Simulation of neutron flux in silicon, cadmium and plumbum using Monte Carlo method

Abstract

The purpose of this study is to investigate the neutron energy flux from reactor after attenuated by silicon, cadmium and plumbum at the end of the beam port. Neutrons can be categorized into thermal neutron, intermediate and fast neutron according to their energies. Fast neutrons in reactor needs to be thermalized before it can be utilized for applications such as in medicine, industrial and research purposes. The fast neutron will lost its energy into thermal neutrons due to scattering and absorption process with nucleuses. The neutron probability of interaction with nucleus depends on the microscopic cross-section, which is different for each material. All the cross-section data for every element have been compiled in ENDF (Evaluated Nuclear Data File) format and internationally recognized. In this study, the neutron transport was simulated using Monte Carlo N-Particle Transport Code, Version 5 (MCNP5). The thickness of the materials used in this research is in the range of 1 cm to 10 cm. The result shows that the neutron was reduced significantly by silicon then follows by plumbum and cadmium. The thermal neutron flux was the lowest in cadmium because it have high thermal neutron cross-section.