denaturing

This study investigates a fusion-fission hybrid reactor concept using a thorium-based molten salt fuel mixture to enhance proliferation resistance and operational sustainability. Neutron transport and reaction rates were modeled using the Monte Carlo N-particle code (MCNP6) with ENDF/B-VIII.0 data. Thorium is mixed homogeneously with 2.25 % depleted uranium (DU) in order to denaturate the 233U fuel. The analysis. showed that with 75 % 6Li enrichment and a 50 cm coolant layer, the tritium breeding ratio (TBR) remained above 1.05 for a period of four years. The energy multiplication factor (M) increased from 1.88 to 2.2, consistently exceeding the minimum target of 1.5. Under the hard fusion neutron flux, more than 96 % of the plutonium produced was 239Pu, heavier plutonium isotopes were burnt in situ. The production of the low enriched 233U fuel increased to about 12 % after 34 months. These results indicate the technical feasibility of a thorium-based fusion-fission hybrid reactor with improved proliferation resistance, efficient energy multiplication, and sustainable fuel cycle characteristics.