Abstract: A method of recovering liquid from an emulsion formed in a liquid-liquid contacting process, where the emulsion comprises at least one of the liquids entrapped by water-permeable membranes, comprises: admixing water and the emulsion in a volume ratio of water:emulsion of between 0.5 to 4.5:1 and raising the temperature of the admixture to between about 45.degree. C. and 99.degree. C., to cause water permeation of the membranes, causing them to disintegrate and release liquid, forming a three-phase composition consisting essentially of the two liquids and the solids.

Abstract: A process is described for the recovery of uranium values from phosphoric acid utilizing an alkyl pyrophosphoric acid (APPA) primary extractant. After extracting the uranium from the phosphoric acid, the APPA extractant is deactivated by heating and the uranium values stripped into a phosphoric acid strip solution containing ferric ion as a salting agent. The uranium values may then be re-extracted directly from this stripping solution without adjustment of its concentration into a dialkyl phosphoric acid trialkyl phosphine oxide synergistic extractant from which a relatively pure yellow cake is precipitated. A new procedure for preparing the requisite APPA primary extractants is also disclosed.

Abstract: A metal containing acidic solution, also containing humic acid impurities and contaminants, is purified by contacting the acidic solution with a scrubbing agent consisting essentially of a metal extractant and from about 50 vol.% to about 90 vol.% of a hydrocarbon, where the scrubbing agent interacts with impurities and contaminants in the acidic solution to form a sludge phase which is removed and a purified acid phase which can be fed into a metal recovery process.

Abstract: Electrostatic separation of aqueous and particulate matter from a primarily organic impurities phase which forms at the organic/aqueous interface in a liquid-liquid extraction process.

Abstract: In the process of uranium recovery from wet-process phosphoric acid, post treatment of the raffinate phosphoric acid is very important to prevent damage to the rubber linings of the acid evaporators and to recover the solvent for reuse. A simple and continuous method using an inclined parallel-corrugated-plate separator to remove the organic solvent entrainment from the raffinate phosphoric acid is established herein.

Abstract: A process as described for extracting at least two desired constituents from a mineral using a liquid reagent which produces the constituents, or compounds thereof, in separable form and independently extracting those constituents, or compounds thereof. The process if especially valuable for the extraction of phosphoric acid and metal values, such as uranium from phosphate rock.

Abstract: Process for the recovery of the ruthenium present in an aqueous nitric solution, wherein the ruthenium is extracted in an organic solvent by contacting the nitric solution with an organic phase comprising an organophosphorous compound having at least one electron donor sulphur atom in the presence of a compound able to displace NO.sup.+ ions of the ruthenium complexes present in the nitric solution. Application to the recovery of ruthenium contained in a nitric solution obtained by dissolving irradiated fuel elements.

Abstract: A process for recovering uranium contained in phosphated solutions including the steps of contacting the solution with a diester of pyrophosphoric acid, reextracting the uranium with an alkaline solution, precipitating the uranium, acidifying and agitating the resulting product, separating the organic phase from the aqueous phase and recycling the organic phase to the initial contacting step. The precipitate is separated prior to recycling.

Abstract: A method and apparatus for processing spent nuclear reactor fuel wherein plutonium is continuously contaminated with radioactive fission products and diluted with uranium. Plutonium of sufficient purity to fabricate nuclear weapons cannot be produced by the process or in the disclosed reprocessing plant. Diversion of plutonium is prevented by radiation hazards and ease of detection.

Abstract: Solvent extraction of metallic mineral values from aqueous acidic solutions is improved by introducing a fine dispersion of air bubbles into the acid-extractant mixture to cause solid material normally occurring at the phase interface to float to the surface. This improved process is particularly applicable to the extraction of uranium from wet process phosphoric acid.

Abstract: Methods for control of excessive corrosion in wet process phosphoric circuits are effected by oxidation of reduced ion species in the acid with maintenance of EMF value of the acid above about 190 millivolts (S.C.E. reference) through the digestion circuit. Supplementary monitoring is by a ceric red-ox titration procedure. In one embodied form, the present invention is utilized to control corrosion in a phosphoric acid plant digestion system (and subsequent processing equipment) operating by a dihydrate (gypsum) process even when such a process employs a calcined western U.S. phosphate ore feed from a relatively highly carbonaceous ore. The invention is also applicable for treating wet process phosphoric acid in phosphoric acid plants having a digester system operating by a hemihydrate process.

