Abstract: The invention provides hydrometallurgical processes for the recovery of rare earth values from ore, using simple crushing without beneficiation to produce an enriched and purified mixed rare earth concentrate. Ore is crushed to a relatively coarse particle size, and then treated with relatively small amounts of acid, at a relatively modest elevated temperature, to render the rare earth elements extractable in a subsequent water leach.

Abstract: Provided is a process for recovering metals from solid radioactive waste, preferably uranium, cesium, mercury, thorium, rare earths or combinations thereof. The process comprises a leaching step and a separation step. The leaching step comprises contacting the solid radioactive waste with an aqueous inorganic acid and a leaching salt to produce a mixture of a metal-rich leachate and a metal-poor waste, which are separated in the separation step. Also provided is a process for recovering metals from solid radioactive waste comprising an attrition step, a leaching step, a washing step, a combination step and a recovery step.

Abstract: Methods for the extraction of metals such as rare earth metals and thorium from metal compounds and solutions. The methods may include the selective precipitation of rare earth elements from pregnant liquor solutions as rare earth oxalates. The rare earth oxalates are converted to rare earth carbonates in a metathesis reaction before being digested in an acid and treated for the extraction of thorium. A two-step extraction method may be applied to precipitate thorium as thorium hydroxide under controlled pH conditions such that pure thorium precipitate is recovered from a first step and a thorium-free rare earth solution is recovered at the subsequent step. The resulting rare earth solutions are of extremely high purity and may be processed directly in a solvent extraction circuit for the separation of rare earth elements, or may be processed for the direct production of a 99.9% bulk rare earth hydroxide/oxide concentrate.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: An in situ recovery of uranium operation involves circulating reactive fluids through an underground uranium deposit. These fluids contain chemicals that dissolve the uranium ore. Uranium is recovered from the fluids after they are pumped back to the surface. Chemicals used to accomplish this include complexing agents that are organic, readily degradable, and/or have a predictable lifetime in an aquifer. Efficiency is increased through development of organic agents targeted to complexing tetravalent uranium rather than hexavalent uranium. The operation provides for in situ immobilization of some oxy-anion pollutants under oxidizing conditions as well as reducing conditions. The operation also artificially reestablishes reducing conditions on the aquifer after uranium recovery is completed. With the ability to have the impacted aquifer reliably remediated, the uranium recovery operation can be considered inherently safe.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: A formic acid aqueous solution that contains Fe (II) ions is produced by dissolving metal iron in a formic acid aqueous solution. Nitrogen is supplied from a nitrogen supply device to a chemical liquid tank and then discharged from a discharge line to reduce the dissolved oxygen concentration in the aqueous solution. The chemical liquid tank is filled with the formic acid aqueous solution sealed with nitrogen, and is transferred from a factory to a nuclear reactor building designated as radiation-controlled areas. Inside the nuclear reactor building, the chemical liquid tank is installed in a film deposition apparatus connected to a reactor water recirculation pipeline. The formic acid aqueous is supplied from the chemical liquid tank to the inside of the reactor water recirculation pipeline, and then a ferrite film is formed on the inner surface of the reactor water recirculation pipeline.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: The present invention relates to a method for the volumetric decontamination of radioactive metals. The method includes the step of precipitating out radioactive gamma and beta emitting nucleotides and then recovering non-radioactive metal compounds.

Abstract: Methods and systems for removing copper minerals from a molybdenite concentrate. One embodiment provides leaching copper from the molybdenite concentrate with a leaching solution comprising ferric chloride, removing molybdenite from the leaching solution, introducing an acid into the leaching solution and introducing O2, O3, or a combination of both, into the leaching solution. A method for regenerating ferric chloride in a leaching solution is also provided. One embodiment provides adding a leaching solution comprising Fe(II) ions, Fe(III) ions, or a combination of both, to a mixture of mineral sulfides, introducing an acid into the leaching solution, and introducing O2, O3, or a combination of both, into the leaching solution.

Abstract: An apparatus for the removal of uranium from a body of material is provided. The apparatus has at least one ultrasonic extractor, having a bottom and a top. The at least one ultrasonic extractor is configured to accept solids at the bottom and acid at the top, and has a mixing screw and at least one source of ultrasonic energy. The mixing screw is configured to transport the solids in a direction countercurrent to the acid in the at least one ultrasonic extractor; and the source of ultrasonic energy is configured to impart ultrasonic energy into the solids and the acid, as the solids and the acid traverse the at least one ultrasonic extractor countercurrently.

