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: The present invention concerns a continuous process for the oxidizing attack of a uraniferous ore by means of an aqueous recycling liquor which contains in a dissolved state alkaline carbonate and bicarbonate and uranium at a concentration close to the limit of solubility thereof at the temperature of the attack operation. The uranium which is precipitated in the course of the attack operation is collected within the solid phase which remains after attack. By redissolution in a suitable aqueous liquor and separation from the sterile materials, a uraniferous liquor is obtained.

Abstract: The level of radioactivity in gypsum produced from phosphate rock can be reduced by adding to the slurry of acid and phosphate rock in the conventional process for producing phosphoric acid a combination of concentrated nitric acid or hydrochloric acid and an ammonium salt, especially ammonium nitrate or ammonium phosphate, before the precipitation of the calcium sulfate.

Abstract: Disclosed is a method of removing uranium from a slurry containing a water-insoluble molybdate and a Group II element, usually calcium. The pH of the slurry is raised with an alkali metal hydroxide and maintained at a level sufficiently high to solubilize insoluble molybdenum compounds. This will generally require a pH of at least 10 which can be achieved by adding sodium hydroxide. Carbonate ion is then added to the slurry in an amount at least sufficient to precipitate the cation of the insoluble molybdenum compound. And finally, the solid portion of a slurry, which now contains calcium carbonate and sodium uranate, is separated from the liquid portion, which contains a soluble molybdenum compound.

Abstract: A process is described for recovering uranium from a pregnant lixiviant containing uranium values and a certain portion of molybdenum values comprising passing the pregnant lixiviant through an anion-exchange resin to capture the uranium and molybdenum values on the resin, eluting the uranium and molybdenum values from the resin with a salt solution, passing the eluate through a weak acid cationic resin in its hydrogen form to capture the uranium values on the resin and treating and resulting eluate to precipitate uranium therefrom to produce the familiar "yellow-cake.

Abstract: Precipitation of Fe.sub.3 HN.sub.4 H.sub.8 (PO.sub.4).sub.6 is prevented in the second cycle extractor, in a two cycle uranium recovery process, by washing ammonia laden organic solvent stream, from the second cycle stripper, with first cycle raffinate iron stream containing phosphoric acid, prior to passing the solvent stream into the second cycle extractor.

Abstract: A method for the heat treatment of a radioactive substance by microwave power comprising preparing a nitrate solution of uranium, thorium, plutonium or a mixture thereof, and applying to the nitrate solution microwave energy sufficient to directly convert the nitrate solution into an oxide powder of uranium, thorium, plutonium or a mixture thereof, respectively. Such oxide powder is suitable for the manufacture of nuclear fuel pellets.

Abstract: Process for the recovery of the uranium present in a phosphoric acid solution by bringing the said solution into contact with an organic solvent suitable for extracting the uranium, wherein the organic solvent comprises a system of extractants constituted by an acid organophosphorus compound and by a neutral phosphine oxide of formula: ##STR1## in which the radicals R.sub.1, R.sub.2 and R.sub.3 are identical or different alkyl, aryl or alkoxyalkyl radicals, at least one of the radicals R.sub.1, R.sub.2 and R.sub.3 being an alkoxyalkyl radical.

Abstract: A process for recovering uranium and/or thorium from a liquid containing uranium and/or thorium is disclosed, which comprises capturing the uranium and/or thorium in the liquid by an amorphous silica precipitate formed by adding water glass to the liquid, making the captured uranium and/or thorium eluted from the precipitate by acid-treatment, recovering the eluted uranium and/or thorium as an acidic solution, and regenerating the precipitate to water glass by use of an alkali metal hydroxide solution. Thus, the uranium and/or thorium can be recovered in high yield and the amorphous silica precipitate, that is, a formed radioactive solid waste can be remarkably reduced.

