Neutronic Analysis on Molten Salt Reactor (MSR) Using OpenMC Code With Variations of Geometry Core Fueled By LiF-BeF2-UF4

Year 2024, Volume: 11 Issue: 2, 152 - 159, 07.07.2024

Ratna Dewi Syarifah , Briyanti Adelia Putri , Indarta Kuncoro Aji

https://doi.org/10.31202/ecjse.1364028

Abstract

Nuclear Power Plant can produce electricity more efficiently and have low carbon emissions. The nuclear reactor used in this study is the MSR (Molten Salt Reactor) FUJI-12. This study aims to conduct an analysis neutronics on MSR FUJI-12 by varying the geometry shape of the reactor core and finding the most effective geometry core to use in MSR. The material used in this study is a mixture of LiF-BeF2-UF4 molten salts. This study uses OpenMC code with nuclear data library ENDF/B VIII.1. The shapes of the geometry core that will be compared are, pancake, balance, and tall. The three geometry core shapes will then be varied into seven kinds. The results show that the geometry of the core is very influential on the reactivity of a nuclear reactor. The k_eff value for all geometry core variants at the beginning of the operating reactor is in a supercritical condition and it will be a critical or subcritical condition at the end of the reactor’s operating life. Balance and tall 1 variants have a high on distribution neutron flux and fission rate. The Balance variant also produces the smallest mass of plutonium nuclides. The neutronic analysis that has been carried out show that the balance variant is the optimal geometry core design that can be used on the MSR FUJI-12.

Keywords

Molten salt reactor, Monte carlo, OpenMC, Geometry core

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