Abstract: The present invention relates to a process for the recovery of uranium in high silica environments comprising the use of a strong base macroreticular ion exchange resin.

Abstract: A method of suppressing deposition of radionuclides on components of a nuclear power plant comprises forming a ferrite film by contacting a first chemical including iron (II) ions, a second chemical for oxidizing the iron (II) ions to iron (III) ions, and a third chemical for adjusting the pH of a processing solution containing a mixture of the first and second chemicals to be 5.5 to 9.0 with the metal member surface in a time period from a finishing stage in decontamination step of removing contaminants formed on the surface of metal member composing the nuclear power plant, and suppressing deposition of radionuclides on the metal member by the ferrite film.

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: A reprocessing process of spent nuclear fuels for roughly separating U and U—Pu from FP, TRU and the like in a nitric acid solution of spent nuclear fuels by utilizing phenomenon of cocrystallization of hexavalent U and Pu. For example, spent nuclear fuels are sheared and dissolved in nitric acid, and insoluble residues in the nitric acid solution are removed. Then, a nitric acid concentration in the solution is adjusted and a valence of Pu in the solution is adjusted to tetravalence. The solution is then cooled to crystallize uranyl nitrate hydrate crystals and separated into the crystals and a mother liquor, and the separated crystals are recovered as a U product. Then, a nitric acid concentration in the separated mother liquor is adjusted and a valance of U and Pu in the mother liquor is adjusted to hexavalence, and the mother liquor is cooled to crystallize uranyl-plutonyl nitrate hydrate crystals which are separated and recovered as a U—Pu mixed product.

Abstract: A process for recovering metal values from a metal containing material including the metal values comprising digesting the metal containing material in a sulfuric acid solution comprising sulfuric acid, a reducing agent, and a carbon source, heating the digestion mixture and separating the resulting solution from the remaining solids. The sulfuric acid solution may additionally include hydrofluoric acid (HF) as a source of fluoride ion.

Abstract: A process for the decontamination of radioactively contaminated metal which comprises contacting the metal with a decontamination reagent solution containing an organic acid and an oxidising agent, allowing said solution to react with the contaminated metal at a pH of up to 4.5, treating the resulting solution to cause substantially complete precipitation of dissolved metal together with radionuclides and separating precipitated material, containing radioactive contaminants, from said solution.

Abstract: A process for oxidation of ferrous ions in solution, and more particularly a process for improved base metal and/or uranium leaching from ores, concentrates or tailings using ferric ion as an oxidizing agent. A reaction vessel (10) holds a ferrous ion-containing solution, for example, a copper sulphide leach slurry or concentrate. An agitator (12) may be provided to promote leaching of the base metal into solution. Some of the ferrous ion-containing solution is drawn off from the reaction vessel (10) and pumped through an in-line mixer (14) via a feed pump (16). Oxygen is injected into the reactor (14) to facilitate oxidation of the ferrous sulphate to form ferric sulphate. The ferric ion-containing solution is then recirculated back to the reaction vessel (10) where the ferric ions are reused in the dissolution of copper sulphide into soluble copper sulphate.

Abstract: The steps for recovering uranium and transuranic elements are simplified, and the generation of waste solvent and waste materials is suppressed. Spent nuclear fuel is dissolved in nitric acid (S100) and the resulting solution is subjected to electrolytic oxidation so that U, Np, Pu, Am is oxidized to VI using Ce as oxidation catalyst. The solution is cooled, and nitrates of valence VI thereby deposit as crystals and are separated from the mother liquor (S104). The mother liquor is heated and concentrated (S114). The mixed crystalline deposit is dissolved in nitric acid (S106), uranyl nitrate is deposited alone by cooling (S108), and the crystals are separated from the U, Np, Pu, Am mixed solution (S110). The uranyl nitrate is dissolved in nitric acid (S112), and the heated and concentrated mother liquor is added to it to prepare another mixed solution.

Abstract: A process for facilitating removal of one or more impurities from titaniferous material containing the impurities in a form which is highly soluble in acid. The material is in turn leached with a sulphuric acid solution and with a hydrochloric acid solution in either order. To enhance the susceptibility of the impurities to removal, the hydrochloric acid leach is augmented by one or more of the following: (i) addition of an effective amount of an added chloride salt; (ii) pre-treatment of the titaniferous material with a solution of an effective amount of a carbonate salt, preferably an alkali metal or alkaline earth metal carbonate; and/or (iii) pre-treatment of the titaniferous material with a solution of an effective amount of a hydroxide, preferably an alkali metal or alkaline earth metal hydroxide.

