Abstract: A method of selectively removing impurities from a scandium-containing feed solution includes contacting an aqueous scandium-containing solution with an organic solvent stream containing an extractant, thereby forming a loaded organic solvent stream containing the impurity or impurities while leaving the scandium in the raffinate. The aqueous stream containing the scandium is washed, diluted and has inorganic salts added before being contacted with a second organic solvent stream to extract the scandium selectively, and followed by stripping the scandium from the scandium-containing loaded organic extractant stream by adding oxalic acid to the loaded organic extractant stream to form scandium oxalate.

Abstract: The invention discloses a method for recovering uranium from an acidic leach solution or leach pulp in salt water using an amino-phosphorus resin, wherein the liquid phase of the leach solution or leach pulp contains greater than 3 g/L chloride ion in solution. The resin may comprise a functional group comprising an amino phosphonic group, an amino-phosphinic group, an amino phosphoric functional group and/or a combination thereof. According to the invention the leach solution or leach pulp may be generated by in-situ leaching, vat leaching, heap leaching and/or agitated leaching at ambient, elevated temperature and/or elevated pressure conditions in saline or hyper-saline water.

Abstract: Process for degrading organic substances in an aqueous composition comprising a step (a) wherein, in a liquid reaction medium, said aqueous composition is reacted with at least one composition comprising hydroxide ions (OH?) and hypochlorite in a molar ratio between hydroxide and hypochlorite higher than or equal to 0.001 and lower than 1.5, in order to oxidize said organic substances.

Abstract: The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. Applications: reprocessing of irradiated nuclear fuels, in particular for recovering plutonium, neptunium, americium, curium and possibly uranium, present in the form of traces, in a pooled but selective fashion with regard to lanthanides, from a solution for the dissolution of an irradiated nuclear fuel, downstream of a cycle for the extraction of uranium.

Abstract: A method for treating process distillate heavies produced during uranium fluoride purification is described. The heavies contain primarily uranium hexafluoride, UF6, and molybdenum oxytetrafluoride, MoOF4. The uranium hexafluoride is removed via distillation at reduced pressure leaving essentially MoOF4 containing <0.1% of residual uranium hexafluoride. This mixture is hydrolyzed in water, then treated with a solution of sodium hydroxide until a pH of at least 7.5 is reached. The precipitated sodium diuranate and sodium fluoride are removed by filtration. The filtrate is reacted with calcium chloride to precipitate the molybdenum values as calcium molybdate containing trace quantities of calcium fluoride.

Abstract: Method for producing a uranium concentrate in the form of solid particles, by precipitation from a uranium-containing solution using a precipitating agent, in a vertical reactor comprising a base, a top, a central part, an upper part, and a lower part, the solid particles of the uranium concentrate forming a fluidized bed under the action of a rising liquid current which circulates from the base towards the top of the reactor successively passing through the lower part, the central part and the upper part of the reactor, and which is created by introducing a liquid recycling current (flow) at the base of the reactor, said liquid recycling current being tapped at a first determined level (A) in the upper part of the reactor and sent back without settling to the base of the reactor, excess liquid being also evacuated via an overflow located at a second determined level (B) in the upper part of the reactor; a method in which the upper limit (C) of the fluidized bed of solid particles is controlled so that it is

Abstract: Formulations useful for preparing hydrous hafnium oxide gels contain a metal salt including hafnium, an acid, an organic base, and a complexing agent. Methods for preparing gels containing hydrous hafnium oxide include heating a formulation to a temperature sufficient to induce gel formation, where the formulation contains a metal salt including hafnium, an acid, an organic base, and a complexing agent.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: Method for coprecipitation (or simultaneous precipitation) of at least one actinide in oxidation state (IV) with at least one actinide in oxidation state (III), wherein: a solution i.e. mixture of actinide(s) in oxidation state (IV) and actinide(s) in oxidation state (III) is prepared by adding to it a singly charged cation whose presence makes it possible to stabilize the aforementioned oxidation states in the mixture, or a singly charged cation which does not act to stabilize the aforementioned oxidation states in the mixture; a solution containing oxalate ions is mixed with the said mixture of actinides in order to carry out coprecipitation, i.e. simultaneous precipitation, of the said actinides in oxidation states (IV) and (III) and a fraction of the singly charged cation. According to another embodiment, a solution i.e.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.

Abstract: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

Abstract: A process for extracting uranium from uranium-containing waste residues. The process dissolves the residue in hot nitric acid, and precipitates the resulting solution with sodium carbonate/sodium bicarbonate.

