Monte Carlo transport code use for optimisation of neutron flux produced with 10-18 MeV electron beam energy

Abstract

This manuscript describes how to calculate the neutron flux produced by irradiation of a heavy target with an electron beam energy from 10 MeV to 18 MeV. It also describes how to optimise the design of the photo-neutrons converter target. For this purpose, we investigate the optimal materials used as a neutron converter by performing several simulations with the MCNP-6 code using heavy materials (tantalum, lead, tungsten and tungsten-beryllium). We optimise the neutron flux by varying thickness and diameter of the target with different electron beam energy. The results show that the electron-based target was able to provide high-neutron flux with a small target geometry dimension. After all calculations, we concluded that the neutron flux produced using the optimal geometry found in this article, during the several simulations using the MCNP method knows a significant increase, and reaches the following values of 2.01 × 1011 n/cm²s, 2.28 × 1011 n/cm²s, 1.17 × 1010 n/cm²s and 1.51 × 1011 n/cm²s, respectively for lead, tantalum, tungsten and tungsten-beryllium targets.