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: A method of substantially removing a trace metallic contaminant from a liquid containing the same comprises,adding an oxidizing agent to a liquid containing a trace amount of a metallic contaminant of a concentration of up to about 10.sup.-1 ppm, the oxidizing agent being one which oxidizes the contaminant to form an oxidized product which is insoluble in the liquid and precipitates therefrom, and the conditions of the addition being selected to ensure that the precipitation of the oxidized product is homogeneous, andseparating the homogeneously precipitated product from the liquid.

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 method of purifying uraniferous aqueous solutions containing impurities, particularly at least one of the elements zirconium and/or hafnium, and also containing at least one of the anions SO.sub.4.sup.=, NO.sub.3.sup.-, Cl.sup.- or F.sup.- acting as complexing agents for the uranium and impurities, by precipitation of said impurities by means of an alkaline agent, characterized in that the following stages are carried out for the purpose of selectively and quantitatively separating the impurities from the uranium:(a) said uraniferous solution is brought to a temperature of at least 40.degree. C.,(b) the content of complexing agent is adjusted to bring the molar ratio: [complexing agent]/[soluble U+Zr and/or Hf] to at least 3,(c) the pH of said solution is adjusted between 2.2 and 4.3 by introducing an alkaline agent.

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: A process for treating a radioactive liquid waste is disclosed, in which a radioactive liquid waste containing uranium and .beta.-decay nuclides, daughter nuclides of uranium, is treated by combination of a flocculation method using water glass as pretreatment and a subsequent ion exchange method. An approximately total amount of the uranium and a part of the .beta.-decay nuclides, daughter nuclides of uranium, in the liquid waste are captured by an amorphous silica precipitate formed by addition of the water glass and a remaining part of the .beta.-decay nuclides, daughter nuclides of uranium, is captured thereafter by the ion exchange treatment. The thus captured radioactive materials are respectively eluted from the filtered out precipitate, a radioactive solid waste, and the ion exchanger by acid treatment to be recovered as an acidic solution. Thus, the radioactive materials in the liquid waste are recovered approximately completely, thereby making the radioactivities of a final drain remarkably reduced.

Abstract: A process is described for the recovery of tetravalent uranium from wet-process phosphoric acid utilizing an alkyl pyrophosphoric acid extractant or the like. After extracting the tetravalent uranium from wet-process acid, iron is stripped from the pregnant extractant into an oxalic acid stripping solution and then the tetravalent uranium is stripped from the pregnant extractant into an oxalate stripping solution. The oxalate stripping solution is an aqueous solution of an alkali metal or ammonium oxalate. The barren extractant is recycled for contacting with fresh wet-process acid. The uranium is oxidized and then precipitated in the oxalate stripping solution. The precipitated solids are separated from the solution, and the uranium is dried to a UO.sub.3 product.

Abstract: Disclosed is a process for the selective separation of uranium from metals accompanying it in a uranium-containing ore, comprising the steps of preparing a uranium-containing solution; adding to the solution (i) hydrochloric acid in an amount sufficient to form complex anions of the type(UO.sub.2 Cl.sub.n).sup.n-2wherein n is 3 or 4, or (ii) sulfuric acid in an amount sufficient to form complex anions of the typeUO.sub.2 (SO.sub.4).sub.m.sup.2-2mwherein m is 2 or 3; adding to the solution a cationic surface active agent which forms a relating insoluble precipitate with the complex anion; subjecting the solution containing said precipitate to a gas flotation step; separating the foam fraction from the liquid fraction; and recovering uranium from the foam fraction. Also disclosed is an apparatus for carrying out the process.

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 is described for the removal of uranium values from contaminated wet-process phosphoric acid solutions using an addition thereto of first an aliphatic ketone and then an inorganic fluoride, whereby high efficiencies in the removal of uranium as a solid, filterable precipitate are achieved.

Abstract: A process for recovering one or more nonradioactive transition metal compounds from an ore containing one or more compounds of said transition metal or metals and further containing at least one complex of a member selected from the group consisting of uranium, thorium, radium, titanium, and rare earth metals, which comprises decomposing said ore in crushed condition by means of an acid so that a portion of the ore is brought into solution in a liquid phase and another portion of the ore remains in a solid phase, said compound or compounds of the transition metal or metals to be recovered passing into only the liquid or into only the solid phase, the uranium in the crushed ore being treated so as to cause substantially all of said uranium to be present in an oxidation state in which it cannot, during the decomposition step, pass into the phase containing the transition metal compound or compounds.

