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My idea is to make a new form of enriched uranium fuel (or more accurately, a substitute for enriched uranium) that's made by mixing uranium-233 (transmuted from thorium in a breeder reactor) with a larger quantity of natural uranium. The uranium-233 is a substitute for the additional uranium-235 in traditional enriched uranium. This would most likely be done by countries that operate both thorium-cycle reactors and uranium-cycle reactors.

As far as I see, this would be advantageous over light-enriched uranium. This might be cheaper than uranium enrichment in the long-term. But the advantage I see in Uranium-233 is that it has a higher fission cross-section & fission/capture ratio than Uranium-235, especially at intermediate neutron energies. This may give better neutron economy for the Reduced-Moderation Water Reactor (in development), which is likely to have a broad range of neutron energies including intermediate spectrum. It also might allow reduced moderation for other reactor types, such as gas-cooled reactors. Fast reactors utilizing U-233 may require less fissile material, due to the higher fission cross-section of U-233 compared to U-235.

Surprisingly, I can't find any research papers on this idea. Would this nuclear fuel I'm proposing have any problems operating in a reactor designed for light-enriched uranium? Would it work for a reactor designed specifically for it? How would it behave differently? Would this fuel be more difficult to reprocess once spent than traditional uranium fuel?

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Yes, you could substitute fissile U235 with fissile U233. Given a U235 enrichment, you could find a consistent U233 enrichment that would give you the same reactivity.

It is a little more complicated because you need to look at the reactivity as a function of burnup, and find the average over the depletion time.

The same process of finding equivalent enrichments is also needed when you use MOX fuel, which contains plutonium. You need to find an equivlent enrichment of U235 and the plutonium isotopes.

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Pu is produced by irradiating 238 U in a reactor; 233 U is produced by irradiating Th in a reactor. The Pu or the 233 U are recovered by reprocessing the spent fuel; then either the Pu or the 233 U can be used to fuel a reactor. There is considerable more experience in reprocessing for Pu than for 233 U due to the nuclear weapons programs. Concerns over nuclear proliferation have curtailed wide-spread reprocessing of either.

Opposition to nuclear power has essentially stopped any rapid advancement in nuclear power technology. Current nuclear power plants in the US are the 1960 era designs, there is no commercial reprocessing in the US, and the US stopped development of fast breeder reactors decades ago with the abandonment of the Clinch River Breeder Reactor and the closure of the Fast Flux Test Facility. There is some work on modular, small, inherently safe reactor designs but none are in commercial use.

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