Abstract: Certain metals, and especially uranium, are recovered from an organic solvent by stripping with an aqueous solution containing polymeric phosphates under conditions in which the polymeric phosphate does not quickly hydrolize and metal preferentially dissolves in the aqueous solution. The metal may then be recovered from the aqueous solution by hydrolizing the polymeric phosphates and either precipitating the metal or reextracting the metal at higher concentrations with an organic solvent containing extractants.

Abstract: A method is described for reducing the volume of radioactive waste produced during the solution mining of uranium and for recovering uranium from it. The recovery leach, which contains uranium in solution and is supersaturated with calcium carbonate, is treated with bicarbonate and made basic which precipitates calcium carbonate and some of the uranium. The precipitated calcium carbonate is dissolved with acid and the uranium in the solution is removed by extraction or precipitation. The remaining solution is contacted with sulfate ions and barium or strontium ions to precipitate BaSO.sub.4.RaSO.sub.4 or SrSO.sub.4.RaSO.sub.4, the principal radioactive constituent in the solid waste product.

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: An improvement in the process for the recovery of uranium from a phosphoric acid solution derived from the acidification of uraniferous phosphate ores including the steps of reducing the uranium in the phosphoric acid solution to the uranous form and contacting the resulting solution with a liquid-liquid solvent extraction agent which is a mixture of mono- and di-(alkyl-phenyl) esters of orthophosphoric acid dissolved in an inert organic diluent, to effect recovery of the uranium with the liquid-liquid solvent extraction agent, the improvement being maintaining the proper ratio of the mono and di-components of the liquid-liquid solvent extraction mixture which comprises removing dissolved and entrained uranium process organics from the raffinate resulting from said liquid-liquid solvent extraction step and recycling the removed organics to said liquid-liquid solvent extraction step.

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: In uranium recovery by solvent extraction processes, uranium is separated from iron in an organic extract by stripping the extract with a dilute aqueous solution containing carbonate ions, hydroxyl ions, and ammonium or alkali metal ions. The iron precipitates during stripping as Fe.sub.2 O.sub.3.nH.sub.2 O which is filtered off. The uranium is then precipitated from the filtrate as UO.sub.3.2H.sub.2 O by steam stripping or as ammonium diuranate by lowering the pH to about 2 to decompose the carbonate then raising it with ammonia or ammonium hydroxide to about 7.

Abstract: The invention described herein relates to a method for improving the acid leachability of aluminum and other metal values found in fly ash which comprises sintering the fly ash, prior to acid leaching, with a calcium sulfate-containing composition at a temperature at which the calcium sulfate is retained in said composition during sintering and for a time sufficient to quantitatively convert the aluminum in said fly ash into an acid-leachable form.

Abstract: An emulsion, formed at the interface of an acidic solution and a solvent, is removed by passing at least one continuous belt through the emulsion by floating action, where portions of the emulsion attach to the surface of the belt material.

Abstract: Impure phosphoric acid containing uranium values is pretreated with a water immiscible organic sulfonic acid, preferably in conjunction with a water immiscible organic acid phosphate compound, for extraction of ionic metallic impurities and organic impurities to produce pretreated phosphoric acid containing uranium values. Pretreated phosphoric acid is contacted with a water immiscible extractant comprising an organic uranium-extracting agent, preferably an organophosphorus compound, dissolved in a water immiscible organic diluent to produce uranium-depleted phosphoric acid and uranium-enriched extractant. Uranium is recovered from the uranium-enriched extractant.

Abstract: A process for the reductive stripping of an organic phase containing either or both uranium and vanadium. Said process comprising contacting said organic phase containing either or both uranium in the hexavalent state and vanadium in the pentavalent state with an aqueous solution containing a predetermined quantity of trivalent vanadium to reduce either or both the uranium and vanadium to a lower valence state which is not soluble in said organic phase. The reduced uranium and vanadium separate from the organic phase and collect in the aqueous solution from which it may be subsequently recovered. The process of this invention yields an aqueous strip solution containing fewer contaminates comprising undesirable alkali metals such as sodium and the like than other known chemical stripping processes. Further, the process limits the quantity of ferrous ions present in the system to alleviate the problem of sludge formation in subsequent treatment of the wet process acid to produce phosphate fertilizer.