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

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

Abstract: An economically advantageous method of recovering uranium from a wet process phosphoric acid solution through the steps of making hemihydrate gypsum contact with the acid solution thereby transferring uranium from the acid solution into the gypsum, dispersing the U-containing gypsum separated from the acid solution in water to convert the gypsum to dihydrate accompanied by the transfer of uranium into water, separating the obtained U-containing aqueous solution from the dihydrate gypsum, and adding precipitant such as an inorganic base to the aqueous solution to form a precipitate comprising an insoluble uranium compound. The contact of hemihydrate gypsum with the phosphoric acid solution is preferably preceded by reduction of hexavalent uranium in the acid solution to tetravalent uranium, and can be achieved either by adding hemihydrate gypsum to the acid solution or by converting dihydrate gypsum to hemihydrate within the acid solution preferably preceded by the addition of sulfuric acid.

Abstract: In an ion exchange process for recovering valuable components from a weak feed solution containing impurities, an improved loading of the desired component on the resin and a purer and more concentrated eluate can be obtained by subjecting the resin to a conditioning step after the step of exhaustion of the feed solution onto the resin and before the resin elution step. In the conditioning step, a predetermined isolated quantity of the loaded resin is contracted with a conditioning solution capable of preferentially removing the unwanted impurity component.

Abstract: In preparing wet process phosphoric acid by decomposing a phosphate rock containing uranium with sulfuric acid and phosphoric acid on condition that hemihydrate gypsum is formed in an acid solution either at the stage of decomposing the phosphate rock or subsequently, uranium contained in the phosphate rock can almost entirely be retained in the obtained phosphoric acid solution by forming the hemihydrate gypsum in the presence of an oxidizing agent, such as a soluble chlorate, hydrogen peroxide or oxygen gas, in the acid solution in a quantity sufficient to render the entire uranium dissolved in the acid solution hexavalent because hemihydrate gypsum adsorbs almost exclusively tetravalent ions of uranium. The uranium is then recovered.

Abstract: A wet process for producing phosphoric acid and phosphohemihydrate by acidulation of phosphate rock. A strong phosphoric acid is obtained with concomitant production of an improved calcium sulfate that is extremely low in radioactivity by operation in Region II for shorter time so as to produce small sized hemihydrate particles, then hydration of a substantial proportion but less than all of the hemihydrate to coarse gypsum substantially free of radioactivity in a dilute acidic aqueous slurry at a temperature and P.sub.2 O.sub.5 concentration selected so as to be within Region I of FIG. 1 and having a substantial weight proportion of coarser gypsum particles that are substantially free of radioactivity.

Abstract: Process for extracting molybdenum from aqueous solutions containing alkali metal carbonate, sulphate, 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, in the form of a concentrate, wherein said solutions are caustified by the addition of a suitable amount of lime, converting the carbonate present into alkali metal hydroxide, a first precipitate is separated from the alkali metal hydroxide-rich liquor, said liquor is concentrated until the alkali metal hydroxide content is at most equal to 50%, to cause the production of a second precipitate formed by alkali metal sulphate and molybdate, which is solubilized in an aqueous recycle liquor, said re-dissolution liquor then being treated with sulphide ions and then acidified to give a pH-v

Abstract: A process for reducing radioactive contamination of phosphogypsum. Phosphogypsum containing radioactive material is calcined to form hemihydrate crystals carrying the radioactive contaminants, and a portion of the crystals is converted to substantially radiation-free gypsum crystals which are readily separated from the hemihydrate crystal relics containing substantially all of the radioactive contamination.

Abstract: A direct precipitation method based on the insolubility of uranyl phosphate in carbonate containing solutions at pH 6-6.5 and its insolubility at pH 8 is described. The method eliminates the use of ion exchange columns for removing uranium values from uranium leachates and can be readily applied to an integrated field operation.

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 method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Abstract: A process of two crystallization stages for reducing radioactive contamination of phosphogypsum is disclosed. In the process anhydrite crystals are obtained through dehydration of the radiation containing phosphogypsum in strong sulfuric acid; a portion of the anhydrite crystals containing the radioactive contamination is converted to substantially radiation free gypsum by crystallizing out on a large solids concentration of radiation free gypsum seed crystals; and coarse radiation free gypsum crystals are separated from small anhydrite crystal relics containing substantially all of the radioactive contamination.

Abstract: A process for removing dissolved radioactive materials from aqueous solution by incorporating lime in a sufficient dosage to adjust the pH of the water to greater than 11.0 while simultaneously adding a material from the group consisting of hydrogen peroxide, oxygen, ferrous sulfate, ferric chloride and potassium permanganate.

