Investigation on Concrete Neutron Shielding Properties Filled by B4C, CdO, and BN Microparticles

Аннотация

In this study, compounds; which are highly absorbent for neutrons due to their high absorption cross-section, are used in different volume proportions within cubic geometry for concrete as the primary shielding substrate. For this purpose, these compounds were investigated using Monte Carlo simulation method using the MCNP code. For achieving different volume percentages, the radius of the absorbent balls must vary from 200 µm(1%) to 750 µm(52%). The designed sample was exposure to an Am—Be neutron source with 5 Ci intensity. 37.5% of total activity of this source is Gamma emission. The goal of designing the geometry in this form is increasing the probability of interactions of fast and energetic neutrons which are moving along diffusion path inside the shield to decrease the speed and energy of these neutrons by inelastic scattering. The compounds have been investigated CdO, B₄C, and BN; which are filled in serpentine concrete. According to the simulation results, based on the efficiency of various compounds and smooth production of these compounds and also the necessary costs for their production and acceptable descending process of equivalent dose, boron carbide and boron nitride compound exhibited satisfactory performance in reducing equivalent doses. so using these compounds as absorbent balls seems to be a reasonable choice.