Calculation of Neutron Fluxes and Radiation Doses for Neutron Irradiator 226 Ra-Be Using the MCNP5 Code

Abstract

The irradiator ²²⁶Ra-Be unit available at the Physics Department of the Sciences Faculty, Damascus University, was simulated by using the MCNP code. Calculations of neutron fluxes and radiation doses were carried out.

Simulation results showed that neutron fluxes, in the energy ranges: thermal (10^-9-10^-6 MeV), fast (0.11-12.0 MeV), and medium (10^-5-10^-1 MeV), had approximately the same magnitudes in some channels (See text for the definition of these channels) where flux percent was about (thermal 70.0%, fast 18.0 % and medium 12.0%). On the other hand, the flux percent in one particular channel was about (thermal 40.0%, fast 39.0% and medium 21.0%) and in another one (thermal %60.0, fast %26.0 and medium %14.0) with presence of a plate of Cadmium whose thickness is 2 mm.

Absorbed radiation doses, in two channels, were calculated by using MCNP5 code and then compared with those measured experimentally by using thermoluminescent dosimeters (TLD). A satisfactory agreement between calculated and measured results was found. The relative differences were about 3.8% and 7.2% in these two channels respectively.

Keywords: Neutron irradiator, ²²⁶Ra-Be, Thermoluminescent dosimeter, MCNP5 code.