Neutronic investigation of the application of certain plutonium-mixed fluids in a fusion-fission hybrid reactor

The nuclear parameters of a fusion-fission hybrid reactor such as neutron flux, heating, fission reaction rate were investigated according to the mixture components, radial energy spectrum in the designed system. In this study, the effect of spent fuel-grade Pu content in the designed system on these nuclear parameters were calculated by using the three-dimensional Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.