Abstract: Soil comprising small soil particles, clay and silt particles, humus, fine vegetation, and contaminated with soluble or insoluble radioactive species is treated by first introducing an aqueous extracting solution comprising a mixture of sodium and potassium carbonate (or bicarbonate), or ammonium carbonate (or bicarbonate) into the soil to solubilize and disperse the radioactive species into solution. The extracting solution has a pH greater than or equal to about 7.5. Contaminated fine vegetation then is separated from the soil and extracting solution. Next, an acid like hydrochloric acid is introduced into the soil. The acid is added in an amount sufficient to lower the pH of the extracting solution at which point desirable organic material will substantially precipitate or coagulate from the extracting solution. The cleansed soil particles, including organic matter, is separated from the contaminated extracting solution.

Abstract: A process for recovering metal and acid values from a source material containing metallic fluorides comprises digesting the source material in sulfuric acid to form a slurry, separating a fluoride containing solid phase and a metal containing first liquid phase. The solid phase is subjected to pyrohydrolysis, sulfuric and hydrofluoric acids are recovered, and the first liquid phase is processed to recover the metal values by solvent extraction or ion exchanges. The tantalum values are extracted from the first liquid aqueous phase by a water immiscible organic extractant such as methylisobutyl ketone to form a first liquid organic phase containing tantalum and a second liquid aqueous phase. The tantalum is stripped from the first organic phase using water. The process includes the additional steps of heating the separated solid phase from about ambient temperature to an elevated temperature in the presence of water vapor to evolve sulfuric acid and render the gangue chemically inert.

Abstract: The process for recovering Ta/Nb values from highly flourinated ore materials by the process of;(a) contacting the materials with one or a mixture of HNO.sub.3 or HCl, and H.sub.3 BO.sub.3 wherein the H.sub.3 BO.sub.3 to other acid molar ratio is from about 1/10 to about 1/1,(b) maintaining the temperature between about 55 C. and about 85 C.,(c) reacting for a sufficient period to substantially solubilize the major portion of the materials and to substantially insolubilize all Ta/Nb values, and(d) separating the solids containing the Ta/Nb values from solution.

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: The invention is a process for selectively extracting strontium values from aqueous nitric acid waste solutions containing these and other fission product values. The extractant solution is a macrocyclic polyether in an aliphatic hydrocarbon diluent containing a phase modifier. The process will selectively extract strontium values from nitric acid solutions which are up to 6 molar in nitric acid.

Abstract: A process for decontaminating radioactive material comprises the step of contacting the material with a dissolving composition to dissolve the contaminants in the material, said composition comprising a dilute solution of about 0.05 molar ethylene diamine tetraacetic acid, about 0.1 molar carbonate, about 10 grams per liter hydrogen peroxide and an effective amount of sodium hydroxide to adjust the pH of the composition to a pH from about 9 to about 11. Also included are the steps of separating the dissolving composition containing the dissolved contaminants from the contacted material and recovering dissolved contaminants from the dissolving composition that has been separated from the material. A composition for dissolving radioactive contaminants in a material, comprising a dilute solution having a basic pH and effective amounts of a chelating agent and a carbonate sufficient to dissolve radioactive contaminants is also provided.

Abstract: A process for removing molybdenum from uranium bearing solutions reduces molybdenum contamination of uranium product. The process employs solvent extraction techniques to remove molybdenum by using a molybdenum-selective extraction reagent of the chelating hydroxy-oxime type. The molybdenum-loaded extractant is stripped with dilute alkali for reuse. Where the uranium bearing solution originates from a solvent extraction process, a pretreatment stage is applied using an agent to remove entrained uranium extractant.

Abstract: A method of removing niobium from a uranium-niobium alloy includes dissolving the uranium-niobium alloy metal pieces in a first aqueous solution containing an acid selected from a group consisting of HCl and H.sub.2 SO.sub.4 and a fluoboric acid as a catalyst to provide a second aqueous solution which includes uranium (U.sup.+4), acid radical ions, the acids and insolubles including uranium oxides and niobium oxides; adding nitric acid to the insolubles to complete the oxidation of the niobium oxides to yield niobic acid and further solubilizing thr uranium oxide; and separating the niobic acid, nitric acid and solubilized uranium oxides.