Abstract: Metal values (especially uranium values) are extracted from aqueous solutions of metal oxyions in the absence of halogen ion using an imidazole of defined formula. Especially preferred extractants are 1-alkyl imidazoles and benzimidazoles having from 7 to 25 carbon atoms in the alkyl group.

Abstract: A method and means for continuous precipitation of granular uranium peroxide. The reaction vessel and agitation method practiced in it avoid filter plugging and caking problems.

Abstract: A process for removing metal salts from an H.sub.4 EDTA precipitate by esterification with an esterification reagent to produce an esterification mixture comprising a solid metal salt an EDTA ester and thereafter separating the solid metal salt from the esterification mixture.

Abstract: Making uranium dioxide pellets of controlled grain size by treating 50-500 g/l UO.sub.2 F.sub.2 with NH.sub.3 in a first and a second stages to form (NH.sub.4).sub.2 U.sub.2 O.sub.7 precipitate, wherein the NH.sub.3 /U molar ratio is between 3-5 in the first stage and between 6-12 in the second stage. The precipitate is then formed into UO.sub.2 pellets having grain size within the range from 10 to 100 .mu.m.

Abstract: The invention has as its object a process for the purification of a uranium solution which is contaminated by zirconium. This process consists of bringing the pH of the solution to a value of 1 to 3, then adjusting the pH of the said solution to, and maintaining it at a value of 2.5 to 3 while raising the temperature to a value of around 30.degree. to around 60.degree. C. for a period sufficiently long that there occurs a selective precipitation of at least 70% of the zirconium while the uranium remains in the dissolved state.

Abstract: A method of controlling the crystal grain size of UO.sub.2 pellets is disclosed, which comprises(1) providing an aqueous uranyl solution which is free of hydrofluoric acid and nitric acid and which contains uranyl fluoride (UO.sub.2 F.sub.2) and/or uranyl nitrate (UO.sub.2 (NO.sub.3).sub.2) as a uranium component,(2) reacting with ammonia said aqueous uranyl solution to precipitate ammonium diuranate (ADU), while adjusting the ratio of said uranium components to a predetermined value within the range varying from the ratio 100% of uranyl fluoride and 0% of uranyl nitrate to the ratio 0% of uranyl fluoride and 100% of uranyl nitrate, adjusting the uranium concentration of the reaction mixture to 50 to 1,000 gU/liter and also adjusting the rate of contact between said uranium component in said aqueous uranyl solution with ammonia to at least 2 moles NH.sub.3 /min/mole U,(3) calcining and reducing said ADU precipitate to form UO.sub.2 powder, and(4) molding and sintering said UO.sub.2 powder, thus producing UO.

Abstract: The invention relates to an ammonium plutonate uranate, its preparation process and its use for producing the mixed oxide (U,Pu)O.sub.2.The ammonium plutonate uranate of formula [Cu,Pu)O.sub.3 ].sub.2 [NH.sub.3 ].sub.x [H.sub.2 O].sub.y in which x and y have a value from 0 to 5 with x+y=5, is prepared by neutralization with ammonia of an aqueous solution of uranium VI and plutonium VI salts, e.g. uranyl and plutonyl nitrates. This solution can be prepared by oxidizing a solution of plutonium IV and uranyl nitrate. By calcining under a reducing atmosphere the ammonium plutonate uranate precipitate formed, a mixed oxide (U,Pu)O.sub.2 is obtained having a solubility in 11N HN O.sub.3 without fluoride of at least 99.9%.

Abstract: The present invention provides a process for producing a uranium oxide by dissolving a yellow cake in sulfuric acid or hydrochloric acid, bringing the obtained solution into contact with a chelating resin of diaminocarboxylic acid type and subjecting the product to neutralizing precipitation followed by heat treatment. By the contact of the solution with the chelating resin, iron, copper, molybdenum and vanadium among the metallic impurities are removed and in the subsequent neutralizing precipitation step, other metallic impurities such as aluminum, calcium, magnesium, sodium and potassium are removed. This process can easily produce uranium oxide having a high purity using a simple apparatus.

Abstract: The present invention relates to a proces for separating rare earths and uranium from a UF.sub.4 concentrate and for placing them into useful form, comprising attacking the UF.sub.4 concentrate with potassium hydroxide, followed by nitric acid dissolution of the resulting precipitate, and extraction of the solution using an organic solvent. The extraction results in an organic solution containing purified uranyl nitrate, and an aqueous solution from which rare earths are precipitated by means of a base.

Abstract: A method of decontaminating mixtures of radioactively contaminated water and radioactively contaminated water-immiscible organic liquids to produce a decontaminated organic liquid which can be disposed of in a conventional manner.

