Neutronic study of ThF<sub>4</sub>-UF<sub>4</sub>-LiF fuel mixture in the molten salt hybrid reactor for <SUP>233</SUP>U denaturing

ŞAHİN H. M., Sahin S., TUNÇ G., Sahinen H.

PROGRESS IN NUCLEAR ENERGY, cilt.193, 2026 (SCI-Expanded, Scopus) identifier

  • Yayın Türü: Makale / Tam Makale
  • Cilt numarası: 193
  • Basım Tarihi: 2026
  • Doi Numarası: 10.1016/j.pnucene.2026.106237
  • Dergi Adı: PROGRESS IN NUCLEAR ENERGY
  • Derginin Tarandığı İndeksler: Science Citation Index Expanded (SCI-EXPANDED), Scopus, Compendex, Environment Index, INSPEC
  • Gazi Üniversitesi Adresli: Evet

Özet

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.