Neutronic Calculation of Mixed Oxide Fuel for Gas Cooled Fast Reactor using Monte Carlo code OpenMC
Muhammad Aldi Kurniawan , Menik Ariani* , Fiber Monado

Physics Department, Faculty of Mathematics and Natural Sciences, Sriwijaya University, Jl. Raya Palembang - Prabumulih Km. 32 Indralaya, OI, South Sumatra 30662, Indonesia.
* menik_ariani[at]unsri.ac.id

Abstract

After its stay in the reactor for several years, the spent UO2 fuel contains an appreciable amount of fissile plutonium which could be separated from it. This Plutonium is then blended with natural uranium (0.7 wt% U-235) to form mixed oxide (MOX) fuel. In this research, the plutonium content could be varied from 0 wt% to 14 wt% keeping the safety parameters of the system in view. Neutronics calculations of MOX fuel for Gas Cooled Fast Reactor are performed by OpenMC - a community-developed Monte Carlo neutron and photon transport code. The physical parameter observed is the infinite multiplication factor value where it has been concluded that increasing the plutonium percentage has an impact on increasing the multiplication factor value.

Keywords: neutronic, mixed oxide, OpenMC

Topic: Physics and Applied Physics