Abstract: A process for separating a feed mixture of zirconium and petroleum coke containing traces amount of radioative materials by flotation process utilizing a plurality of flotation cells. The process comprises grinding the feed mixture, slurrying the ground feed mixture with water, treating the slurry with a flotation agent and a collector for the coke and subjecting the treated slurry to air sparging and agitation to create an overflow and an underflow. The overflow is then filtered to collect substantially zircon-free coke for further processing.

Abstract: A method of obtaining uranium metal from an oxidized uranium compound, characterized in that the oxidized compound is treated with chlorine and carbon at a first stage, to obtain a chloride which is reduced by electrolysis or metallothermy using a reducing metal at a second stage.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein.

Abstract: A method of separating uranium and plutonium from a mixed solution containing uranium nitrate and plutonium nitrate comprises cooling the mixed solution to a temperature ranging from -40.degree. to -20.degree. C. to thereby selectively precipitate uranyl nitrate. The precipitated uranyl nitrate is separated from the solution while leaving plutonium nitrate to remain in the solution.

Abstract: A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride.

Abstract: A method for separating plutonium from uranium and from fission products with the aid of crown ether compounds comprising contacting an aqueous solution A0 containing Pu, U and fission products with an organic solvent O0 containing at least one crown ether compound to obtain an organic solution O1 containing Pu and U; extracting U from the organic solution O1 with an aqueous solution A4 such as water or nitric acid to obtain an aqueous solution A5 containing U and an organic solution O3 containing mainly of Pu and recovering Pu from the organic solution using an aqueous solution A6 such as sulfuric acid.

Abstract: A process for controlling the oxidation reaction of oxides of uranium and fixing the ratio of oxygen to uranium in uranium oxide compounds by means of a passification process, and the stabilized uranium oxide compounds produced therefrom. The method is especially useful in the production of uranium oxide fuel for nuclear reactors.

Abstract: Actinides metals are recovered from spent nuclear fuel oxides containing fission products by a pyrochemical process. The process comprises, in part, electrorefining the metal complex from an anode by electrolytically oxidizing actinides into a salt and then electrodepositing actinides onto a cathode to form an actinide metal deposit. The actinide metal deposit is then melted to separate the salts and the actinide metals. The separated salt is recycled into an electrorefiner and the actinide metals are recovered and then transferred to a fuel fabrication system.

Abstract: A process for reducing an uranyl salt solution to an uranous salt solution employing a gas diffusion membrane comprising a reaction layer and a gas supply layer. The uranyl salt solution can be reduced to the corresponding uranous salt solution in the reaction layer of the membrane by a hydrogen gas supplied from the gas supply layer with the reduced power consumption.

Abstract: The process of the present invention provides a gas steam of a mixture of UF.sub.6 isotopes and an inert gas, e.g. nitrogen, which is adiabatically expanded through a nozzle into a laser light excitation zone and photodissociated to form U-235 enriched UF.sub.5.After the gas stream has passed through the laser light excitation zone, XeF.sub.6 is fed into the process gas stream so the the xenon hexafluoride remains protected against the dissociating radiation. The XeF.sub.6 may be mixed with the same inert gas that is employed for the adiabatic cooling of the UF.sub.6. The XeF.sub.6 reaction with U-235 enriched UF.sub.5 produces a stable complex of UXeF.sub.11, which polymerizes to poly(pentafluoroxenonium(+1)-hexafluorouranate V), an intermediate product of the present process. The intermediate product may be thermally decomposed to form U-235 enriched UF.sub.6 or U-235 enriched .beta.-UF.sub.5.

Abstract: A process for converting UF.sub.6 gas into UO.sub.2 powder comprising blowing UF.sub.6 gas and steam into a fluid bed to produce UO.sub.2 F.sub.2 particle, hydrating and dehydrating the UO.sub.2 F.sub.2 particle to UO.sub.2 F.sub.2 powder, and defluorinating and reducing the UO.sub.2 F.sub.2 powder to UO.sub.2 powder. The UO.sub.2 powder is suitable for manufacturing a reactor fuel owing to its high activity, low remaining fluorine amount and excellent fluidity.

Abstract: A process for conversion of gaseous UF.sub.6 to UO.sub.2 powders by using a fluidized bed reaction apparatus comprising pyrohydrolizing gaseous UF.sub.6 and steam to obtain UO.sub.2 F.sub.2 particles, hydrating and dehydrating the UO.sub.2 F.sub.2 particles to UO.sub.2 F.sub.2 anhydride and reducing the UO.sub.2 F.sub.2 anhydride to UO.sub.2 powders. The obtained UO.sub.2 powders are suitable for production of nuclear fuels in power plant owing to its good ceramic properties, low fluorine contents and free flowability.

