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

Abstract

In this study, a fusion–fission hybrid reactor system was designed by using 9Cr2WVTa Ferritic steel structural material and the molten salt-heavy metal mixtures 99–95% Li20Sn80-1–5% SFG-Pu, 99–95% Li20Sn80-1–5% SFG-PuF4, or 99–95% Li20Sn80-1–5% SFG-PuO2, as fluids. The fluids were used in the liquid first wall, blanket and shield zones of a fusion–fission hybrid reactor system. Beryllium (Be) zone with the width of 3cm was used for neutron multiplication between the liquid first wall and blanket.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.