Abstract: A cyclic process for the treatment of a uraniferous liquor containing, in addition to the desired metal, sodium carbonate and/or bicarbonate and sodium sulphate, obtained by the alkaline attack of a sulphur-containing uraniferous ore, the said process comprising the neutralization of the uraniferous liquor then the fixation of the uranium and the SO.sub.4.sup.= ions over a strong anion-exchange resin, the elution thereof with an ammonium carbonate and/or bicarbonate solution in the form of an ammonium uranyl carbonate which is subsequently decomposed and ammonium sulphate, the precipitation and the calcination of the precipitated uranates and/or diuranates with regeneration of the eluent solution of ammonium carbonate and/or bicarbonate, the causticiation of the liquor freed from uranium which leaves the anion-exchange resin to eliminate the SO.sub.4.sup.= ions in calcium form, recycling a carbonated and/or bicarbonated liquor which is free from SO.sub.4.sup.

Abstract: A process for recovering uranium and/or thorium from a liquid containing uranium and/or thorium is disclosed, which comprises making a precipitate composed mainly of amorphous silica formed by adding water glass to the liquid containing uranium and/or thorium, capturing the uranium and/or thorium in the liquid by the precipitate, treating the precipitate with acid to elute the captured uranium and/or thorium, adding ammonia or hydrogen peroxide to an acidic solution containing the eluted uranium and/or thorium to form a precipitate composed of ammonium salt or peroxide of the uranium and/or thorium, and filtering out the ammonium salt or peroxide precipitate. Thus, high purity solid uranium and/or thorium reutilizable for manufacture of nuclear fuel material can be recovered in high yield.

Abstract: Alkaline solutions containing sulfate and sodium ions such as uranium leach liquors, are treated with cationic ion exchange resins in the ammonium form to form a solution containing soluble ammonium salts and complexes; the solution is further treated to convert the soluble sulfates into insoluble sulfates and, in the case of uranium, further treating to separate the uranium by heat and/or acidification. The ammonium can be recovered and reused as ammonium carbonate to regenerate the ion exchange resin whereby sodium ions are eluted from the resin in the form of sodium carbonate which is converted to sodium hydroxide and re-cycled for use as leaching solution.

Abstract: This present invention provides a process of recovering uranium from a uranium-containing acid lixiviation solution comprising extracting the lixiviation solution with an organic extractant comprising an organic diluent, a dialkyl pyrophosphoric acid, and a stabilization agent, and treating the mixture of solutions as follows:(a) bubbling anhydrous ammonia gas into the solution mixture to precipitate the ammonium salt of excess dialkyl pyrophosphoric acid; and(b) separating the precipitate from the solution and recovering the uranium from the solution.

Abstract: A method of controlling the molybdenum content of uranium yellowcake which is precipitated from an eluate of ammonium carbonate. The pH of the eluate is maintained at a relatively high value, i.e., never less than 9.3, while the eluate is heated to strip off ammonia, carbon dioxide, and water. Ammonia is added to the eluate during heating to keep the pH high and the molybdenum in solution in the eluate. Also, the precipitated uranium yellowcake may be treated by adding aqueous ammonia and heating to form an insoluble uranium salt and a soluble molybdenum salt which can be filtered to produce a uranium yellowcake having a reduced molybdenum content.

Abstract: Uranium is recovered from phosphate ore in a process comprising making phosphoric acid, partially reacting same with NH.sub.3, removing aluminum with dilute sodium carbonate and sodium bicarbonate and extracting the uranium with a concentrated solution of said carbonates.

Abstract: High purity ammonium uranyl tricarbonate is produced from crude plant yellow cake by treatment of the yellow cake with a mineral acid, neutralization of the acidic solution with a conversion of the uranium therein to ammonium uranyl tricarbonate, precipitation of the ammonium uranyl tricarbonate with a non-solvent, followed by separation. Optionally, the ammonium uranyl tricarbonate can be calcined to produce high purity uranium oxides.

Abstract: Process for the recovery of uranium from a pregnant lixiviant employed in uranium leaching operations in which the uranium is concentrated by ion exchange resin. The ion exchange resin is eluted with an eluant containing 0.05-0.5 molar ammonium carbonate. The eluant also contains a base such as ammonium hydroxide in an amount sufficient to increase the pH to a value within the range of 9.0-11.0, thus increasing the carbonate/bicarbonate ratio. This enables the use of a relatively small quantity of ammonium carbonate to provide the desired carbonate ion concentration during the elution step.

