Neutronics analysis of thorium-uranium based molten salt blanket in stellarator-type FFHR

Abstract

• Detailed CAD model has been designed for thorium-uranium based molten salt blanket. • Complex geometry structure has been analyzed in coupling with Monte Carlo calculation. • Neutronics features have been simulated in consideration of the burnup effect. • The thickness and the 6 Li enrichment have been optimized in the simulation process. The stellarator-type Fusion-Fission Hybrid Reactor (FFHR) is an alternative conceptual design for the tokamak-based devices in order to realize the stable power output. It can bring the viable energy source for the fusion technology in a foreseeable future, and has been widely studied with a lot of advanced concepts being proposed. However, there are some challenges in the neutronics design of the FFHR blanket for the purpose of obtaining the appropriate Tritium Breeding Ratio (TBR) and blanket energy multiplication factor, since the blanket system has a complicated three-dimensional structure in helical shape, which can not be easily translated into native geometry representation syntax of most Monte Carlo neutron transport codes. In this paper, a liquid fuel, F-Li-Be molten salt, with uranium and thorium dissolved has been proposed for the preliminary design of stellarator-type FFHR blanket. Besides, the CAD-PSFO code [ 19 ] has been employed for the simulation of complex geometry structure in the helical blanket, and the OMCB package [ 22 ] has been used in the burnup analysis of the neutronics features in the molten salt blanket. Moreover, detailed geometry design of the blanket arrangement and thickness have been studied in order to meet the requirement of energy multiplication factor in the stellarator-type FFHR. Also, the composition and concentration of the fuel material and 6 Li in molten salt have been discussed for the purpose of tritium self-sufficiency with steady energy supply in the whole lifetime.