Comparative Evaluations of Nuclear Fuel Breeding Rate in Hybrid Fission-Fusion Reactor with U and Th

A nuclear fuel-breeding blanket of a hybrid fission-fusion reactor is considered in this work. The blanket design uses the tight layout of cylindrical heat-generating elements. The neutronic analysis of nuclear fuel breeding in a blanket is performed for source materials (metallic uranium and thorium, uranium dioxide, thorium dioxide, uranium nitride, and thorium nitride) in combination with various coolants—lead, sodium–potassium eutectic, carbon dioxide, water, steam–water mixture, and heavy water. The 239 Pu breeding rate is about 6 kg/(m 2 yr) at the neutron wall load of 0.4 MW/m 2 or 15 kg/(MW yr) being normalized to the neutron power on the first wall in front of the fuel part of a blanket (for the most efficient combinations of initial material and coolant). The 233 U breeding rate is about 3.5 kg/(m 2 yr) at the neutron wall load of 0.4 MW/m 2 or 9 kg/(MW yr) being normalized to the neutron power on the first wall in front of the fuel part of a blanket (for the most efficient combinations of initial material and coolant).