Abstract: In a process for converting UF.sub.6 into UO.sub.2, the UF.sub.6 is brought into contact with an aqueous aluminum nitrate solution. The resultant product is solvent extracted with tributyl phosphate to remove uranyl nitrate. The raffinate has a fluorine/aluminum (F/Al) weight ratio within the range of from about 0.5 to about 1.2. A sufficient quantity of hydrofluoric acid is added to the raffinate to minimize the solubility of aluminum fluoride (AlF.sub.3) therein and thereby maximize the precipitation potential of AlF.sub.3. Generally this occurs when sufficient hydrofluoric acid has been added to cause the F/Al weight ratio to be within the range of from about 1.8 to about 2.2. As a result of this treatment, the raffinate is divided into an uranium-containing aqueous solution and an AlF.sub.3 precipitate which contains substantially no uranium.

Abstract: A process for removing arsenic from an aqueous solution containing arsenic in the form of a soluble arsenate includes adjusting the pH of the solution if necessary to at least about 10, adding a soluble barium salt to precipitate arsenic as barium arsenate, and removing the precipitated barium arsenate from the solution.

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

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

Abstract: A method of accelerating and increasing the solubility in nitric acid of plutonium dioxide and uranium/plutonium oxide fuels and insoluble residues from such nitric acid dissolution, includes the steps of arranging metal loadings to be at least 200 g (Pu) or (U+Pu) per liter of dissolving solution so that the plutonium in solution acts as an autocatalyst.

Abstract: Zirconium is selectively precipitated from aqueous solutions containing zirconium and uranium by treatment with a precipitant consisting of tartaric acid or a tartrate.

Abstract: Phosphoric acid and gypsum are produced from phosphate rock by digesting same with sulfuric acid to produce a high concentration of phosphoric acid and calcium sulfate hemihydrate. The solubilized uranium in the rock is rendered recoverable by reducing same from the +6 valence state to the +4 valence state during the digestion step followed by oxidizing back to the +6 valence state in a subsequent step.

Abstract: A cyclic process for the treatment of a uraniferous liquor containing, in addition to the desired metal, sodium carbonate and/or bicarbonate and sodium sulphate, obtained by the alkaline attack of a sulphur-containing uraniferous ore, the said process comprising fixing the uranium over a first strong anion-exchange resin, eluting it with an ammonium carbonate and/or bicarbonate solution in the form of an ammonium uranyl tricarbonate which is subsequently decomplexed, precipitating and calcinating the precipitated uranates and/or diuranates with regeneration of the eluant solution of ammonium carbonate and/or bicarbonate, passing the liquor which is freed from uranium and leaves the first strong anion-exchange resin into a neutralization zone then over a second anion-exchange resin, eliminating the SO.sub.4.sup..dbd. ions which are subsequently precipitated in calcium form, recycling a carbonate and/or bicarbonated liquor which is free from SO.sub.4.sup..dbd.

Abstract: A cyclical and automatable process and apparatus for continuously removing undesired components from solid particles by means of a solvent, the process including maintaining an expanded bed of solid particles in a vertical column, supplying a suitable solvent to the bed to flow through the bed either upwardly or downwardly in a selectably alternating manner, supplying solid particles to the top of the column while the solvent flows upwardly therethrough and interrupting the supply of solid particles and removing a bottom portion of the bed by the solvent when it flows downwardly.

Abstract: Conversion of plutonium nitrate containing americium as a contaminant into plutonium dioxide while concurrently preventing americium contaminant from passing into the plutonium dioxide product. Discontinuous precipitation of plutonium oxalate is effected by maintaining the Pu nitrate-nitric acid solution containing americium in a reaction vessel at 60.degree.-95.degree. C., adding granular solid oxalic acid or solid ammonium oxalate in freeflowing form substantially free of clusters and lumps in stoichiometric deficiency of up to 7% of oxalic acid for reaction with the Pu in the solution; separating resultant Pu oxalate precipitate into which at least a portion of the americium contaminant has been prevented from entering, and calcining the purified Pu oxalate precipitate to PuO.sub.2.

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: Molybdenum is recovered from an ammonium molybdate solution containing phosphate anions by digesting in an ammonium molybdate solution at least one water-soluble compound of at least one metal selected from the group consisting of aluminum, calcium, iron and magnesium in small but effective amounts to precipitate at least about 50% of the phosphate anions for a time sufficient to precipitate the phosphate anion, separating the phosphate precipitate from the ammonium molybdate solution, then acidifying the ammonium molybdate solution with at least one mineral acid selected from the group consisting of sulfuric acid and nitric acid to lower the pH value of the solution to between about 2.5 and about 4.5 to precipitate ammonium polymolybdate, and calcining the ammonium polymolybdate at a temperature below about 750.degree. C. to produce a molybdenum trioxide product.