Abstract: Process for the selective separation of uranium in carbonated aqueous solution from soluble impurities including in particular at least one of the elements zirconium and/or hafnium and/or molybdenum, thereby making it possible to produce a uraniferous liquor in a greatly improved state of purity, comprising precipitation of the zirconium and/or hafnium by introducing an alkali metal hydroxide, then separation of the zirconiferous precipitate which possibly contains hafnium and/or molybdenum from the resulting uraniferous solution with its reduced impurities content, which still possibly contains molybdenum, which is characterized in that in order to increase the removal of zirconium and/or hafnium and/or molybdenum, the uraniferous carbonated aqueous solution is treated by means of a dilute alkaline aqueous solution containing at most 17 g/l of at least one alkali metal hydroxide expressed in respect of OH.sup.

Abstract: UO.sub.2 base powder exhibiting any specific surface and crystallite diameter properties is mixed with rare earth (SE) oxide-containing powder, the particles of which exhibit at least in one surface layer, a crystal lattice of the fluorite type, with the stoichiometric composition (SE.sub.0.5, U.sub.0.5) 0.sub.2.00 and/or form it in sintering; and is compacted to form compacts which are sintered in a gas atmosphere with reducing action at 1500.degree. C. to 1750.degree. C. to form high-density sintered bodies.

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 is described for the recovery of uranium from wet-process phosphoric acid utilizing an alkyl pyrophosphoric extractant. The extractant also contains a modifier for retaining the alkyl pyrophosphoric acid in solution in an essentially water-immiscible organic diluent during stripping. After extracting the uranium from wet-process acid, the APPA extractant is treated with an oxidizing agent such as hydrogen peroxide and the uranium stripped into an alkaline stripping solution. The alkaline stripping solution is an aqueous solution of an alkali metal or ammonium carbonate or hydroxide. The barren extractant is recycled for contacting with fresh wet-process acid. The uranium and any ferric iron present is precipitated in the stripping solution. The uranium is dissolved in an aqueous solution such as an aqueous solution of ammonium uranyl tricarbonate (AUT), any precipitated ferric solids are separated from the solution, and the uranium is reprecipitated, separated, dried and calcined to a U.sub.3 O.sub.

Abstract: A process is described for the recovery or uranium from wet-process phosphoric acid utilizing an alkyl pyrophosphoric extractant. The extractant also contains a modifier for retaining the alkyl pyrophosphoric acid ester in solution in an essentially water-immiscible organic diluent during stripping. After extracting the uranium from wet-process acid, the APPA extractant is treated with an oxidizing agent such as hydrogen peroxide and the uranium stripped into an alkaline stripping solution. The alkaline stripping solution is an aqueous solution of an alkali metal or ammonium carbonate unsaturated in uranium. The barren extractant is recycled for contacting with fresh wet-process acid. Any ferric iron present is precipitated in the stripping solution and then separated from the solution. The uranium is precipitated, separated, dried and calcined to a U.sub.3 O.sub.8 product.

Abstract: A process for the selective removal of arsenical material in the course of a process for the hot oxidizing attack on a uraniferous and/or molybdeniferous ore containing arsenical materials, comprising attacking said ore with an aqueous liquor of sodium or potassium carbonate and/or bicarbonate, said attack operation being carried out under conditions with respect to levels of concentration, temperatures and pressures which cause solubilization of the uranium and/or molybdenum and the arsenic present in the ore, then collecting a suspension of a solid phase in a liquid phase, and finally, separating said phases, wherein the arsenic which is solubilized in the attack operation is extracted in the form of magnesium arsenate by treating the material containing the arsenic with a magnesium compound.

Abstract: The invention relates to a process for recovering and purifying the molybdenum from a solution containing more molybdenum than vanadium.The process of the invention comprises the step of contacting the solution which contains molybdenum and vanadium, originally at a pH at least equal to 11 with an excess of an ammonium salt sufficient to precipitate the greatest part of vanadium in the form of solid ammonium vanadate, this excess being yet hot higher than a value such that after precipitation of vanadium, the quantity of ammonium having not reacted with the vanadate would be over a concentration which would account for risks of precipitating also molybdenum which is contained in the liquid phase of the supernatant.The molybdenum which is obtained is of industrial grade.

