precipitated from a molten fluoride mixture by addition of thorium oxide, and that the precipitate can be returned to solution by treatment with HF. Experimental results also indicated that treatment of protactinium-containing salt with ZrO2 leads to oxide precipitation of the protactinium and that after beta decay of the protactinium, the resulting UO2 will react with ZrF4 to give UF4. More recent experimental results have indicated another method for removing protactinium directly from the blanket fluid. This involves treating the molten blanket salt with a stream of bismuth containing dissolved thorium metal. The thorium reduces the protactinium (and also any uranium) to metal, which can then be accumulated on a stainless-steel-wool filter, or recovered directly from the liquid metal. The metal can be hydro-fluorinated and/or fluorinated to return the protactinium (and any uranium) to the fuel-recycle process as the fluoride. Thus there is experimental evidence that direct processes are available for removal of protactinium from the blanket stream of molten-salt reactors. If protactinium is not removed directly from the blanket stream, then the blanket salt is processed by the fluoride-volatility process alone. Any uranium not removed during the blanket processing would be returned to the blanket and removed by subsequent processing.
5. GENERAL CONSIDERATIONS Only statements of a qualitative nature can be made relative to the economic and technical impact of processing thorium fuels because of the developmental nature of this fuel cycle. It is believed that reprocessing of thorium fuels is technically feasible in several ways; and that the reprocessing cost, in mills/kWh, should not be significantly different than for the U-Pu cycle, provided that the two cycles are compared on the same terms, i.e., the sizes of plant should be about the same, and if "equilibrium" fuel is used in one case, it should be used in the other also. Isotopic impurities are often mentioned as a liability of U-233, but the plutonium cycle will also accumulate them, and will require at least semi-remote fabrication (70). While Pu-239 emits only a small amount of hard radiation, its emission is the source of very high toxicity if ingested. While the plutonium fuel could be handled in a glovebox to guard against hazards, long irradiations and high burnups, on the other hand, produce substantial amounts of other isotopes with a consequent increase in the radiation level. The gloveboxes would then no longer be adequate and shielding would be required around the equipment and remote control techniques would be used in handling the fuel. The absolute amount of actinide activity present would be a function of the isotopic ratio and cooling time before and after processing. If thorium fuels are reprocessed in equipment designed for uranium, or with processes such as Thorex which use reagents developed for a different system, complications will occur and good economy cannot be expected. Processing of thorium fuel should be based on plants designed for that specific purpose and an intensive research program on new separation methods would be desirable and economically justified.