Abstract: Sodium silicate (waterglass) is added to a waste process stream containing metal to form a waterglass sludge containing the metal contaminants. The waterglass sludge matrixed with the metal is removed from the stream. Thereafter, a caustic is added to the separated waterglass sludge which dissolves the sodium silicate, leaving a metal solid in suspension. The sodium silicate solution is filtered off and recycled for use in the waterglass precipitation process. The metal solid remains and concentrated acid is added to form a metallic acid solution which is substantially free of silicates. This solution may then be treated by solvent extraction or other means to recover the metal. The process provides substantial metal recovery from the process waste stream and eliminates the need for burial of the waste, thereby eliminating burial costs.

Abstract: A metallurgical processing system for economically recovering metal values, such as columbium, tantalum, thorium, and uranium from dilute source solids, such as digestion sludges, by a series of steps including:1) slurrying the source solids with dilute hydrofluoric acid to produce a solid phase and a liquid phase containing dissolved tantalum and columbium, then extracting tantalum and/or columbium from the liquid phase by means of a liquid ion-exchange process and then, additionally;2) roasting the solid phase with sulfuric acid to recover and recycle hydrofluoric acid, leaching the roasted solids with dilute sulfuric acid to produce a disposable solid phase and a liquid phase containing thorium and uranium, and extracting thorium and uranium from the liquid phase by means of a liquid-liquid amine extraction process.

Abstract: In a method of producing uranium (IV) fluoride, a feedstock comprising uranium metal or alloy is dissolved in hydrochloric acid and hydrofluoric acid to produce a clear solution. The solution is heated and excess hydrofluoric acid added to produce a precipitate comprising uranium (IV) fluoride.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group colon hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: The separation of uranium, iron and scandium is achieved by the following process wherein a material containing these values is dissolved in mineral acid to form an aqueous solution, thereafter, an iminodiacetic acid cationic ion exchange resin is used to retain a major portion of the scandium and uranium. A raffinate containing the iron is formed. The ion exchange resin having said scandium and uranium retained thereon is rinsed with a dilute acid to remove residual metals other than scandium and uranium. The resin is eluted with an aqueous solution of an organic chelating acid to remove a major portion of the scandium from said resin and to form an aqueous solution containing scandium. The resin is then eluted with an aqueous solution of a mineral acid to remove a major portion of the uranium from the resin to form an aqueous solution containing the uranium. The scandium and uranium are recovered as solids.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably at least one substance from the group: hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on pressurized water reactors, boiling water reactors, metallic substances, high temperature alloys and brickworks as well. The used decontamination solution can, after regeneration, by recycled into the decontamination process. Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries. The decontamination agent (NBF.sub.

Abstract: The invention relates to a process for improving the effectiveness of the removal of zirconium from an aqueous, zirconium-containing nuclear fuel and/or fertile material solution in nitric acid in a liquid-liquid extraction process.Accordingly, the object of the invention is to improve the removal of zirconium from the reprocessing solutions and, at the same time, to simplify the course of the process. The invention seeks to improve decontamination of the uranium product and the plutonium product while, at the same time, reducing the outlay involved.According to the invention, this object is achieved in that, in a process step carried out before the first extraction of the nuclear fuels and/or fertile materials, the zirconium is converted from the dissolved state into a filterable or centrifugable solid phase by the use of an adsorbent from the group of inorganic ion exchangers and is removed from the aqueous solution together with the adsorbent.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group, hexafluorosilicate acid, fluoroboric acid, and the salts of both these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: A process for separately recovering uranium and hydrofluoric acid from a waste liquor containing uranium and fluorine comprises a neutralizing precipitation step wherein a magnesium compound is added to the waste liquor containing uranium and fluorine to form a precipitate and the thus formed precipitate is then separated; a distillation step wherein an aqueous solution of sulfuric acid is added to the precipitate separated in said neutralizing precipitation step to dissolve the precipitate and the thus formed solution is then distilled to recover hydrofluoric acid as a distillate; and a uranium recovery step wherein uranium is recovered from a residue produced by said distillation step.

Abstract: Uranium is leached from water slurries of uranium ore by incorporating a mixture of sulfur dioxide and air therein to provide the oxidizing and acidifying requirements to accomplish leaching.