Abstract: In a process for the recovery of uranium from a wet-process phosphoric acid, comprising treating in an extraction step the preliminarily oxidized acid first with an organic solvent consisting essentially of a dialkylphosphoric acid and a trialkyphosphine oxide dissolved in an inert and unreactive organic solvent whereby there are obtained a uranium-free phosphoric acid and an organic extract consisting essentially of the solvent containing the major portion of uranium; then, in a reextraction step, separating the uranium from the organic extract as ammonium uranyl tricarbonate by reacting the organic extract with ammonium hydroxide and ammonium carbonate; and recycling the uranium-free solvent to the extraction step; an improvement comprises treating the organic extract in a reextraction apparatus having at least two stages, by (a) introducing the extract at the head of the first stage; (b) countercurrently introducing ammonia or ammonium hydroxide solution at the bottom of the first stage; the pH of the firs

Abstract: During the reprocessing of irradiated nuclear fuel by solvent extraction techniques a primary separation to give a uranium containing product stream and a plutonium containing product stream occurs. The plutonium in the plutonium containing stream is separated from neptunium and uranium by bringing a solution containing plutonium, neptunium and uranium in an organic solvent into contact first with an aqueous solution of a hydroxylamine and/or a hydrazine salt at 30.degree. to 35.degree. C. to preferentially reduce the neptunium and to extract it into the aqueous phase and then bringing the organic solution containing plutonium and uranium into contact with an aqueous phase containing a hydroxylamine and a hydrazine salt at about 50.degree. C. to preferentially reduce the plutonium and to extract it into the aqueous phase leaving the uranium in the organic solvent.

Abstract: A process for the simultaneous coextraction of uranium and vanadium from an aqueous acidic solution containing the same comprising contacting said aqueous acidic solution with an organic phase comprising a mixture of (a) an alkyl substituted diaryl phosphoric acid, (b) a trialkylphosphine oxide compound, and (c) an organic diluent whereby the uranium and vanadium are coextracted into the organic phase and thereafter recovering the uranium and vanadium from the organic phase.

Abstract: A process for rendering actinide values recoverable from sodium carbonate scrub waste solutions containing these and other values along with organic compounds resulting from the radiolytic and hydrolytic degradation of neutral organophosphorous extractants such as tri-n butyl phosphate (TBP) and dihexyl-N,N-diethyl carbamylmethylene phosphonate (DHDECAMP) which have been used in the reprocessing of irradiated nuclear reactor fuels. The scrub waste solution is preferably made acidic with mineral acid, to form a feed solution which is then contacted with a water-immiscible, highly polar organic extractant which selectively extracts the degradation products from the feed solution. The feed solution can then be processed to recover the actinides for storage or recycled back into the high-level waste process stream. The extractant is recycled after stripping the degradation products with a neutral sodium carbonate solution.

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

Abstract: A method for handling the problem of crud formation in the solvent extraction of wet-process phosphoric acid is described. Crud is the name ordinarily given to the thick interfacial layer of semi-floatable material which tends to form and accumulate, and which interferes with many of the extraction processes and, in particular, with the solvent extraction of wet-process phosphoric acid. The method described provides for the operation of the solvent extraction stages in a manner that allows the formation and temporary accumulation of crud in those stages whereupon the solvent and the crud are then removed from the extraction stages and treated in a series of operations which include a clarification step, an acid removal step, a water wash step and a caustic treatment step, which steps effectively separate the solvent from the crud, and remove the crud components from the system and regenerate otherwise unusable solvent.

Abstract: Uranium oxide is recovered from an aqueous solution of uranyl halides by extracting uranyl halide from the aqueous solution with an organic liquid, forming a crown ether uranyl halide complex in the organic liquid, and then contacting the uranyl halide crown ether complex with water, carboxylate ion, and light under suitable conditions. Also disclosed are certain novel crown ether uranyl halide complexes and their preparation. Also disclosed is the use of 18-crown-6 to selectively recover uranyl halide from a solution thereof containing other metal salts.

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: A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values.

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

Abstract: The rate of reduction of Pu(IV) to Pu(III) in nitric acid solution containing a reducing agent is enhanced by exposing the solution to 200-500 nm electromagnetic radiation. Pu values are recovered from an organic extractant solution containing Pu(IV) values and U(VI) values by the method of contacting the extractant solution with an aqueous nitric acid solution in the presence of a reducing agent and exposing the aqueous solution to electromagnetic radiation having a wavelength of 200-500 nm. Under these conditions, Pu values preferentially distribute to the aqueous phase and U values preferentially distribute to the organic phase.

