Abstract: A method and system are provided for the creation of vertical and horizontal freeze wells, in a dome-like pattern around the ore body, as a hydraulic barrier to ensure the ISR mining solution and the mined minerals do not flow out of the ore body. A method to formulate a suitable mining solution used for ISR mining, where the lixivant does not freeze when using the freeze dome containment method and where the resulting PLS has a high concentration of dissolved minerals and thus eliminates the need for the solvent extraction/ion exchange step during processing is also described.

Abstract: Disclosed are a method and a device for recovering uranium (U) using a process for chemically treating washing wastewater of a uranium hexafluoride (UF6) cylinder. The method and the device are provided to separate uranium (U) from the wastewater released during a process of washing the uranium hexafluoride (UF6) cylinder and to release a filtrate that satisfies atomic energy licensing standards and environmental regulation standards using evaporation and condensation. Accordingly, an independent technology and process for treating the wastewater released during the process of washing the uranium hexafluoride (UF6) cylinder are ensured, which provides easier maintenance and greatly reduces costs compared to the purchase and operation of apparatuses manufactured by foreign makers.

Abstract: Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

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

Abstract: Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Abstract: Disclosed is a process for recovery of uranium from a spent nuclear fuel using a carbonate solution, characterized by excellent proliferation resistance of preventing leaching of transuranium element (TRU) nuclides such as Pu, Np, Am, Cm, etc. from the spent nuclear fuel as well as environmental friendliness of minimizing waste generation, wherein a highly alkaline carbonate solution is used to separate uranium alone from the spent nuclear fuel.

Abstract: Disclosed is a process for recovery of uranium from a spent nuclear fuel using a carbonate solution, characterized by excellent proliferation resistance of preventing leaching of transuranium element (TRU) nuclides such as Pu, Np, Am, Cm, etc. from the spent nuclear fuel as well as environmental friendliness of minimizing waste generation, wherein a highly alkaline carbonate solution is used to separate uranium alone from the spent nuclear fuel.

Abstract: A two-cycle countercurrent extraction process for recovery of highly pure uranium from fertilizer grade weak phosphoric acid. The proposed process uses selective extraction using di-(2-ethyl hexyl) phosphoric acid (D2EHPA) and tri-n-butyl phosphate (TBP) with refined kerosene as synergistic extractant system on hydrogen peroxide treated phosphoric acid, and stripping the loaded extract with strong phosphoric acid containing metallic iron to lower redox potential. The loaded-stripped acid is diluted with water back to weak phosphoric acid state and its redox potential raised by adding hydrogen peroxide and re-extracted with same extractant system. This extract is first scrubbed with sulfuric acid and then stripped with alkali carbonate separating iron as a precipitate, treated with sodium hydroxide precipitating sodium uranate, which is re-dissolved in sulfuric acid and converted with hydrogen peroxide to highly pure yellow cake of uranium peroxide.

Abstract: A process for treating composite materials comprising metals and having significant radioactivity for separation into concentrated individual products is provided. The process of the invention recovers valuable metals and semi-metal elements including rare earth, transition metal, radioactive elements, and compounds and composites thereof as commercially viable products while isolating the radioactive components. They key components in the composite material which are desired to be recovered are tantalum, niobium, and scandium due to their high commercial values and significant quantities. The process further includes the capability to recover uranium, thorium, rare earth, and zirconium products. Generally, the process consists of an initial sulfate roast followed by a series of hydrometallurgical unit operations designed to separate and purify the desired components into commercially usable products, such as tantalum and/or niobium in the form of metal oxides, M.sub.2 O.sub.

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: Process for obtaining fritted oxide pellets of the MxOy type for nuclear fuels from solutions of soluble salts of the element or elements M, involving stage of precipitating the elements M by hydrogen peroxide in an acid medium. During this precipitation, there is an instantaneous dispersion of one of the reagents (solution of salts or peroxide) in the other, in order to obtain a homogenous mixture and an also instantaneous precipitation of the nuclei in a continuous liquid phase confined in an enclosure having minimum dimensions, the mother liquors being rapidly exhausted. The process makes it possible to obtain sintered pellets whose density exceeds 96% of the theoretical density.

