Transition to thorium fuel cycle in a small modular molten salt reactor based on a batch reprocessing mode

Short life fission products extracted from molten salt reactor fuel for radiopharmaceutical applications

This work studies the potential of using short life fission product (AFp) radioisotopes e.g. 82Br, 86Rb, (90Sr) - 90mY, (99Mo) - 99mTc, 103Ru - 103mRh, 111Ag, 127Sb - 127(m)Te, 126I, 131I, 133Xe, 136Cs, 141Ce, 143Ce, 143Pr, 147Nd - 147Pm, 149Pm, 153Sm, 156Eu, 159Gd and 161Tb, extracted from a molten salt reactor and their separation using specific thermodynamic and radiochemical conditions. Their utilisation for coupled radiodiagnostics and radiotherapy is a key consideration. A molten salt reactor produces fission products during operation. These radioisotopes can be separated at line from the liquid fuel by evaporation/distillation, chemical reduction (using H2 doped gas), electro-deposition and/or chemical oxidation (using Cl2 doped gas). They can be refined and chemically treated for radiopharmaceutical use for imaging and radiodiagnostics utilising γ radioscopy or positron emission tomography, and potentially in radiotherapy to target specific cancers or viral diseases using β- emitters. Some of the AFp isotopes are currently used for radiodiagnostics because they emit γ rays of energy 50-200 keV. However, some may also be used in parallel for radiotherapy utilising their β- (EMean ≈ 100 keV) emission whose mean free pathway of c.a. 100 nm in biological tissue is much smaller than their penetration depth. Focus is given to 86Rb, 90Y, 99mTc, 131I and 133Xe as well as on the ALn isotopes (141Ce, 143Ce - 143Pr, 147Nd - 147Pm, 149Pm and 153Sm) because of their strong potential for complexation with bio-ligands (e.g. DOTA) or for their ability to form micro-nano-spheres, and because of their potential for dual radiodiagnostics and radiotherapy. It is shown that these radio-lanthanides could also replace 177Lu for the treatment of specific cancers.

Analysis of thorium and transuranium utilization in a small modular reactor

This study presents neutronic analyses of a small modular reactor utilizing transuranium and thorium. Two different fuel cases are considered in the analyses as the transuranium extracted from PWR-MOX spent fuel (a form of a mixture of minor actinide and Pu isotopes) (Case A) and 4.5 % enriched UO2 with ThO 2(the form of separate fuel rods) (Case B). The total power of the considered small modular reactor containing 69 assemblies is 450 MW thermal. In both fuel cases, the time-dependent critical burnup calculations are carried out by using MCNPX 2.7 code until their effective neutron multiplication factors decrease to 0.99. The calculations bring out that the small modular reactor can operate for quite a long time without refueling and that a new fuel with a richness of 1.05 % can be obtained from ThO as well 2 as energy production.