Abstract: For the recovery of uranium from a solution in which the uranium is in valence (VI), a solution is formed containing ammonium (NH.sub.4.sup.+), fluoride (F.sup.-) and sodium hydrosulfite (Na.sub.2 S.sub.2 O.sub.4). This forms a precipitate (NaNH.sub.4 UF.sub.6), a previously unreported compound. It can be dissolved in an aqueous solution of nitric acid, aluminum nitrate nonahydrate, Al(NO.sub.3).sub.3.9H.sub.2 O, or a mixture of the two. This produces uranyl nitrate hexahydrate, which can be purified by processes conventionally used for that purpose.

Abstract: A process for manufacturing uranium oxide powder from UF.sub.6 which comprises converting UF.sub.6 to UO.sub.2 F.sub.2 by its reaction with excess alcohol in gas phase and further converting the formed UO.sub.2 F.sub.2 to uranium oxide by combusting hydrocarbon formed in the gas phase reaction and the excessive part of alcohol with oxygen containing gas supplied separately and supplying a regulated amount of steam separately to the combustion reaction zone.

Abstract: In the process for treating irradiated nuclear fuel to effect separation of uranium plutonium other higher actinides, and fission products, in which nitric acid treatment, followed by solvent extraction, then backwashing the reduction of tetra- and hexa-valent plutonium to the tri-valent form, then a second solvent extraction by which the tri-valent plutonium remains in the aqueous phase while uranium goes into the solvent phase, the reduction step is performed by hydrazine with or without tetra-valent uranium nitrate and catalyzed by technetium in the tetra-valent form with or without technetium in one or more higher valency states. The technetium can be present in the system as an irradiation product or be added to the process stream in a combined form.

Abstract: An improved process for producing powdered uranium dioxide from a solution of uranyl nitrate which is suitable for the manufacture of fuel for nuclear reactors. The process is continuous and comprises an incremental precipitation of soluble uranyl nitrate with ammonium hydroxide which is interrupted with an intermediate aging period. The precipitate of ammonium uranate solids is dried and thermally converted to a powdered oxide of uranium.

Abstract: Uranium dioxide powder produced by a gas phase process in which uranium hexafluoride is reacted with dry steam and then with steam and/or hydrogen at a higher temperature is subjected to mechanical treatment, e.g. milling, to break down its structure and increase its packing density. Other powders may be included with the uranium dioxide. The treated powder is mixed with a limited quantity (e.g. 0.5% by weight) of binder, preferably a high strength adhesive, to produce a free flowing powder and formed into pellets by pressing. The pellets are then sintered. Optionally the free flowing powder is spheroidised by tumbling prior to pressing into pellets.

Abstract: Hydroxymethane diphosphonic acid and alkali metal or ammonium salt of such acid are prepared. They are useful in detergent compositions and in sequestering and chelating polyvalent metals.

Abstract: A method of dissolving impure uranium tetrafluoride in a hot state in a nitric acid solution in the presence of an aluminum compound. For the purpose of obtaining a uranyl nitrate solution which can easily be separated from the solid phase formed during treatment, the dissolving is carried out in two stages at an appropriate temperature. The first stage comprises introducing quantities of nitric acid and of the aluminum compound which are insufficient to dissolve the impure uranium tetrafluoride completely, and keeping the resultant suspension agitated for a period of at least 0.5 hour. The second stage comprises introducing quantities of nitric acid and of the aluminum compound which are at least sufficient to dissolve the uranium not dissolved in the first stage, while keeping the suspension agitated.

Abstract: A chemical process for isotopic enrichment of uranium involving the steps of exciting a chelated uranium compound (e.g. hydrated uranyl acetate complex in aqueous solution) to an excited electronic state wherein the excited state preferentially reacts at different rates by virtue of dissimilar nuclear magnetic moment contributions to the chemical kinetics of alternative excited state reaction pathways (e.g. return to ground state by intersystem crossing by electron-nucleus hyperfine coupling vs free radical formation and subsequent precipitation of the hydrated basic salt of uranyl acetate).

Abstract: A physico-chemical cleaning process for the inner walls of a reactor which serves to maintain these walls in a state close to their main initial characteristics necessitated by a fluorination reaction, said characteristics being degraded by the deposition of a parasitic phase during the reaction, said initial characteristics being maintained by the use of a protective agent belonging to the group constituted by at least one of the reagents, a product resulting from the reaction, a product foreign to the reaction, but compatible with the substances of the main reaction or by the reaction of a third substance with at least one of the reaction substances or a mixture thereof, forming a renewable protective film on the said walls. The protective agent can be deposited by condensation on the walls before, during or after the main reaction, and then vaporized. The protective agent can be in liquid form, and trickling along the walls.

