Abstract: A method for the separation of a mixture of fine particles wherein the fraction to be separated has a surface which is more reactive chemically than the remaining fractions by grinding the mixture so that the fraction to be separated has a particle size of less than 50.mu., suspending the ground mixture at a temperature less than 100.degree. C. in a diluted aqueous solution of zirconium oxychloride or zirconium oxynitrate, allowing the solution to settle and decanting the deposited material and then adding a base to the decantate to precipitate the residue therein. This process allows selective separation of the fractions having a more chemically reactive surface without the use of complex equipment and with a minimum use of expensive additives and chemicals.

Abstract: Radiation-contaminated ammonium nitrate is heated in solution to about 100.degree. C. in the presence of finely powdered calcium oxide or lithium hydroxide. Ammonia and water vapor are given off leaving an alkaline or alkaline earth nitrate which can then be safely decomposed by calcination into a metal oxide and oxides of nitrogen. The metal oxide can be recycled in a continuation of the process. The oxides of nitrogen can be passed through water to produce nitric acid useable in dissolving oxides of fissionable materials and the ammonia may be used in aqueous solution to react with nitrates of nuclear fuel or breeder metals in the very process that produces the by-product ammonium nitrate. Thus, all by-products and reagents can be reconverted and recycled.

Abstract: 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: The addition of a normally insoluble fluoride to a reaction being a chemical treatment process provides a surprisingly improved rate is dissolution in many cases. Specifically, where the fluoride is added to a reaction using hydrochloric acid or ferric chloride, the yield is more rapid than has previously been shown in the art. The invention is most applicable where the stability constant of the fluoride or fluoride complex of the element to be recovered is higher than that of the cation with which the fluoride is originally associated.

Abstract: Uranium oxide hydrate is produced by irradiating with light a solution of a suitable diluent, water-soluble uranium salt, carboxylate ion, and a rate-promoting amount of at least one suitable crown ether.

Abstract: UO.sub.2 for nuclear fuel is made from UF.sub.6. The method involves injecting UF.sub.6, with or without a nitrogen carrier, into a solution containing 1) an inert reaction medium, 2) water, 3) a Lewis base. The precipitate from the above reaction is then reduced in H.sub.2 at a temperature below 750.degree. C. to give ceramic grade UO.sub.2.

Abstract: A process for producing UO.sub.2 F.sub.2 from a soluble uranyl salt. The uranyl salt is combined with a soluble fluoride salt in a solvent to form a reaction solution. The solvent exhibits Lewis base characteristics. The reaction product is a crystalline solid which is separated from the reaction solution. The UO.sub.2 F.sub.2 may then be obtained from the crystalline solid.

Abstract: A method is disclosed for the manufacture of phosphoric acid directly from phosphate rock wherein the crushed phosphate rock is mixed with dilute phosphoric acid to form a slurry and the slurry is then heated to produce calcium monophosphate. Thereafter, oxalic acid is added to the slurry to precipitate the calcium therein as calcium oxalate which is separated. The liquid resulting therefrom which contains the phosphoric acid from the rock is then treated conventionally to recover the phosphoric acid therein.

Abstract: A nitric acid solution having a nitric acid concentration within the range of 0.01 to 15 M and containing plutonium(IV) ions and/or plutonium(III) ions is reacted with a formic acid solution in order to obtain a precipitate of plutonium(III) formiate.

Abstract: In the recovery of plutonium from irradiated nuclear fuel elements especially those coming from fast nuclear reactors, the improvement consisting in that the spent nuclear fuel elements are subjected to the action of a molten nitrate bath (mixed nitrates of alkali metals or alkaline earth metals), whereafter the plutonium thus obtained is further decomposed in a nitrate bath at a higher temperature, then the plutonium is recovered and a further thermal decomposition of the remaining material at a still higher temperature enables the uranium to be recovered. The recovery of plutonium requires the action of the nitric vapors, the recovery of uranium does not. Molten nitrates can likewise be recovered and recycled.

Abstract: Metal powders, metal oxide powders, and mixtures thereof of controlled particle size are provided by reacting an aqueous solution containing dissolved metal values with excess urea. Upon heating, urea reacts with water from the solution leaving a molten urea solution containing the metal values. The molten urea solution is heated to above about 180.degree. C. whereupon metal values precipitate homogeneously as a powder. The powder is reduced to metal or calcined to form oxide particles. One or more metal oxides in a mixture can be selectively reduced to produce metal particles or a mixture of metal and metal oxide particles.

Abstract: There is provided a process for continuous preparation of uranium tetrafluoride hydrate which comprises the steps of continuously feeding uranous solution and hydrofluoric acid into the lower section of a reaction vessel to produce crystal particles of uranium tetrafluoride hydrate causing the crystal particles to float up and be suspended in the upper section of the reaction vessel by agitation; in which section the crystal particles grow and then precipitating and discharging the thus grown crystal particles from the bottom of the vessel, while causing waste solution to overflow from the top of the vessel. There is also provided an apparatus to accomplish the aforementioned process.

Abstract: A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

Abstract: The reduction of UF.sub.5 to UF.sub.4 in a molten fluoride salt by sparging with hydrogen is catalyzed by metallic platinum. The reaction is also catalyzed by platinum alloyed with gold reaction equipment.

Abstract: An actinide dioxide, e.g. uranium dioxide, is prepared by reacting an actinide nitrate or hydrate or tetrahydrofuranate thereof, e.g. uranyl nitrate, a hydrate of uranyl nitrate or a tetrahydrofuranate of uranyl nitrate with an alkali or alkaline earth metal adduct of a monocyclic or polycyclic hydrocarbon in the presence of an inert organic solvent. Typically, the starting material may be uranyl nitrate dihydrate or uranyl nitrate ditetrahydrofuranate (the latter material is a novel composition of matter) with a reactant such as the sodium adduct of naphthalene in the presence of a solvent such as tetrahydrofuran. The resultant uranium dioxide may be further purified by heating it in the presence of hydrogen.

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: Salts or materials containing plutonium and americium are dissolved in hydrochloric acid, heated, and contacted with an alkali metal carbonate solution to precipitate plutonium and americium carbonates which are thereafter readily separable from the solution.

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

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

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

Abstract: A method of recovering uranium hexafluoride from gaseous mixtures employing as an absorbent a liquid composition at least one of the components of which is chosen from the group consisting of ethanolamine, diethanolamine, and 3-methyl-3-amino-propane-diol-1,2.

Abstract: A method for recovering substantially all of the fluorine and uranium values and at least 90 percent of the rare earth metal values from brine raffinate obtained as by-product in the production of phosphoric acid by the hydrochloric acid decomposition of tricalcium phosphate minerals. A basically reacting compound is added to the brine raffinate to effect a pH of at least about 9, whereby fluorine, uranium and rare earth metal values are simultaneously precipitated therefrom. These values may then be separately recovered from the precipitate by known processes.

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).