Comparative validation of Monte Carlo codes for the conversion of a research reactor

Abstract This paper presents the calculation results of the set of test problems for a research reactor with a tube-type low enriched uranium (LEU, 19.7 w/o, U-9%Mo) fuel and oxide high enriched uranium (HEU, 90 w/o) fuel, a light water moderator, and a beryllium reflector. The static cases and the depletion problem were examined. Calculations were performed using continuous energy Monte Carlo codes: MCNP (+MCREB for burnup calculation), MCU-PTR, and SERPENT 2. The impact of the cross-section libraries used for a particular problem on the calculated results was investigated.