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

Abstract: 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: A process is disclosed for purifying a composition comprising at least one rare-earth element. The process is useful for removing radioactive impurities from these elements. The process comprises the steps of:(1) forming an aqueous solution of the composition to be purified having a pH between 4.0 and 6.5;(2) precipitating a portion of the rare earth and the radioactive impurities by adding to the solution, while maintaining the pH between 4.0 and 6.5;(i) a soluble barium or strontium salt;(ii) a soluble aluminum, magnesium or zirconium salt;(iii) a selective precipitating agent; and(iv) a sulfate ion-providing salt.The preferred selective precipitating agent is 8-hydroxyquinoline. After precipitation, the process is completed by the steps of:(3) separating the solution from the precipitate and(4) recovering purified rare earth from the solution.The process produces highly purified rare earth which can be used to form phosphors for X-ray screens.

Abstract: Poly[di-1,2-(diazinylidene)-ethene-1,2-diols] having the structure:X--C.sub.4 H.sub.2 N.sub.2 --C(OH).dbd.C(OH)--C.sub.4 H.sub.2 N.sub.2 --C(OH) ].sub.n C(OH)--C.sub.4 H.sub.2 N.sub.2 --Yin which the C.sub.4 H.sub.2 N.sub.2 group is a diazinylidene diradical having each C(OH) group adjacent (ortho) to one but not two ring nitrogen atoms and not adjacent (ortho) to each other; X and Y are terminal formyl (CHO) groups or groups derived from them by oxidation; and n has a value of zero to seven--are prepared by a cyanide ion catalyzed condensation of pyrazine-2,5-dicarboxaldehyde, pyridazine-3,6-dicarboxaldehyde; or pyrimidine-4,6-dicarboxaldehyde as the only aldehydes which provide the requisite structures. The polymers form complexes with metal ions including manganese, which is insoluble, and vanadium, which is soluble.

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: 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: This invention is a simple and effective method for removing uranium from aqueous HF solutions containing trace quantities of the same. The method comprises contacting the solution with particulate calcium fluoride to form uranium-bearing particulates, permitting the particulates to settle, and separting the solution from the settled particulates. The CaF.sub.2 is selected to have a nitrogen surface area in a selected range and is employed in an amount providing a calcium fluoride/uranium weight ratio in a selected range. As applied to dilute HF solutions containing 120 ppm uranium, the method removes at least 92% of the uranium, without introducing contaminants to the product solution.

Abstract: An actinide nitride-fueled nuclear reactor and a method of operation therefor, including continuous in situ removal of fission products and optional addition of fuel-forming actinide material as the reaction proceeds. The reactor employed has a fuel system comprising a critical mass of a nitride of an actinide metal in contact with a non-critical solution of the actinide metal in a molten metal solvent of low neutron adsorption cross section such as tin, said fuel system being maintained under a nitrogen atmosphere in an inert, refractory vessel such as graphite which is non-conducive to the formation of actinide oxides. Fission products formed are continuously exchanged with the actinide metal dissolved in the molten metal solvent as the nuclear reaction proceeds, with equivalent amounts of actinide nitride being formed and precipitated into the critical mass as fission products are dissolved in the molten metal solution.

Abstract: A process for selectively removing tetravalent and/or hexavalent uranium ions from a solution by using a chelate-forming cation exchanger. The solution may be prepared by contacting uranium bearing ore with either an acid or a dilute carbonate or bicarbonate solution. To facilitate removal of the uranium ions, the solution may have a pH value greater than 1.0 and a redox potential of from -50 to +250 mV.

Abstract: This invention is a process for decomposing ammonium nitrate and/or selected metal nitrates in an aqueous solution at an elevated temperature and pressure. Where the compound to be decomposed is a metal nitrate (e.g., a nuclear-fuel metal nitrate), a hydroxylated organic reducing agent therefor is provided in the solution. In accordance with the invention, an effective proportion of both nitromethane and nitric acid is incorporated in the solution to accelerate decomposition of the ammonium nitrate and/or selected metal nitrate. As a result, decomposition can be effected at significantly lower temperatures and pressures, permitting the use of system components composed of off-the-shelf materials, such as stainless steel, rather than more costly materials of construction. Preferably, the process is conducted on a continuous basis. Fluid can be automatically vented from the reaction zone as required to maintain the operating temperature at a moderate value--e.g., at a value in the range of from about 130.

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 concentrating and recovering uranium material from an aqueous solution, comprises passing a feed solution containing uranium through at least one reverse osmosis membrane system to concentrate the uranium, and then flushing the concentrated uranium solution with water in a reverse osmosis membrane system to further concentrate the uranium.