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

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

Abstract: A method and apparatus are provided for producing uranyl peroxide (UO.sub.4.2H.sub.2 O) from uranium-bearing metal pieces. The uranium-bearing metal pieces are dissolved in a first aqueous solution containing nitric acid and between 0.5% and 5.0% fluoboric acid to provide a second aqueous solution which includes uranyl ions (UO.sub.2.sup.+2) and nitric and fluoboric acids. Hydrogen peroxide is added to the second aqueous solution to precipitate uranyl peroxide out of that solution and provide a third aqueous solution which contains nitric and fluoboric acids. The uranyl peroxide is then separated from the third aqueous solution.

Abstract: A method and apparatus are provided for producing uranyl peroxide (UO.sub.4.2H.sub.2 O) from uranium-bearing metal pieces. The uranium-bearing metal pieces are dissolved in a first aqueous solution containing nitric and fluoboric acids to provide a second aqueous solution which includes uranyl ions (UO.sub.2.sup.+2) and nitric and fluoboric acids. Hydrogen peroxide is added to the second aqueous solution to precipitate uranyl peroxide out of that solution and provide a third aqueous solution which contains nitric and fluoboric acids. The uranyl peroxide is then separated from the third aqueous solution.

Abstract: A stabilized acidic aqueous composition comprising hydrogen peroxide, metallic ions, and 3-amino-1,2,4-triazine, said triazine being present in an amount sufficient to improve the stability of said composition against decomposition of said peroxide and the methods of utilizing such composition to treat metallic surfaces and to precipitate hydrated uranium peroxide from a solution containing uranium dissolved therein.

Abstract: A process for preparing uranium peroxide in a fluidized bed by precipitation of a uranium peroxide solution obtained from the reaction of uranyl sulfate or uranyl nitrate solutions having a concentration of uranium between about 0.5 and 300 g/L and a hydrogen peroxide solution having a concentration of from about 5 to about 70% by weight at a pH of from about 2.5 to about 4 and a temperature of from about 0.degree. to about 60.degree. C.; together with the easily handled uranium peroxide product so produced, these products being characterized by their substantially spherical form, their mean diameter of from about 30 to about 130 microns, their untamped bulk density of about 1.9 to about 2 g/cm.sup.3, and their Carr scale flowability of at least 95, with a sodium ion content below 300 ppm, such uranium peroxides being useful intermediates for producing nuclear fuel with great safety of manipulation and low sodium content.

Abstract: Uranium peroxide is precipitated from an acidified carbonate strip solution by the addition of hydrogen peroxide and a sufficient quantity of the alkaline carbonate strip solution to maintain the pH at an acid level which is suitable for the precipitation of uranium peroxide.

Abstract: A process as disclosed for recovering uranium values from a carbonate leach solution which comprises directly eluting a column of resin onto which uranium has been sorbed by flowing a concentrated acidic eluant through the column without preconditioning and/or post-conditioning the resin. The concentrated acidic eluant may be flowed upward or, preferably, downward through the column.

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 is described for recovering uranium from a pregnant lixiviant containing uranium values and a certain portion of molybdenum values comprising passing the pregnant lixiviant through an anion-exchange resin to capture the uranium and molybdenum values on the resin, eluting the uranium and molybdenum values from the resin with a salt solution, passing the eluate through a weak acid cationic resin in its hydrogen form to capture the uranium values on the resin and treating and resulting eluate to precipitate uranium therefrom to produce the familiar "yellow-cake.

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 process for the recovery of uranium by precipitation from a rich eluate by contacting the eluate with hydrogen peroxide to produce uranium peroxide yellowcake and reacting the yellowcake with a reducing agent to produce uranium trioxide. The reaction between the yellowcake and the reducing agent may be carried out at a temperature less than 100.degree. C. Subsequent to the reducing step, the uranium trioxide may be washed with water in order to remove water soluble salts present as impurities. Thereafter, the uranium trioxide is dried at a temperature less than 200.degree. C.

Abstract: Uranium values are recovered as uranyl peroxide from wet process phosphoric acid by a solvent extraction-precipitation process. The preferred form of this process comprises a first solvent extraction with DEPA-TOPO followed by reductive stripping of the extractant with Fe.sup.++ - containing phosphoric acid. After reoxidation, the uranium-containing aqueous stripping solution is extracted again with DEPA-TOPO and the pregnant organic is then stripped with a dilute ammonium carbonate solution. The resulting ammonium uranyl tricarbonate solution is then acidified, with special kerosene treatment to prevent wax formation, and the acidified solution is reacted with H.sub.2 O.sub.2 to precipitate a uranyl peroxide compound.

