Research Article
Neutronic Calculation of Mixed Oxide Fuel for Gas-Cooled Fast Reactor using Monte Carlo code OpenMC
@INPROCEEDINGS{10.4108/eai.3-11-2023.2347943, author={M Aldi Kurniawan and Menik Ariani and Fiber Monado and Akmal Johan}, title={Neutronic Calculation of Mixed Oxide Fuel for Gas-Cooled Fast Reactor using Monte Carlo code OpenMC}, proceedings={Proceedings of the 3rd Sriwijaya International Conference on Basic and Applied Sciences, SICBAS 2023, November 3, 2023, Palembang, Indonesia}, publisher={EAI}, proceedings_a={SICBAS}, year={2024}, month={8}, keywords={neutronic mixed oxide openmc infinite multiplication factor}, doi={10.4108/eai.3-11-2023.2347943} }- M Aldi Kurniawan
Menik Ariani
Fiber Monado
Akmal Johan
Year: 2024
Neutronic Calculation of Mixed Oxide Fuel for Gas-Cooled Fast Reactor using Monte Carlo code OpenMC
SICBAS
EAI
DOI: 10.4108/eai.3-11-2023.2347943
Abstract
A fundamental aspect of nuclear power is using the prepared nuclear fuel once and then dumping it as waste, most of it can be recycled, thus closing the fuel cycle. The current means of doing this is by separating and recycling the plutonium. This plutonium is then blended with natural uranium to form mixed oxide (MOX) fuel. Neutronics calculations for the Gas-Cooled Fast Reactor are performed by OpenMC - Monte Carlo neutron and photon transport code. The parameter observed is the infinite multiplication factor (kinf), where it has been concluded that increasing the plutonium percentage has an impact on increasing the value. Depletion was carried out on the MOX composition with the best results, namely 9%, 10%, and 11% based on the value kinf ≈ 1. The depletion results also gave good results as shown by the cell successfully reaching a critical condition with an effective multiplication factor ≈ 1.
