Uranium Minerals

Kadapa district of Andhra Pradesh

Oxidative pressure leaching of uranium from a dolomitic limestone ore:

India has a medium-tonnage, low-grade uranium ore deposit of siliceous dolomitic phosphatic limestone type, in Kadapa district of Andhra Pradesh. Detailed exploration carried out in a stretch of about 9 km in this area, established a resource of 29000 t of U3 O8 with a cut-off grade of 0.025% U3 O8. Mineralogical studies on an exploratory mine ore sample from this area, indicated the occurrence of uranium values predominantly as ultra-fine dissemination, in lighter gangue minerals (specific gravity less than 3.2). It also occurs, albeit to a minor extent, in the form of ultra-fine pitchblende in association with pyrite, as disseminations in collophane-rich parts, coffinite and as U-Ti complex. Carbonate minerals constitute the major gangue present in the form of dolostone (~80%). Siliceous minerals in the ore are quartz, feldspar and chlorite (13%). Collophane (4%) is the only phosphate bearing phase. Pyrite is the predominant sulphide ore mineral, along with few grains of chalcopyrite and galena. The iron bearing oxides are magnetite, ilmenite and goethite. Heavy media separation of various closely-sized feed fractions, using bromoform (BR) and methylene iodide (MI) liquids, have indicated that about 91% of the uranium values are present in lighter minerals (specific gravity <3.2), as ultra-fine disseminations. The remaining 9% of uranium values reported in methylene iodide heavy fraction, are accounted by discrete pitchblende, which is mostly associated with pyrite and collophane. Pitchblende occurring with pyrite is present as fine orbicular cluster, separated by thin disconnected rims of pyrite or as garlands around pyrite.

Leaching Chemistry of Uranium Minerals

The common oxidation states of uranium, in its minerals like uraninite, pitchblende, coffinite and numerous others, are +4 and +6. Amongst the two oxidation states, U+6 is soluble in aqueous media under suitable EH – pH conditions, while U+4 is practically insoluble. The uranium minerals occurring in various ore deposits consist predominantly of uranous ion (U+4), necessitating the use of an oxidant and other lixiviants, for quantitative dissolution during leaching. The type of leaching - acid or alkaline mode depends upon the host rock. Sulfuric acid is the common leachant in acid leaching process, while Na2 CO3 - NaHCO3 , (NH4 ) 2CO3 and NH4 HCO3 are the widely used lixiviants in alkaline leaching of uranium ores. The oxidant reagents could be either chemical or gaseous in nature. A typical chemical reaction in alkaline leaching of UO2 with carbonate ions and oxidant (X) is given in Equation 1, a similar equation can be written for the sulfuric acid leaching process. UO2 + 3CO3 -2 + X  [UO2(CO3) 3 ] -4 + X-2 .

Atmospheric alkaline leaching studies, carried out on this ore sample, by varying important process parameters like mesh-of-grind, temperature, contact time, dosages of leachants - sodium carbonate and sodium bicarbonate, solids concentration and type of oxidant, gave a maximum U3O8 leachability of 65%. Studies with other oxidants like NaOCl, Cu-NH3, oxygen and air gave poor leachability in comparison to KMnO4, emphasizing the need for strong oxidizing conditions during the dissolution process. However, as KMnO4 cannot be used as an oxidant on commercial scale due to its expensive nature, the only alternative is to carryout the leaching reaction in a pressure reactor, using a gaseous oxidant. Since the solubility of oxygen diminishes with increasing temperature, adoption of higher partial pressure aids in increased dissolved oxygen concentration. Detailed analysis of the leach residue obtained in the atmospheric leaching experiments indicated, that uranium values associated with pyrite are not completely leached at temperatures <1000 C. Further, some of the locked-up uranium values in various gangue phases, require more aggressive diffusion conditions for penetration of the leachant to the desired mineral interface. Both these requirements can be realized only at elevated temperature and under sustained oxidizing conditions, possible in an autoclave reactor. Leaching at elevated temperature and pressure was initially carried out in a laboratory, 5 liter S.S. autoclave reactor equipped with necessary instrumentation and control to maintain preset temperature, overpressure and agitation speed of the impeller. All the autoclave leaching experiments were carried out, at optimum dosage combination of sodium carbonate and sodium bicarbonate evolved in atmospheric leaching, that is - 50 kg/ton and - 70 kg/ton respectively. The autoclave leaching studies mainly addressed the dissolution of uranium associated with pyrite and the scope of replacing KMnO4 with industrial oxygen. Figs. 3 and 4, illustrate the effect of temperature and contact time on the leachability of uranium values, observed under aggressive conditions. About 75% of uranium values were leached at a reaction temperature of 125 - 130°C in 3 h of contact time, using a feed ground to 65% weight finer than 200#. Increasing the fineness of grind in -200# to 85% showed, an enhancement in leachability to about 80%. Based on these results, large-scale leaching studies were carried out, both on batch and continuous leach reactor, to generate necessary scale-up and engineering data for industrial scale reactor, besides verifying the reproducibility of results at higher-scale of operation. Both the batch and cigar type continuous reactor were of 850 liter capacity with inconel 600 as material of construction. Largescale studies confirmed the results generated in batch scale experiments. At present, DAE is setting-up a 3000 tpd capacity uranium mill at site, wherein two 720 m3 capacity autoclave reactors with inconel 600 cladding for the wetted parts will be used. This will be the first uranium plant, using autoclave leaching technology in India.

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