Abstract: The injection of an ammonium sulfite or bisulfite solution increases the permeability of an uranium ore bed that has suffered permeability losses during the in-situ mining of uranium with an alkaline leach solution containing a peroxide or dissolved oxygen oxidant. Such an injection recovers much of the lost formation permeability, thus decreasing costs and effort required to put needed restoration solutions or further leach solutions through the ore bed. In addition, uranium contamination of the ground water normally occurring after cessation of leaching is significantly lowered by such injection.

Abstract: Improvement in the process for recovering uranium from wet-process phosphoric acid solution derived from the acidulation of uraniferous phosphate ores by the use of two ion exchange liquid-liquid solvent extraction circuits in which in the first circuit (a) the uranium is reduced to the uranous form; (b) the uranous uranium is recovered by liquid-liquid solvent extraction using a mixture of mono- and di-(alkyl-phenyl) esters of orthophosphoric acid as the ion exchange agent; and (c) the uranium oxidatively stripped from the agent with phosphoric acid containing an oxidizing agent to convert uranous to uranyl ions, and in the second circuit (d) recovering the uranyl uranium from the strip solution by liquid-liquid solvent extraction using di(2-ethylhexyl)phosphoric acid in the presence of trioctylphosphine oxide as a synergist; (e) scrubbing the uranium loaded agent with water; (f) stripping the loaded agent with ammonium carbonate, and (g) calcining the formed ammonium uranyl carbonate to uranium oxide, the i

Abstract: The invention is an improvement in the process for recovering uranyl uranium from solutions in which the uranyl uranium is recovered with an liquid-liquid solvent extraction agent dissolved in an inert solvent, the loaded agent scrubbed with water, the scrubbed agent stripped with ammonium carbonate solution to form a slurry of ammonium uranyl tricarbonate, the stripped agent returned to the liquid-liquid solvent extraction step, and the ammonium uranyl tricarbonate calcined to a uranium oxide product, the improvement which comprises stripping the uranium from the ion exchange agent with an alkali metal carbonate rather than ammonium carbonate, scrubbing the loaded agent with acidified alkali carbonate strip solution prior to stripping, and regenerating the stripped liquid-liquid solvent extraction agent with a mineral acid before return to the liquid-liquid solvent extraction circuit.

Abstract: Uranium is recovered from alkaline leach liquors of uranium ores by treating the leach liquors with an ion exchange resin in the ammonium form to provide a solution of salts and complexes of uranium and ammonium from which the uranium is separated by treating with heat or an acid.

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: Uranium ore deposits which contain uranium in the relatively insoluble tetravalent state are readily selectively leached in situ to recover relatively pure uranium compounds, by: (a) passing through the ore deposit a relatively dilute aqueous leach solution of ammonium bicarbonate, ferric ammonium ethylenediaminetetraacetic acid (EDTA), and a source of oxygen, the leach solution converting the tetravalent uranium to hexavalent uranium which readily dissolves in the leach solution; (b) withdrawing the reacted leach solution enriched in dissolved uranium from the ore deposit; and (c) stripping the uranium from the withdrawn leach solution. The stripping of the uranium from the leach solution is preferably accomplished by countercurrent flow of the enriched leach solution to a column of base anion exchange material which preferentially extracts the uranium.

Abstract: A process for treating a particulate uranium oxide composition containing ammonium diuranate precursors such as uranyl fluoride and uranyl nitrate, along with possible impurities, is disclosed. The process comprises the steps of (a) contacting the composition with a sufficient amount of an ammonium hydroxide solution to convert the ammonium diuranate precursors to ammonium diuranate, and (b) heating the composition under a controlled atmosphere at a temperature sufficient to dry the composition, sublime any ammonium fluoride present, and convert the ammonium diuranate to uranium dioxide. In a first modification of the invention, there is practiced the additional step of moving the composition, preferably by mechanically agitating the composition, prior to the heating step, or during the heating step, or both prior to and during the heating step, for a sufficient time to produce rounded agglomerates in the form of a readily flowable powder.

Abstract: Molybdenum is separated from molybdenum-containing activated charcoal or char also containing small amounts of uranium obtained as a by-product in uranium leaching processes by stripping with an alkaline solution to provide a molybdenum-containing solution containing substantially less than 500 ppm U.sub.3 O.sub.8.

