Summary
Using low enrichment uranium as driver fuel under once‐through mode in molten salt reactor (MSR) attracts more and more attention because of its fuel availability, no new technology, and nuclear nonproliferation. It is regarded as a wise research and development road to shorten deployment time of MSRs and to prepare techniques and experiences for thorium‐uranium breeding of MSRs in the future. However, this fuel management is still faced with some different technical routes, such as the selection of carrier salts, the enrichment of uranium, with or without thorium, and the recycle necessary of spent nuclear fuel. Therefore, various fuel cycle schemes were compared and analyzed using an in‐house developed fuel management code MOBAT. Different graphite assemblies were optimized by changing the salt volume fraction in core and dimension to find a region for best fuel utilization and negative temperature reactivity coefficient. Prismatic block with 10% volume fraction of molten salt is considered as a good assembly type because of its significant space shielding effect of U‐238. For carrier salts, LiF‐BeF2 with 99.995% enriched Li‐7 displays higher fuel utilization and lower cost of fuel cycle than NaF‐BeF2, while the tritium production at the beginning of life will be two orders of magnitude higher. For fuel enrichment, 20% enriched uranium is recommended because the background of neutron absorption from carrier salt and graphite will be more significant in lower enrichment condition. Importantly, it shows that thorium is a good breed and burned fuel in situ and could improve the fuel utilization by 20%. Also, offline reprocessing to recover the uranium is a commendable scheme when the cost of offline reprocessing is lower than 400 $/kgHN.
The 235,238UF4 was irradiated by photo-neutrons, distribution and behavior of the fission product 95Nb from irradiated 235,238UF4 in FLiBe salt were investigated by the measurement of its activity in the salt with the γ-ray spectroscopy. The experiments indicated that a part of 95Nb deposited on the surfaces of graphite and Hastelloy, as the moderator and the structural materials of molten salt reactor (MSR), respectively, and the majority of 95Nb maintained in molten salt. Addition of lithium metal made 95Nb in salt to be reduced and settled, leading to the decrease in its activity. Degree of the decrease was found to be correlated with niobium concentration. The experimental results supported the statement proposed early by ORNL, that 95Nb might be used as a redox indicator for MSR. Finally, the problem met with on-site monitoring for redox potential in MSR was pointed, and a possible protocol to resolve the problem was proposed.
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