Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket

Abstract

Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO 2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57 MW/cm 2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of 233Pa on the 233U production and k eff are also discussed.