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: Process for the in-situ leaching of uranium from a subterranean ore deposit comprising introducing into the deposit an aqueous lixiviant having a pH of at least 6.0 and containing an alkali metal sulfate leaching agent. The alkali metal sulfate may be employed in combination with an alkali metal carbonate or bicarbonate with the sulfate comprising the predominant leaching agent. The lixiviant may be at a pH of at least 7.5 and contain an alkali metal sulfate leaching agent and a hypochlorite oxidizing agent.

Abstract: A method of recovering uranium from wet-process phosphoric acid wherein the acid is treated with a mixture of an ammonium salt or ammonia, a reducing agent, and then a miscible solvent. Solids are separated from the phosphoric acid liquid phase. The solid consists of a mixture of metal phosphates and uranium. It is washed free of adhering phosphoric acid with fresh miscible solvent. The solid is dried and dissolved in acid whereupon uranium is recovered from the solution. Miscible solvent and water are distilled away from the phosphoric acid. The distillate is rectified and water discarded. All miscible solvent is recovered for recycle.

Abstract: Uranium ore deposits which contain certain proportions of other metals and elemental components, such as are present in redox roll front ore deposits, are selectively leached in situ by passing therethrough relatively dilute aqueous leach solutions comprising essentially from about 0.5 to 5 grams per liter of ammonium bicarbonate and from about 0.1 to 3 grams per liter of peroxide, preferably introduced as aqueous H.sub.2 O.sub.2, and sufficient ammonia to bring the solution to a pH of from about 7.4 to 9, and preferably from 7.5 to 8.5, thereafter withdrawing from the ore deposit the aqueous leach solution enriched in uranium which it preferentially extracts along with a generally lower proportion of other metals and elements as compared to their respective ratios in the ore deposit, and contacting the enriched leach solution with a strong base anion exchange material to strip the uranium from the leach solution.

Abstract: Carbon dioxide is admixed with water to form a carbonic acid solution for use as a leaching solution for extraction of uranium by solution mining. An oxidizing agent is also present in the solution.

Abstract: The present invention relates to the recovery of uranium from carbonate leach solutions. More particularly, it relates to the recovery of uranium from ammonium carbonate leach solutions having molybdenum as an impurity.

Abstract: The invention relates to the purification of homo- and copolymers of mono- and diolefines, as obtained from a solution polymerization in the presence of uranium catalysts, from the catalyst residues, and particularly to a process for the removal of the uranium from the hydrocarbon solution of the polymer, which solution is mixed with an aqueous solution of ammonium carbonate, the thus resulting aqueous phase being then separately treated for the recovery of the ammonium carbonate and of the uranium.

Abstract: Process for the preparation of fine particle size uranium dioxide from a uranium trioxide feed comprising the following steps:A. reacting solid uranium trioxide with aqueous ammonium nitrate to form an insoluble ammonium uranateB. neutralizing the thus formed slurry with ammonium hydroxide to precipitate out as an insoluble ammonium uranate the remaining dissolved uraniumC. recovering the thus formed precipitates in a dry state andD. reducing the dried precipitate to uranium dioxide.

Abstract: The present invention relates to an improved method for the solution mining of a mineral from a subterranean formation. More specifically, the invention relates to an improved method which enhances significantly the recovery of a mineral from a subterranean formation via solution mining utilizing an injection and production well. The improvement comprises positioning an injection well in the reduced zone and a production well in the oxidized zone of a formation.

Abstract: Uranium is separated from analytical Group II and Group III metal ions in an aqueous liquor containing uranyl ions. The liquor is extracted with a non-interfering, water-immiscible, organic solvent containing a reagent which will react with the uranyl ions to form a complex soluble in the solvent. If the liquor is acidic, the solvent is washed with water. Then to the solvent is added an aqueous solution containing about 0.5 to 1.0 mole per liter of (NH.sub.4).sub.2 CO.sub.3 or NH.sub.4 HCO.sub.3 ions and sufficient sulfide ions to precipitate the metal ions as sulfides. The solvent and the aqueous solution are separated and the sulfides filtered from the aqueous solution. The ammonium-uranyl-tricarbonate in the aqueous solution can then be precipitated by increasing the concentration of (NH.sub.4).sub.2 CO.sub.3 or NH.sub.4 HCO.sub.3 ions to about 1.5 to 2.5 moles per liter. The precipitate is filtered and calcined to obtain U.sub.3 O.sub.8 or UO.sub.2.