Abstract: Process for the recovery of uranium from wet process phosphoric acid solution in which an organic extractant, containing uranium values and dissolved iron impurities and comprising a dialkylphosphoric acid and a trialkylphosphine oxide dissolved in a water immiscible organic solvent, is contacted with a substantially iron-free dilute aqueous phosphoric acid to remove said iron impurities. The removed impurities are bled from the system by feeding the resulting iron-loaded phosphoric acid to a secondary countercurrent uranium extraction operation from which they leave as part of the uranium-depleted acid raffinate. Also, process for recovering uranium in which the extractant, after it has been stripped of uranium values by aqueous ammonium carbonate, is contacted with a dilute aqueous acid selected from the group consisting of H.sub.2 SO.sub.4, HCl, HNO.sub.3 and iron-free H.sub.3 PO.sub.4 to improve the extraction efficiency of the organic extractant.

Abstract: Uranyl ion in solution in tri-n-butyl phosphate is readily photochemically reduced to U(IV). The product U(IV) may effectively be used in the Purex process for treating spent nuclear fuels to reduce Pu(IV) to Pu(III). The Pu(III) is readily separated from uranium in solution in the tri-n-butyl phosphate by an aqueous strip.

Abstract: A method is provided for increasing the lifetime of an extraction medium containing an organophosphorus acid ester and a hydrocarbon and being used for reprocessing spent nuclear fuel and/or breeder materials. Impurities resulting from chemical and/or radiolytic decomposition and interfering compounds of such impurities with radionuclides are removed from the extraction medium by bringing the extraction medium, after use, into intimate contact with an aqueous hydrazine hydrate solution having a concentration of between about 0.1 molar and about 1.0 molar at a temperature between about 20.degree. C to about 75.degree. C. The aqueous hydrazine hydrate solution is then separated from the extraction medium.

Abstract: A method of preparation for non-polluting storage of liquids containing phosphoric acid ester and hydrocarbons used in the reprocessing of spent nuclear fissile and/or fertile materials. The hydrocarbons are recycled for the production of fresh phosphoric ester-hydrocarbon solutions and the residue is brought into a form which can be fixed.

Abstract: A process for chemical enrichment of uranium with respect to a lighter one of its isotopes. The process consists in contacting uranium of valence state III and uranium of valence state IV, or a compound of uranium of valence state III and a different compound of uranium of valence state III. One of the phases which are contacted or the only phase is liquid. The system should be substantially free of elements which would cause uranium III to oxidize to valence IV.

Abstract: A process for separately recovering uranium, plutonium and neptunium substantially free of fission products from irradiated nuclear fuel is presented in which the fuel is dissolved in a strong mineral acid forming an aqueous dissolved nuclear fuel solution and treated to separate the uranium, plutonium and neptunium therefrom substantially free of said fission products by the sequential steps of solvent extraction, ion exchange and fluorination. The process has an improvement comprising the addition of a sufficient quantity of an additive of a stable metallic complex to the aqueous dissolved nuclear fuel solution prior to solvent extraction. This achieves improved purity of the separated uranium, plutonium and neptunium.

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: A liquid-liquid extraction process for the recovery and partitioning of actinide values from acidic nuclear waste aqueous solutions, the actinide values including trivalent, tetravalent and hexavalent oxidation states is provided and includes the steps of contacting the aqueous solution with a bidentate organophosphorous extractant to extract essentially all of the actinide values into the organic phase. Thereafter the respective actinide fractions are selectively partitioned into separate aqueous solutions by contact with dilute nitric or nitric-hydrofluoric acid solutions. The hexavalent uranium is finally removed from the organic phase by contact with a dilute sodium carbonate solution.

Abstract: A method of regenerating fuel and/or breeder elements with thorium as brer material for nuclear reactors. The fuel and/or breeder material is dissolved in an acid such as nitric acid, sulfuric acid or the like and the thus obtained solution is subjected to a liquid-liquid reaction with an extraction substance such as tributyl phosphate dissolved in an organic solvent. The fuel material and breeder material are enriched in the organic phase and the fission products remain in the aqueous phase, whereafter the fuel material and/or breeder material are converted into oxides, carbides etc. Prior to the above mentioned liquid-liquid extraction, the acid solution containing the fuel material and/or breeder material is passed over a sorption column in such a way that the number of mols of adsorbed .sup.233 Pa per time unit equals the number of mols of .sup.233 Pa disintegrating per time unit in the sorption column to form .sup.233 U.

