Uranium extraction of its ore

Leaching

Roasted uranium ores are leached of their uranium values by both acidic and alkaline aqueous solutions. For the successful operation of all leaching systems, uranium must either be initially present in the more stable hexavalent state or be oxidized to that state in the leaching process.

Acid leaching is commonly performed by agitating an ore-leach mixture for 4 to as long as 48 hours at ambient temperature. Except in special circumstances, sulfuric acid is the leachant used; it is supplied in amounts sufficient to obtain a final leach liquor at about pH 1.5. Sulfuric acid leaching circuits commonly employ either manganese dioxide or chlorate ion to oxidize the tetravalent uranium ion (U4+) to the hexavalent uranyl ion (UO22+). Typically, about 5 kilograms (11 pounds) of manganese dioxide or 1.5 kilograms of sodium chlorate per ton suffice to oxidize tetravalent uranium. In any case, the oxidized uranium reacts with the sulfuric acid to form a uranyl sulfate complex anion, [UO2(SO4)3]4-.

Uranium ores that contain significant amounts of basic minerals such as calcite or dolomite are leached with 0.5 to 1 molar sodium carbonate solutions. Although a variety of reagents has been studied and tested, oxygen is the uranium oxidant of choice. Typically, candidate ores are leached in air at atmospheric pressureand at 75° to 80° C (167° to 175° F) for periods that vary with the particular ore. The alkaline leachant reacts with uranium to form a readily soluble uranyl carbonate complex ion, [UO2(CO3)3]4-.

Treatment of uranium leachates

The complex ions [UO2(CO3)3]4- and [UO2(SO4)3]4- can be sorbed from their respective leach solutions by ion-exchange resins. These special resins—characterized by their sorption and elution kinetics, particle size, stability, and hydraulic properties—can be used in a variety of processing equipment—e.g., fixed-bed, moving-bed, basket resin-in-pulp, and continuous resin-in-pulp. Conventionally, sodium and ammonium chloride or nitrate solutions are then used to elute the sorbed uranium from the exchange resins.

Uranium can also be removed from acidic ore leach-liquors through solvent extraction. In industrial methods, alkyl phosphoric acids—e.g., di(2-ethylhexyl) phosphoric acid—and secondary and tertiary alkyl amines are the usual solvents. As a general rule, solvent extraction is preferred over ion-exchange methods for acidic leachates containing more than one gram of uranium per litre. Solvent extraction is not useful for recovery of uranium from carbonate leach liquors, however.

Precipitation of yellow cake

Prior to final purification, uranium present in acidic solutions produced by the ion-exchange or solvent-extraction processes described above, as well as uranium dissolved in carbonate ore leach solutions, is typically precipitated as a polyuranate. From acidic solutions, uranium is precipitated by addition of neutralizers such as sodium hydroxide, magnesia, or (most commonly) aqueous ammonia. Uranium is usually precipitated as ammonium diuranate, (NH4)2U2O7. From alkaline solutions, uranium is most often precipitated by addition of sodium hydroxide, producing an insoluble sodium diuranate, Na2U2O7. It can also be precipitated by acidification (to remove carbon dioxide) and then neutralization (to remove the uranium) or by reduction to less soluble tetravalent uranium. In all cases, the final uranium precipitate, commonly referred to as yellow cake, is dried. In some cases—e.g., with ammonium diuranate—the yellow cake is ignited, driving off the ammonia and oxidizing the uranium to produce uranium trioxide (UO3) or the more complex triuranium octoxide (U3O8). In all cases, the final product is shipped to a central uranium-purification facility.

Refining of yellow cake

Uranium meeting nuclear-grade specifications is usually obtained from yellow cake through a tributyl phosphate solvent-extraction process. First, the yellow cake is dissolved in nitric acid to prepare a feed solution. Uranium is then selectively extracted from this acid feed by tributyl phosphate diluted with kerosene or some other suitable hydrocarbon mixture. Finally, uranium is stripped from the tributyl phosphate extract into acidified water to yield a highly purified uranyl nitrate, UO2(NO3)2.

Conversion and isotopic enrichment

Uranyl nitrate is produced by the ore-processing operations described above as well as by solvent extraction from irradiated nuclear reactor fuel (described below, see Conversion to plutonium). In either case, it is an excellent starting material for conversion to uranium metal or for eventual enrichment of the uranium-235 content. Both of these routes conventionally begin with calcining the nitrate to UO3 and then reducing the trioxide with hydrogen to uranium dioxide (UO2). Subsequent treatment of powdered UO2with gaseous hydrogen fluoride (HF) at 550° C (1,025° F) produces uranium tetrafluoride (UF4) and water vapour, as in the following reaction:

This hydrofluorination process is usually performed in a fluidized-bed reactor.

Uranium tetrafluoride can also be fluorinated at 350° C (660° F) with fluorine gas to volatile uranium hexafluoride (UF6), which is fractionally distilled to produce high-purity feedstock for isotopic enrichment. Any of several methods—gaseous diffusion, gas centrifugation, liquid thermal diffusion—can be employed to separate and concentrate the fissile uranium-235 isotope into several grades, from low-enrichment (2 to 3 percent uranium-235) to fully enriched (97 to 99 percent uranium-235). Low-enrichment uranium is typically used as fuel for light-water nuclear reactors.

After enrichment, UF6 is reacted in the gaseous state with water vapour to yield hydrated uranyl fluoride (UO2F2 · H2O). Hydrogen reduction of the uranyl fluoride produces powdered UO2, which can be used to prepare ceramic nuclear reactor fuel