Abstract: A method for treating process waste consisting of uranium hexafluoride, UF.sub.6, molybdenum hexafluoride, MoF.sub.6 and molybdenum oxytetrafluoride, MoOF.sub.4 is provided wherein the foregoing mixture is hydrolyzed in an aqueous solution of ammonium carbonate and ammonium hydroxide. The resulting mother liquor is digested at 60.degree.-80.degree. C. to drive off carbon dioxide and convert the ammonium uranyl carbonate, (NH.sub.4).sub.4 UO.sub.2 (CO.sub.3).sub.3 to ammonium diuranate, (NH.sub.4).sub.2 U.sub.2 O.sub.7. The pH is maintained above 9 to prevent premature precipitation of the molybdenum values as molybdic acid or uranyl molybdate.The ammonium diuranate which forms as a yellow solid is filtered, slurried in aqueous ammonium hydroxide, filtered and dried. The yield is quantitative. Pyrolysis results in conversion of the diuranate salt to uranium oxide, U.sub.3 O.sub.8, of suitable purity to be recycled in the uranium hexafluoride production process.

Abstract: A method for manufacturing Pu.sub.2 -containing crystals by oxidizing plutonium in a heavily contaminated nitric acid starting solution to Pu.sup.6+, precipitating ammonium plutonyl carbonate containing crystals by reaction with ammonium- and carbonate ions and subsequent separating and calcining of these crystals. The nitric acid solution with the oxidized Pu.sup.6+ as well as carbonate and/or acetate ions and NH.sub.3 are dosed into an aqueous starting solution of ammonium carbonate or ammonium acetate in a concentration such that the mol ratio of carbonate and/or acetate ions to the Pu.sup.6+ -ions in the solution is at least 4:1 and plutonium containing carbonato- or acetato complex compounds which remain dissolved in the receiver solution, are formed, while difficult-to-dissolve metal hydroxide compounds are precipitated. Thereupon NH.sub.3 and carbonate ions are added to precipitate the ammonium plutonyl carbonate containing crystals.

Abstract: Method for precipitating and removing soluble metal compounds from solutions of phosphoric acid. The method is useful in the disposal of metal-containing phosphoric acid waste from electrolytic operations, including such acid solutions contaminated with uranium compounds.

Abstract: This invention relates to a process for reducing to environmentally acceptable levels the amount of soluble or mobile contaminants in formations that have been subjected to in situ leaching to recover mineral values therefrom. The present process is also applicable to surface milling operations. In accordance with the present invention, an aqueous solution containing orthophosphates is introduced into the formation to immobilize contaminants by forming a relatively stable precipitate. For example, radionuclides such as uranium, thorium or radium form stable precipitates with orthophosphates.When the formation fluids contain a high concentration of carbonates, the formation fluids are treated to reduce the carbonates concentration to levels which do not substantially interfere with the phosphate precipitation of contaminants.

Abstract: A process for recovering from a wet process phosphoric acid which contains uranium, a uranium containing concentrate and a purified phosphoric acid. The wet process phosphoric acid is treated with a precipitant in the presence of a reducing agent and an aliphatic ketone.

Abstract: Method for the manufacture of oxidic sintered nuclear fuel bodies by compacting UO.sub.2 -starting powder or a mixture of UO.sub.2 -and PuO.sub.2 starting powder which contains up to 10% by weight rare-earth oxide, especially Gd.sub.2 O.sub.3, as an additive into blanks and subsequent densification of these blanks by a heat treatment in a sintering atomsphere with reducing action. The UO.sub.2 -starting powder used for compacting has a specific surface in the range of 2 to 4.5 m.sup.2 /g and/or a mean crystallite diameter in the range of 80 nm to 250 nm, and the heat treatment in the sintering atmosphere with reducing action is carried out at a temperature in the range of 1,500.degree. C. to 1,750.degree. C.

Abstract: A process for extracting arsenic from aqueous solutions containing alkali metal carbonate, sulfate and hydroxide or hydrogen carbonate, at least one metal selected from the group consisting of vanadium, uranium and molybdenum in the form of an alkali metal salt, and inorganic and/or organic impurities, comprising the steps of caustifying the solution with lime to convert the alkali metal carbonates into hydroxides to precipitate the insoluble calcium salts formed, concentrating the solution by evaporation of the effluent liquids originating from the caustification step to obtain a precipitate substantially comprising alkali metal sulfate wherein the caustification is performed in two steps, consisting of (a) treating the solutions with an amount of lime approximately equal to but less than the stoichiometric amount necessary to convert the alkali metal carbonates into hydroxides as a first precipitate, and (b) after separating and washing the first precipitate, treating the resulting liquor with at least the

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 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: The present invention relates to a process for recovery of uranium from a carbonate lixiviant additionally containing other contaminants such as molybdenum and silica, by adjusting the pH of the lixiviant to a value of at least 4 and treating the lixiviant to obtain a carbonate concentration of at least 50 ppm by weight. Subsequently the lixiviant is treated with an aqueous solution containing a sufficient amount of ferric iron to precipitate the contaminants without substantial precipitation with the uranium values. The precipitate is separated from the lixiviant. The treated lixiviant is passed through an ion exchange to retain uranium values.