Abstract: A recovery process of uranium comprising:(1) extracting uranium ions with an organic solvent containing one or more compounds selected from the group consisting of alkyl phosphoric acid, alkyl-aryl phosphoric acid, alkyl dithio phosphoric acid, aryl dithio phosphoric acid, neutral phosphoric acid ester and alkyl amine together with a petroleum hydrocarbon as a diluent; and(2) stripping the uranium ions in the resultant organic solvent from the step (1) to an aqueous phase with contact of an aqueous solution containing one or more compounds selected from the groups of NH.sub.4 F, NH.sub.4 HF.sub.2, KF or KHF.sub.2.

Abstract: A novel easily handled substantially particulate ammonium uranate with a mean diameter between 30 and 150 micrometers, an apparent untamped bulk density of 2 to 2.8 g/cm.sup.3, and a flowability measured on the Carr scale equal to or greater than 95, with a low sulfate ion content between 0.5 and 1%, together with a fluidized bed process for preparing such ammonium uranate by precipitation of a super-saturated solution of ammonium uranate obtained by reacting an uranium sulfate solution and an ammoniacal solution, operating at a pH of about 6.6 to 7.2.

Abstract: Uranium is recovered from a carbonate leach solution containing a dissolved uranium salt and a monovalent ion. The pH of the leach solution is adjusted to about 5 to about 7.5, and preferably to about 6 to about 7. Phosphate ion is then added to typical in-situ leach solutions in an amount from about 10 to about 30 mole % in excess of the amount needed to stoichiometrically react with the uranium in said solution. This results in the precipitation of a compound made up of the monovalent ion, uranium, and the phosphate ion, which is insoluble in the solution. The precipitate is then separated from the solution preferably by means of a centrifuge or a vortex clarifier. It can then be dissolved in acid, and the uranium extracted into an organic solvent such as DEHPA-TOPA in kerosene.

Abstract: A method of recovering uranium dioxide from a sodium uranyl carbonate solution obtained by the alkaline carbonate leaching of uranium ore in which a solution is reacted at a temperature above 130.degree. C. and at superatmospheric pressure with particular metallic iron. The precipitated UO.sub.2 is recovered from the solution.

Abstract: Salts of the formula NF.sub.4.sup.+ MF.sub.7.sup.- are produced by the fowing reactionNF.sub.4 HF.sub.2 nHF+MF.sub.6 .fwdarw.NF.sub.4 MF.sub.7 +(n+1)HFwherein M is uranium (U) or tungsten (W).

Abstract: Metallic phosphates are prepared by heating mixtures of BPO.sub.4 and a metallic oxide or salt.

Abstract: The invention is a novel method for the recovery of uranium from dry, particulate uranium tetrafluoride. In one aspect, the invention comprises reacting particulate uranium tetrafluoride and calcium oxide in the presence of gaseous oxygen to effect formation of the corresponding alkaline earth metal uranate and alkaline earth metal fluoride. The product uranate is highly soluble in various acidic solutions wherein the product fluoride is virtually insoluble therein. The product mixture of uranate and alkaline earth metal fluoride is contacted with a suitable acid to provide a uranium-containing solution, from which the uranium is recovered. The invention can achieve quantitative recovery of uranium in highly pure form.

Abstract: A continuous process is disclosed for precipitating uranium from an aqueous solution formed by the hydrolysis of uranium hexafluoride gas. Undersized ammonium diuranate particles are formed in a first container by mixing the aqueous solution and an ammonium hydroxide solution containing about 10 to about 30% by weight ammonium. The ratio of the two solutions should be such that there are about 20 to about 30 moles of ammonium per mole of uranium. The temperature is maintained at about 30.degree. to about 50.degree. C. and the residence time in the first container is about 14 to about 57 seconds. The slurry is then transported to a second container for further particle growth where it is agitated at a temperature of about 20.degree. to 40.degree. C. for a residence time of about 2 to about 9 minutes. The resulting ammonium diuranate precipitate has a surface area of about 10 to about 20 meters squared per gram. The precipitate can then be calcined and pressed into pellets.

