Monte Carlo simulation of the ETRR-2 research reactor using the MCNP Code

Abstract

Abstract The MCNP computer Code is used to model the ETRR-2 research reactor. A computer program was designed to evaluate the axial burn-up of the fuel elements. The excess reactivity of the reactor core is calculated for different core configurations and compared with the existing measurements. The thermal flux is also calculated and compared with measurements. Several factors that affect the safety of the reactor such as power peak and the effect of control rod insertion on the reactor power and flux were studied and analysed. The agreement between the MCNP results and the experimentally determined values is good.