Abstract: Ferric ions are added into the aqueous feed of a plutonium scrap recovery process that employs a tributyl phosphate extractant. Radiolytic degradation products of tributyl phosphate such as dibutyl phosphate form a solid precipitate with iron and are removed from the extraction stages via the waste stream. Consequently, the solvent extraction characteristics are improved, particularly in respect to minimizing the formation of nonstrippable plutonium complexes in the stripping stages. The method is expected to be also applicable to the partitioning of plutonium and uranium in a scrap recovery process.

Abstract: The method of selective stripping from plutonium-loaded organic solvents which may also contain uranium consists in passing the organic solvent in countercurrent contact with an aqueous solution of a salt of hydroxylamine and formic acid, the concentration of formic acid in the reaction medium being within the range of 1M to 5M.The method finds an application in the stripping of plutonium from trilaurylamine diluted in an aromatic hydrocarbon such as tert-butyl benzene or in the separation of uranium and plutonium contained in tributylphosphate diluted in an aliphatic hydrocarbon such as dodecane.

Abstract: The method of selective stripping from plutonium-loaded organic solvents which may also contain uranium consists in bringing the initial organic solvent into countercurrent contact with a solution containing a reducing agent consisting of an aromatic organic compound selected from the group comprising the substituted hydrazines, the polyamines, the aminophenols, and an agent for the removal of the nitrites which are formed.One field in which the method finds an application is the extraction of uranium and plutonium from nitric acid solutions derived from the reprocessing of irradiated nuclear fuels.

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: Uranium is separated from contaminating metal ions in an aqueous feed liquor containing the uranyl ion. The liquor is extracted with a first, non-interfering, waterimmiscible, organic solvent containing a reagent which reacts with the uranyl ion to form a complex soluble in the organic solvent. The organic solvent is scrubbed with water if necessary, then stripped with a stripping liquor of an aqueous sulfuric acid liquor having a pH of about 0.5 to about 6 containing a reducing ion or an aqueous carbonate solution having a pH of about 8 to about 9. If the sulfuric acid liquor is used the stripped uranous ion is oxidized and the sulfuric acid liquor is diluted to prevent the precipitation of a uranium complex. The stripping liquor is extracted with an amine liquor comprising a second, noninterfering, water-immiscible, organic solvent and a tertiary or quaternary amine. The amine liquor is stripped with an ammonium carbonate solution to precipitate a uranium complex.

Abstract: An improved Purex wet recovery process including the step of extracting and separating uranium and plutonium simultaneously from the fission products in the presence of nitric acid and nitrous acid by using a multistage extractor unit having an extracting section and a washing section is provided for separating and recovering neptunium simultaneously with uranium and plutonium contained in spent nuclear fuel. The improved method comprises the steps of maintaining the nitrous acid concentration in said extracting section at a level suited for effecting oxidation of neptunium from (V) to (VI) valence, while lowering the nitrous acid concentration in said washing section so as to suppress reduction of neptunium from (VI) to (V) valence, and maintaining the nitric acid concentration in said washing section at a high level.

Abstract: A process for the treatment of irradiated fissile material in solution in nitric acid in which tributylphosphate diluted to less than 7% by volume with an inert diluent is contacted with the nitric acid solution and backwashed with sulphuric acid.

Abstract: Fast breeder fuel elements which have been highly burnt-up are reprocessed by extracting uranium and plutonium into an organic solution containing tributyl phosphate. The tributyl phosphate degenerates at least partially into dibutyl phosphate and monobutyl phosphate, which form stable complexes with tetravalent plutonium in the organic solution. This tetravalent plutonium is released from its complexed state and stripped into aqueous phase by contacting the organic solution with an aqueous phase containing tetravalent uranium.

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 metallic reducing agent such as iron, aluminum or zinc, and then a miscible solvent such as methanol. The precipitated solids, which are separated from the purified phosphoric acid, consist of a mixture of metal phosphates and uranium. This solid is dissolved in acid and the uranium recovered from the solution by liquid-liquid solvent extraction. The miscible solvent and some water are distilled away from the purified phosphoric acid. The distillate is rectified, the water discarded, and the miscible solvent recovered for recycle.When the miscible solvent is methanol, the optimum ranges of ammonia and methanol are 0.05 to 0.20 gram atom nitrogen per gram atom phosphorus and 1.93 to 3.15 pounds methanol per pound of orthophosphoric acid. The amount of reducing agent added should be sufficient to reduce all uraniun to U(IV).