Neutronic Analysis of a Thorium-Uranium Fueled Water Cooled Fusion-Fission Hybrid Blanket

In this paper, a light water cooled fusion-fission hybrid reactor blanket fueled with thorium and uranium is presented. The major objective is to study the feasibility of this new concept with multi-purposes, including high energy gain, tritium self sufficiency and ²³³U breeding. The basic logic of this concept is to use the excess neutrons generated in the natural uranium fuel region to breed ²³³U in the thorium fuel region, while maintaining high energy amplifying factor (M) and tritium self-sufficiency. The guiding principle for the blanket design is to obtain a good neutron economy. The main method is to maximize the available neutrons and optimally distribute them in the blanket via competing processes of fission, tritium breeding and fissile fuel breeding by adjusting the neutron spectrum and system geometry. The COUPLE code developed by INET of Tsinghua University is used to simulate the neutronic behavior in the blanket. The simulation results show that a combined soft and hard neutron spectrum could yield M>15 while maintaining TBR>1.10 and conversion ratio of fissile materials (including ²³⁹Pu and ²³³U) CR>1.0 in a reasonably long refueling cycle (about 5 years). The results also demonstrates that under the constraint condition of tritium self sufficiency, this water cooled concept can only reach one optimized purpose at one time, energy gain M or ²³³U breeding.