Comparative Assessment of Molten Salt Reactor Neutronic Performance with Various U-233 Purity

Abstract

Molten salt reactor (MSR) can be deployed as a thermal breeder reactor in a thorium fuel cycle. The fissile nuclide mostly uses U-233, which is nonexistent in nature and must be synthetised. Researches on thermal breeder MSR usually assume that the U-233 is pure, but in technical reality, U-233 synthesis always accompanied by other uranium isotopes. These impurities can affect the reactor physics performance and altering the operational safety consideration. This research studies the impact of using impure U-233 on the neutronic performance Passive Compact Molten Salt Reactor (PCMSR). Four U-233 vectors with various level of purity were used as comparison. The investigated parameters were reactor criticality, temperature coefficient of reactivity (TCR), kinetic parameters, and conversion ratio (CR). The calculations were performed using MCNP6.2 code with ENDF/B-VII.0 nuclear data library. From the calculation, impure U-233 fuels were proved to improve the TCR as a result of weaker moderator temperature coefficient (MTC). Whilst impurity does not particularly affect delayed neutron fraction, it reduces neutron generation time. Impure U-233 vectors slightly altered CR value, but rather insignificant. Overall, operational safety and CR value can be maintained even if the MSR core is started using low-purity U-233.