Neutronic Analysis On Molten Salt Reactor Fuji-12 Using 235U as Fissile Material In Lif-BEF2-UF4 Fuel

Abstract

Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiFBeF2-UF4 has been carried out. The problem faced in the use of thorium-based fuel is that the amount of 233U is small and not available in nature. 233U was produced through the 232Th breeding at a cost of $46 million/kg. That is a very high price when compared to 235U enrichment, which is only $100/kg. The MSR FUJI-12 used in this study is a generation IV reactor with a mixture of liquid salt fuel LiF-BeF2-ThF4-UF4 and thorium-based fuel ( 232Th+233U). In this study, neutronic analysis was carried out by replacing thorium-based fuel with uranium-based fuel ( 235U+238U). Neutronic analysis was performed using the OpenMC 0.13.0 code, which is a Monte Carlo simulation-based neutron analysis code. The nuclear data library used for neutronic calculations is ENDF B-VII/1. The fuel is used in a LiF-BeF2-UF4 molten salt mixture with three eutectic compositions: fuel 1, fuel 2, and fuel 3. Each fuel composition is optimized by enriching 235U in UF4 by 3 % to 8 %. The optimization results show the stability of the reactor criticality value, which is the main parameter so that the reactor can operate for the specified time. The optimization results show that fuel 1 cannot reach its optimal state in each variation of 235U enrichment. Fuel 2 and fuel 3 can reach optimal conditions at a minimum enrichment of 8 % and 7 % 235U. The results of the analysis of the distribution of the neutron flux in the reactor core show the distribution of nuclear reactions that occur in the core. The distribution of flux values in fuel 1 shows that the fission chain reaction is not running perfectly. Fuel 2 and fuel 3 are more stable by maintaining maximum flux at the center of the reactor core