Abstract: In a process for the conversion of uranium hexafluoride to an uranium oxide by injecting uranium hexafluoride and dry steam into a first region of a vessel so as to form a plume of particles of uranyl fluoride and reacting the uranyl fluoride in a second region of the vessel with a countercurrent flow of steam and/or hydrogen the operation of the process is such that a major proportion of the uranyl fluoride is caused to circulate within the first region so that the original uranyl fluoride particles are able to grow and agglomerate in a dendritic manner.

Abstract: A nuclear fuel material green body of density from about 30 to 70% of theoretical density having tensile strength and plasticity adequate to maintain the integrity of the body during processing leading to ultimate sintered condition is produced by adding one or more amines to a particulate mass of the nuclear fuel containing about five percent of ammonium uranyl carbonate under conditions resulting in reaction of the amine with the ammonium uranyl carbonate, liberation of ammonia and formation of a water-soluble uranyl compound more effective as a binder than the ammonium uranyl carbonate.

Abstract: Sodium uranate which is easy to handle, characterized by the spherical form of the grains, the average diameter thereof, between 40 and 150 .mu.m, their uncompacted apparent bulk density of 2.5 to 3.5 g/cm.sup.3 and a flow value higher than or equal to 95, measured on the CARR scale produced by diluting the sodium uranate-containing solution introduced into a fluid bed reactor to between about 0.5 and 5 g/l within a period of less than about 2 seconds before the uranium-containing solution enters the reactor.

Abstract: Uranium values are recovered as uranyl peroxide from wet process phosphoric acid by a solvent extraction-precipitation process. The preferred form of this process comprises a first solvent extraction with DEPA-TOPO followed by reductive stripping of the extractant with Fe.sup.++ - containing phosphoric acid. After reoxidation, the uranium-containing aqueous stripping solution is extracted again with DEPA-TOPO and the pregnant organic is then stripped with a dilute ammonium carbonate solution. The resulting ammonium uranyl tricarbonate solution is then acidified, with special kerosene treatment to prevent wax formation, and the acidified solution is reacted with H.sub.2 O.sub.2 to precipitate a uranyl peroxide compound.

Abstract: Process waste waters at a pH of about 7 contaminated with radioactive isotopes are decontaminated by (a) adjusting the pH to about 5.8, (b) adding CaO or Ca(OH).sub.2 to raise the pH to about 8.5, (c) agitating the mixture for at least 5 minutes to effect intimate contact and produce a suspension of solids containing radioactive contaminants, and (d) separating the suspension of solids from the water by centrifuging. Removal of radioactive uranium isotopes with an alpha emission is effected at a pH of about 10. The process provides a method for concentrating radioactive contaminants in water for subsequent ultimate storage and also purifies the contaminated water so it may be safe to discharge it into the sewer. The treatment may be carried out in a plurality of stages in series.

Abstract: In the production of UO.sub.2, ammonium uranyl carbonate is an intermediate product wet with water and contaminated with ammonium carbonate and is washed with methanol to remove water and ammonium carbonate. The spent methanol containing 50% water and up to 10% ammonium carbonate is subjected to rectification in a column under subatmospheric pressure with cooling the top of the tower to a low temperature to retard decomposition of ammonium carbonate and condense a liquid water fraction. Clogging of the column, vapor lines and condenser by recombination of the decomposition products is prevented. The purified methanol contains less than 5% water and may be returned for further washing of ammonium uranyl carbonate.

Abstract: A method of preparing uranium (VI) peroxide hydrate from uranium tetrafluoride hydrate, comprising the steps of digesting uranium tetrafluoride hydrate in an aqueous acid in the presence of a fluoride complexing agent to produce an aqueous uranium solution, adjusting the aqueous uranium solution to a pH between about 1 to about 3, filtering the aqueous uranium solution to remove undissolved material, reacting the aqueous uranium solution with peroxide to precipitate uranium (VI) peroxide hydrate, and separating the precipitated uranium (VI) peroxide hydrate.

Abstract: A process is disclosed for improving the quality of an ammonium diuranate (ADU) precipitate. Ammonium hydroxide is added to a solution of uranyl fluoride in the presence of a polymer such as polyacrylic acid, polyacrylonitrile, or polyacrylamide. The presence of the polymer reduces the particle size of the precipitate and increases its settling rate. A reduced particle size provides an ADU powder which is suitable for nuclear fuel fabrication and an increased settling rate enhances the dewatering operation of the ADU slurry.

Abstract: A method of preparing uranium (VI) peroxide hydrate from uranium tetrafluoride hydrate, comprising the steps of digesting uranium tetrafluoride hydrate in an aqueous acid in the presence of a fluoride precipitating agent to produce an aqueous uranium solution, filtering the aqueous uranium solution to remove precipitated fluorides and undissolved material, adjusting the aqueous uranium solution to a pH between about 1 to about 3, reacting the aqueous uranium solution with peroxide to precipitate uranium (VI) peroxide hydrate, and separating the precipitated uranium (VI) peroxide hydrate.