Abstract: Process and apparatus for producing ammonium diuranate (ADU) from uranium hexafluoride (UF.sub.6) wherein gaseous UF.sub.6 is first hydrolyzed, then partially neutralized with aqueous ammonium hydroxide to a pH of from 5 to 6, the resultant solution is conveyed to a precipitation vessel wherein it is admixed with additional aqueous ammonium hydroxide to a pH of from 9.8 to 10.

Abstract: A process for treating the aqueous effluents that are produced in converting gaseous UF.sub.6 (uranium hexafluoride) into solid UO.sub.2 (uranium dioxide) by way of an intermediate (NH.sub.4).sub.4 UO.sub.2 (CO.sub.3) .sub.3 ("AUC" Compound) is disclosed. These effluents, which contain large amounts of NH.sub. 4.sup.+ (ammonium), CO.sub.3.sup.-.sup.- (carbonate), F.sup.- (fluoride), and a small amount of U (uranium), are mixed with H.sub.2 SO.sub.4 (sulfuric acid) in order to expel CO.sub.2 (carbon dioxide) and thereby reduce the carbonate concentration. The uranium is precipitated through treatment with H.sub.2 O.sub.2 (hydrogen peroxide) and the fluoride is easily recovered in the form of CaF.sub.2 (calcium fluoride) by contacting the process liquid with CaO (calcium oxide). The presence of SO.sub.4.sup.-.sup.- (sulfate) in the process liquid during CaO contacting seems to prevent the development of a difficult-to-filter colloid. The process also provides for NH.sub.3 (ammonia) recovery and recycling.

Abstract: A method of coprecipitating a mixed oxalate to be used as a laser material in which one of the reactants is introduced as droplets formed by an aerosol into a bath containing the second reactant.

Abstract: Uranium is separated from contaminating cations in an aqueous liquor containing uranyl ions. The liquor is mixed with sufficient recycled uranium complex to raise the weight ratio of uranium to said cations preferably to at least about three. The liquor is then extracted with at least enough non-interfering, water-immiscible, organic solvent to theoretically extract about all of the uranium in the liquor. The oganic solvent contains a reagent which reacts with the uranyl ions to form a complex soluble in the solvent. If the aqueous liquor is acidic, the organic solvent is then scrubbed with water. The organic solvent is stripped with a solution containing at least enough ammonium carbonate to pecipitate the uranium complex. A portion of the uranium complex is recycled and the remainder can be collected and calcined to produce U.sub.3 O.sub.8 or UO.sub.2.

Abstract: Waste water containing large amounts of fluorides and ammonia and trace amounts of uranium, as is produced during the hydrolysis and ammonium hydroxide treatment of uranium hexafluoride (UF.sub.6) to produce ammonium diuranate (ADU) therefrom, is rendered suitable for cyclic reuse by initially treating the waste water with sufficient lime to precipitate substantially all of the fluorides present in the waste water to a relatively insoluble CaF.sub.2 precipitate, the treated solution is subjected to distillation to drive off ammonia for reuse in the ADU precipitation, the CaF.sub.2 precipitate is separated from the aqueous distilland leaving water with dissolved calcium, the distilland is treated by a cationic ion-exchange material to remove substantially all of the calcium and other cationic metal impurities and the resulting water containing small amounts of uranium, fluoride and ammonia is recycled to react with UF.sub.

Abstract: In conventional processes for separating metals by selective precipitation, the precipitate is frequently in an amorphous form, eg a slime, which is difficult to filter, wash and handle generally. In the present process mixed solutions and/or sols of the metals are mixed with a suitable soluble organic polymer to form a viscous mixture. The latter is formed into discrete entities, e.g. drops, which are contacted with a reagent which converts each drop to a gelled entity containing at least one of the metals as a precipitate.In one form of the process, the same chemical reagent also serves to dissolve the other metal present in the initial drops. The metals are then easily separated by separating the gelled drops from the reagent. In another form, the reagent includes two different chemical reagents which contact the drop simultaneously, one to precipitate one metal in the gelled drops and the other to dissolve the other metal.