Abstract: A method for recovering uranium and/or related values which include means for protecting ion-exchange resins in the recovery operation from oxidative degradation due to contact with hydrogen peroxide. A guard chamber is positioned in the elution circuit so that barren eluant, after it is stripped of its uranium and/or related values by treatment with hydrogen peroxide, will flow through the chamber. The guard chamber contains catalytic material, e.g. activated carbon, which decomposes hydrogen peroxide upon contact into water and oxygen. The barren eluant, after it passes through the catalytic material, is used to make up fresh eluant for reuse in the recovery method without the risk of the fresh eluant causing oxidative degradation of the resins.

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 method of preparing uranium (VI) peroxide hydrate from uranium tetrafluoride hydrate, comprising the steps of digesting uranium tetrafluoride hydrate in an aqueous acid in the presence of a fluoride complexing agent to produce an aqueous uranium solution, adjusting the aqueous uranium solution to a pH between about 1 to about 3, filtering the aqueous uranium solution to remove undissolved material, reacting the aqueous uranium solution with peroxide to precipitate uranium (VI) peroxide hydrate, and separating the precipitated uranium (VI) peroxide hydrate.

Abstract: A method of preparing uranium (VI) peroxide hydrate from uranium tetrafluoride hydrate, comprising the steps of digesting uranium tetrafluoride hydrate in an aqueous acid in the presence of a fluoride precipitating agent to produce an aqueous uranium solution, filtering the aqueous uranium solution to remove precipitated fluorides and undissolved material, adjusting the aqueous uranium solution to a pH between about 1 to about 3, reacting the aqueous uranium solution with peroxide to precipitate uranium (VI) peroxide hydrate, and separating the precipitated uranium (VI) peroxide hydrate.

Abstract: There is provided a process for the production of pourable, directly compactable uranium dioxide powder by continuous precipitation of uranium peroxide from a stirred uranyl nitrate solution at a pH of 1 - 2.5 with hydrogen peroxide, calcining of the peroxide at 500.degree.- 800.degree. C. and subsequent reduction at 550.degree.- 750.degree. C. The precipitation solution contains 70 - 150 g/l of uranium and 0.80 g/l of ammonium nitrate and is treated with 15 - 20% aqueous hydrogen peroxide solution in the ratio of U:H.sub.2 O.sub.2 from 1:1.5 to 1:3 and there is led through the solution a mixture of ammonia and air having an air-ammonia volume ratio of 1:0.3 - 0.6. An apparatus for carrying out the process is also described. SUBACKGROUND OF THE INVENTIONThe present invention is directed to a process and apparatus for the production of directly pressable (moldable) uranium dioxide powder from uranium peroxide (UO.sub.4.xH.sub.

Abstract: A process of recovering uranium values as uranium peroxide from an aqueous uranyl solution containing dissolved vanadium and sodium impurities by treating the uranyl solution with hydrogen peroxide in an amount sufficient to have an excess of at least 0.5 parts H.sub.2 O.sub.2 per part of vanadium (V.sub.2 O.sub.5) above the stoichiometric amount required to form the uranium peroxide, the hydrogen peroxide treatment being carried out in three sequential phases consisting of I, a precipitation phase in which the hydrogen peroxide is added to the uranyl solution to precipitate the uranium peroxide and the pH of the reaction medium maintained in the range of 2.5 to 5.5 for a period of from about 1 to 60 minutes after the hydrogen peroxide addition; II, a digestion phase in which the pH of the reaction medium is maintained in the range of 3.0 to 7.0 for a period of about 5 to 180 minutes and III, a final phase in which the pH of the reaction medium is maintained in the range of 4.0 to 7.

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: Phosphorus impurities are removed from yellowcake by dissolving it in hydrochloric or sulfuric acid to a U.sub.3 O.sub.8 assay of at least 150 g/l at a pH of 2; precipitating uranium peroxide with hydrogen peroxide while keeping the pH between 2.2 and 2.6 and recovering the uranium peroxide from the phosphorus impurities remaining in solution.