Accuracy of Multi-Group Transport Calculation in D-T Fusion Neutronics

Abstract

Both multi-group and point (continuous energy) Monte Carlo calculations have been performed on two types of benchmark problems, and the results have been compared to investigate the accuracy of the multi-group transport calculations in D-T fusion neutronics. In the multi-group calculations, space dependent self-shielding factors were used. As the first benchmark problem, spheres of four materials were analyzed. The discrepancy in the total fluxes obtained by the multi-group and point calculations was small in the lithium sphere, but it was great in the regions composed of iron. Nuclear design calculations of two types of fusion reactor blankets were performed as the second benchmark problem. The tritium breeding ratios obtained by the multi-group calculations agreed well with those obtained by the point calculations. However, the total fluxes in the reflectors or in the shields and the fast neutron leakage fluxes from the inboard shield were greatly underestimated by the multi-group calculations.