Assessment of the neutronic performance of some alternative fluids in a fusion-fission hybrid reactor by using Monte Carlo method

This study analyzes the nuclear parameters such as neutron flux, tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fissile fuel breeding in liquid first wall, blanket and shield zones and investigates effects of spent fuel grade Pu content in the designed system on these nuclear parameters. Three-dimensional analyses were performed by using the Monte Carlo code MCNPX-2.7.0 and nuclear data library ENDF/B-VII.0.