Abstract: An actinide dioxide, e.g. uranium dioxide, plutonium dioxide, neptunium dioxide, etc., is prepared by reacting the actinide nitrate hexahydrate with sodium dithionite as a first step; the reaction product from this first step is a novel composition of matter comprising the actinide sulfite tetrahydrate. The reaction product resulting from this first step is then converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g. nitrogen) to a temperature of about 500.degree. to about 950.degree. C. for about 15 to about 135 minutes. If the reaction product resulting from the first step is, prior to carrying out the second heating step, exposed to an oxygen-containing atmosphere such as air, the resultant product is a novel composition of matter comprising the actinide oxysulfite tetrahydrate which can also be readily converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g. nitrogen) at a temperature of about 400.degree.

Abstract: A process for recovering uranium from a wet-process phosphoric acid crude solution is provided in which the phosphoric acid crude solution is contacted with an organic extractant consisting of octylphenyl phosphoric acid, di(2-ethylhexyl)phosphoric acid and trioctylphosphine oxide dissolved in an organic diluent to extract uranium from the phosphoric acid crude solution. The thus uranium loaded organic extractant is then contacted with mixed acid consisting of hydrofluoric acid and sulfuric acid, or alternatively with concentrated phosphoric acid to back-extract the uranium from the organic extractant.

Abstract: Radiation-contaminated ammonium nitrate is heated in solution to about 100.degree. C. in the presence of finely powdered calcium oxide or lithium hydroxide. Ammonia and water vapor are given off leaving an alkaline or alkaline earth nitrate which can then be safely decomposed by calcination into a metal oxide and oxides of nitrogen. The metal oxide can be recycled in a continuation of the process. The oxides of nitrogen can be passed through water to produce nitric acid useable in dissolving oxides of fissionable materials and the ammonia may be used in aqueous solution to react with nitrates of nuclear fuel or breeder metals in the very process that produces the by-product ammonium nitrate. Thus, all by-products and reagents can be reconverted and recycled.

Abstract: The material H(UO.sub.2)XO.sub.4.nH.sub.2 O, where X is P, As or I(OH).sub.2, conducts protons and when pressed under steadily increasing pressure (subsequently gradually released) may find use as a component of a battery, fuel cell, electrochromic cell, water vapor pressure meter or the like.

Abstract: A method of resolving water-in-oil emulsions resulting from the organic solvent extraction of uranium from aqueous leach liquors which comprises treating said emulsions with at least 20 parts per million of a water-soluble acrylamide copolymer which contains from 5 - 50% by weight of a lower alkyl substituted tertiary aminoethyl methacrylate and quaternary ammonium salts thereof.

Abstract: A hydrolysis column, used to hydrolyze uranium hexafluoride gas with water in an ammonium diuranate conversion process, which includes a pipe having a water inlet, a connector inserted in the pipe intermediate its length, and a gas nozzle connected to the connector to feed uranium hexafluoride gas into the water. Since the uranium hexafluoride gas will freeze at 147.degree. F, the gas nozzle is heated by steam which flows through internal passageways, thus imparting sufficient heat to the nozzle which then acts as a heat sink to maintain the gas in a fluid state. The gas-water mixture is then discharged through the pipe outlet to the next step in the process.

Abstract: Compositions of matter are described comprising the intercalation of Lewis bases into the layer lattice structure of UO.sub.2 F.sub.2 or by formation of directed chemical bonds between an electron donor atom of the Lewis base and the uranium ions in UO.sub.2 F.sub.2. Thermal treatment of these compositions results in the release of the Lewis base unchanged and the recovery of the uranyl fluoride.

Abstract: A process for producing UO.sub.2 F.sub.2 from a soluble uranyl salt. The uranyl salt is combined with a soluble fluoride salt in a solvent to form a reaction solution. The solvent exhibits Lewis base characteristics. The reaction product is a crystalline solid which is separated from the reaction solution. The UO.sub.2 F.sub.2 may then be obtained from the crystalline solid.

Abstract: A process for increasing the average reaction rate of reduction of UO.sub.2 F.sub.2. The UO.sub.2 F.sub.2 is treated with an organic compound which interacts with the UO.sub.2 F.sub.2. The combination is decomposed to yield UO.sub.2 F.sub.2 in a kinetically reactive state.

Abstract: An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.