Abstract: A process for the purification of solutions containing sodium or potassium carbonate, sulphate, hydroxide or hydrogen carbonate, and mainly at least one of the metals belonging to the group formed by vanadium, uranium or molybdenum, in the form of sodium or potassium salts, and inorganic and/or organic impurities, wherein the above-mentioned solutions are completely or partially caustified by the addition of an adequate amount of lime, whereby a first precipitate essentially containing calcium carbonate is separated, and the separated liquor is concentrated by evaporation until the hydroxide content is at most equal to 50%, to cause the production of a second precipitate which essentially comprises sodium or potassium sulphate, then, after separation thereof, a hydroxide-rich liquor is collected.This process is more particularly adapted for treatments of liquors resulting from the alkaline attach of vanadiferous and uraniferous ores.

Abstract: A process for obtaining molybdenum as a useful product from aqueous solutions to be purified, according to claim 1 of French patent No.

Abstract: Entrainment of contaminated water in the organic phase and poor phase disengagement is prevented in the second cycle scrubber, in a two cycle uranium recovery process, by washing the organic solvent stream containing entrained H.sub.3 PO.sub.4 from the second cycle extractor, with a dilute aqueous sulfuric or nitric acid solution in an acid scrubber, prior to passing the solvent stream into the second cycle stripper.

Abstract: Process for the purification of an aqueous solution containing alkali metal carbonate, sulphate, and hydroxide or hydrogen carbonate and primarily one at least of the metals belonging to the group formed by vanadium, uranium and molybdenum, in the form of alkali metal salts, and mineral and/or organic impurities, said solutions resulting from an ore attack cycle and being taken off after extraction of the metal being sought, which comprises treating said solutions by means of lime at a temperature which is at most equal to the boiling temperature, to convert the carbonate and hydrogen carbonate present into alkali metal hydroxide, then effecting separation and washing of a first precipitate of calcium carbonate and a liquor which is enriched in respect of alkali metal hydroxide, concentrating it by evaporation to cause production of a second precipitate which essentially comprises alkali metal sulphate, separating it from the hydroxide-rich liquid phase, re-dissolving it in an aqueous medium and treating the

Abstract: A process for the selective removal of arsenical materials, in the course of a continuous hot oxidizing attack process, comprising attacking a uraniferous ore containing arsenical materials as impurities in the presence of an oxidizing agent in the reaction medium by means of an aqueous liquor formed by a recycling solution containing alkali metal carbonate and bicarbonate, and uranium close to the limit of solubility thereof, under concentration, temperature and pressure conditions which causes solubilization of the uranium present in the ore, and re-precipitation thereof in the attack medium, then collecting a suspension of a solid phase in a liquid phase which, after cooling, is subjected to a separation operation, recycling the liquid phase to the attack operation, and treating the separated solid phase by means of an aqueous liquor to re-dissolve the precipitated uranium which is characterized in that the arsenic which is solubilized in the attack operation is extracted by means of a magnesium compound w

Abstract: A method for co-conversion of aqueous solutions of one or more heavy metal nitrates wherein thermal decomposition within a temperature range of about 300.degree. to 800.degree. C. is carried out in the presence of about 50 to 500% molar concentration of ammonium nitrate to total metal.

Abstract: A method is disclosed for separating uranium from molybdenum in an aqueous solution. The pH of the solution is lowered to about 1 to about 4 using an inorganic acid such as hydrochloric acid. The carbon dioxide is then removed from the solution which can be done by aeration. The pH of the solution is raised to about 9 in the presence of ammonium ions which result in the precipitation of ammonium diuranate. The precipitated ammonium diuranate may then be filtered or otherwise separated from the molybdenum which remains in the solution.

Abstract: A process for recovering uranium values from a sulphate solution containing dissolved uranium and molybdenum and with a pH not exceeding about 5.5, includes reacting the solution with ammonia at a pH in the range of from about 8 to about 10, without the solution existing for any significant time at a pH of around 7, with resultant precipitation of uranium values relatively uncontaminated by molybdenum. The uranium containing precipitate is separated from the remaining solution while maintaining the pH of the remaining solution within the same range.

Abstract: A process for treating water containing chemicals comprising ammonium nitrate, fluorides, hydrogen peroxide and uranium compounds, wherein valuable and/or hazardous components of the chemicals-containing water are separated therefrom and concentrated for recovery. The process includes an ion exchange operation in addition to manipulations of chemical components and pH conditions of the water by means of the application of reagents and adsorbents.