Abstract: A process for the conversion into a usable condition of actinide ions contained in the solid residue of a sulfate reprocessing process for organic actinide-containing radioactive solid waste, which are present in the form of water soluble sulfato complexes. The residue is absorbed with water of 1 to 2 molar nitric acid so that the residue or the largest amount of residue goes in the solution. The resulting solution is separated from the insoluble constituents of the residue in case of any insoluble residue, and heated to a temperature in the range of 40.degree. C. below the boiling point of the solution to form a hot solution. To the hot solution is added an aqueous barium nitrate solution having an amount of barium nitrate which corresponds to a small excess of barium ions over the amount required stoichiometrically for complete precipitation of the sulfate ions. The resulting reaction solution is held at a selected temperature in the same range as above for a period in the range of 0.5 to 2 hours.

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 obtaining magnesium sulfate substantially free from radioactive uranium from a slag containing the same and having a radioactivity level of at least about 7000 pCi/gm. The slag is ground to a particle size of about 200 microns or less. The ground slag is then contacted with a concentrated sulfuric acid under certain prescribed conditions to produce a liquid product and a solid product. The particulate solid product and a minor amount of the liquid is then treated to produce a solid residue consisting essentially of magnesium sulfate substantially free of uranium and having a residual radioactivity level of less than 1000 pCi/gm. In accordance with the preferred embodiment of the invention, a catalyst and an oxidizing agent are used during the initial acid treatment and a final solid residue has a radioactivity level of less than about 50 pCi/gm.

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: The invention relates to a method for recovering rare metals from the combustion residues of various coals, in particular brown coals. The recovery is performed by digestion, more particularly by aqueous and/or dilute alkaline and/or dilute acidic digestion, where two or three of these steps can be combined in any desired order or they may be carried out separately. During digestion the concentration of the solution is monitored and the subsequent digestion steps are terminated at a desired concentration. The solid and liquid phase are then separated and the rare metals are isolated from the liquid phase while the solid phase, optionally after neutralization and/or washing can be utilized for example as a source of energy.

Abstract: A process for the recovery of the uranium present in phosphoric acid produced by a wet process in (A) by means of a suitable solvent (L.sub.9), which is carried out after separation of the gypsum in (B) formed in the attack operation, and elimination of the solid materials which are still in suspension, characterized in that, in order to avoid the formation of dross in the course of the operation of extracting the uranium which is previously reduced in (D), the step of eliminating the solid materials which are still in suspension is effected by a final solid-liquid separation operation in (C) and/or in (E), prior to extraction of the uranium in (F), in the presence of an added fraction of gypsum (S.sub.22) and/or (S.sub.23) resulting from the production of H.sub.3 PO.sub.4.

Abstract: The uranium, yttrium, thorium and rare earth values contained in a phosphate rock are essentially totally recovered therefrom by, during the wet-process preparation of phosphoric acid from said phosphate rock, acidulating same in the presence of aluminum and/or iron, and optionally silica, and thence recovering said values from the phosphoric acid thus produced.

Abstract: A process for separating radium-226 from an aqueous liquid by adding a precipitant to the liquid. The liquid, including the precipitant, is passed upwardly through a particulate bed to assist co-precipitation and removal of the radium-226. The effect is to produce a fluidized bed of the particulate bed. The process is efficient, rapid and undemanding of space, all advantages not found in the prior art.

Abstract: The present invention relates to insoluble compositions, which are capable of removing metal (e.g. selectively) from solution (e.g. Fe.sup.3+ from a liquid nutrient medium so as to lower the Fe.sup.3+ content to less than 0.1 .mu.M); the insoluble compositions comprise: a suitable insoluble carrier and organic co-ordinating sites covalently fixed to the surface of said carrier, said co-ordinating sites being capable of chelating Fe.sup.3+, Th.sup.4+ and/or UO.sub.2.sup.2+.

Abstract: A process for selectively separating plutonium from uranium and other metals is disclosed wherein, in order to precipitate and remove the pertaining complexes, use is made of a sufficient quantity of a cationic compound containing at least one group adapted to assert affinity to polar surfaces and containing a radical which has little affinity to water, e.g. surface active agents and the like, and use is made of the capability of Pu.sup.4+ and U.sup.4+ to form nitrato-complexes and of the capability of Pu.sup.4+ and UO.sub.2.sup.2+ to form sulfato-complexes.

Abstract: The rate of separation of insoluble material from an acid-leached uranium slurry is improved by adding a flocculating agent that is a combination of polyacrylamide and a copolymer of acrylamide with 2-acrylamide-2-methylpropanesulphonic acid or a water soluble salt thereof with acrylamide.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group.Iadd.: .Iaddend..[.colon.]. hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid.