Abstract: A method for converting uranium hexafluoride to uranium dioxide includes steps of hydrolysis of UF6 to uranium oxyfluoride (UO2F2) in a hydrolysis reactor (4) by reaction between gaseous UF6 and dry water vapour injected into the reactor (4), and pyrohydrolysis of UO2F2 to UO2 in a pyrohydrolysis furnace (6) by reaction of UO2F2 with dry water vapour and hydrogen gas (H2) injected into the furnace (6). The hourly mass flowrate of gaseous UF6 supplied to the reactor (4) is between 75 and 130 kg/h, the hourly mass flowrate of dry water vapour supplied to the reactor (4) for hydrolysis is between 15 and 30 kg/h, and the temperature inside the reactor (4) is between 150 and 250° C.

Abstract: Methods and apparatus are provided for producing and extracting Mo-99 and other radioisotopes from fission products that overcome the drawbacks of previously-known systems, especially the excessive generation of radioactive wastes, by providing gas-phase extraction of fission product radioisotopes from a nuclear fuel target using a mixture including halide and an oxygen-containing species with heat to convert the fission product radioisotopes to gas (e.g., Mo-99 to MoO2Cl2 gas). The gaseous species are evacuated to a recovery chamber where the radioisotopes solidify for subsequent processing, while the substantially intact uranium target made available for further irradiation and extraction cycles.

Abstract: The rod contains substantially cylindrical oxide nuclear fuel pellets based on enriched uranium oxide. The H/D ratio of the height over the diameter of the pellets lies in the range 0.4 to 0.6. The initial diametral clearance between the pellets and the cladding does not exceed 200 ?m.

Abstract: A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300° C.

Abstract: A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300° C.

Abstract: This invention relates to the integration of ammonium carbonate leach processes with established acid and alkaline uranium leach processes as multifunctional industrial processes for the extraction, high degree purification and conversion of processed or semi-processed uranium as U3O8, UO2, or most tetra or hexa-valent forms of uranium, and where applicable, for the recovery of uranium from uranium ores, using advanced multiple stage membrane based technologies for the separation and concentration of uranium in solution from heavy metals and lighter elements that may be present in the solution, and the selective leach and precipitation properties of an ammonium carbonate leach.

Abstract: A method for manufacturing a porous fuel comprising uranium, optionally plutonium and at least one minor actinide is provided. The method may comprise the following successive steps: a) a step for compacting as pellets a mixture of powders comprising uranium oxide, optionally plutonium oxide and at least one oxide of a minor actinide, at least one portion of the uranium oxide being in the form of triuranium octaoxide U3O8, the other portion being in the form of uranium dioxide UO2; b) a step for reducing at least one portion of the triuranium octaoxide U3O8 into uranium dioxide UO2.

Abstract: There is provided a method of producing U3O8 powder having large surface area and small particle size by oxidizing defective UO2 pellets and manufacturing nuclear fuel pellets which are stable in a pore structure and high in density through the use of a mixture comprising UO2 powder and U3O8 powder. The method includes producing an U308 powder having a surface area of at least 1 m2/g by oxidizing defective UO2 pellets at a temperature of 300 to 370° C. in such a way that a maximum weight increase rate per 1 g of the UO2 pellets is up to 0.06 wt %/min; producing a mixed powder by mixing the U3O8 powder with an UO2 powder by 2 to 15 wt %; producing a compact by compression molding the mixed powder; and sintering the compact in a reducing gas atmosphere at a temperature of 1600 to 1800° C. In addition, a small amount of an Al-compound may be added to the oxidized U3O8 powder before the U3O8 powder is mixed with the UO2 powder.

Abstract: A dry conversion reactor for converting uranium hexafluoride to uranium dioxide, the dry conversion reactor including a gas-phase reaction segment and a fluidized bed segment, wherein at least one of the gas-phase reaction segment and the fluidized bed segment is a replaceable segment. A method for operating a dry conversion reactor utilizing a uranium hexafluoride to uranium dioxide conversion process, the method including replacing at least one conversion reactor segment.

Abstract: Method for removing the epoxy and/or phenolic polymer encapsulating a nuclear fuel pellet comprising uranium dioxide UO2, the method comprising the following successive steps: a) the polymer is pyrolysed in a reducing atmosphere; and b) the carbon residues obtained after the pyrolysis step (a) are selectively oxidized, the oxidation being carried out at temperature above 1000° C. in an atmosphere comprising carbon dioxide CO2. Such a method makes it possible to remove the epoxy and/or phenolic polymer encapsulating the pellet while avoiding or limiting the risk of radiological contamination by the formation of U3O8.

Abstract: Catalyst for oxidation reactions which comprises at least one constituent active in the catalysis of hydrogen chloride oxidation and support therefor, which support is based on uranium oxide. The catalyst is notable for a high stability and activity.

Abstract: Method for removing the epoxy and/or phenolic polymer encapsulating a nuclear fuel pellet comprising uranium dioxide UO2, the method comprising the following successive steps: a) the polymer is pyrolysed in a reducing atmosphere; and b) the carbon residues obtained after the pyrolysis step (a) are selectively oxidized, the oxidation being carried out at temperature above 1000° C. in an atmosphere comprising carbon dioxide CO2. Such a method makes it possible to remove the epoxy and/or phenolic polymer encapsulating the pellet while avoiding or limiting the risk of radiological contamination by the formation of U3O8.

Abstract: A dry conversion reactor for converting uranium hexafluoride to uranium dioxide, the dry conversion reactor including a gas-phase reaction segment and a fluidized bed segment, wherein at least one of the gas-phase reaction segment and the fluidized bed segment is a replaceable segment. A method for operating a dry conversion reactor utilizing a uranium hexafluoride to uranium dioxide conversion process, the method including replacing at least one conversion reactor segment.

Abstract: The present invention provides a two-step process for producing nuclear grade, active uranium dioxide (UO2) powder in which the first step comprises reacting uranium hexafluoride (UF6) with steam in a flame reactor to yield uranyl fluoride (UO2F2); and the second step comprises removing fluoride and reducing UO2F2 to uranium dioxide (UO2) in a kiln under a steam/hydrogen atmosphere. The two-step process, each step separated by a positive sealed valve means to prevent gas, particularly H2 flow back, tightly controls the exothermicity of the reaction, which allows for a very tight temperature control which controls the growth of the particles and results in UO2 powder that is active and of consistent morphology.

Abstract: The present invention provides a two-step process for producing nuclear grade, active uranium dioxide (UO2) powder in which the first step comprises reacting uranium hexafluoride (UF6) with steam in a flame reactor to yield uranyl fluoride (UO2F2); and the second step comprises removing fluoride and reducing UO2F2 to uranium dioxide (UO2) in a kiln under a steam/hydrogen atmosphere. The two-step process, each step separated by a positive sealed valve means to prevent gas, particularly H2 flow back, tightly controls the exothermicity of the reaction, which allows for a very tight temperature control which controls the growth of the particles and results in UO2 powder that is active and of consistent morphology.

Abstract: Fluorine extraction systems and associated processes are described herein. In one embodiment, a fluorine extraction process can include loading a mixture containing a uranium fluoride (UxFy, where x and y are integers) and an oxidizing agent into a reaction vessel. The reaction vessel has a closed bottom section and an opening spaced apart from the bottom section. The fluorine extraction process can also include heating the mixture containing uranium fluoride and the oxidizing agent in the reaction vessel, forming at least one uranium dioxide and a non-radioactive gas product from the heated mixture, and controlling a depth of the mixture in the reaction vessel to achieve a desired reaction yield of the non-radioactive gas product.

Abstract: A method of declogging at least one filter of a plant for manufacturing uranium oxide from uranium hexafluoride, including separating, from the wall of the filter, uranium oxyfluoride particles deposited, by a stream of inert gas such as nitrogen, injected into the filter, in a counter-currentwise direction to the flow of hydrofluoric acid.

Abstract: A substance in a condensed state, for example a powdered solid, is in continuous movement in the longitudinal direction (6) of a furnace (4, 5). A reactive gas mixture is brought into contact with the substance in the condensed state. A plurality of samples of the gaseous mixture are removed at a plurality of reference points (14) spaced apart from one another along the longitudinal direction (6) of the furnace (4, 5); each of the gas samples is analyzed outside the furnace to determine the composition of the gas mixture and for each point (14), the extent of a chemical reaction between the condensed substance and the reactive gas mixture is deduced from the composition of the gas mixture at each of the reference points (14). In particular, the apparatus comprises a sampling and injection rod (10) introduced into the furnace (4, 5) and disposed in its longitudinal direction (6).

Abstract: Method for producing metal oxide nanoparticles. The method includes generating an aerosol of solid metallic microparticles, generating plasma with a plasma hot zone at a temperature sufficiently high to vaporize the microparticles into metal vapor, and directing the aerosol into the hot zone of the plasma. The microparticles vaporize in the hot zone into metal vapor. The metal vapor is directed away from the hot zone and into the cooler plasma afterglow where it oxidizes, cools and condenses to form solid metal oxide nanoparticles.

Abstract: Sulphidation method for a UO2 powder, in which said powder is sulfurated by bringing it into contact with a gaseous sulphidation agent. Method for manufacturing nuclear fuel pellets based on uranium oxide, or mixed oxide of uranium and plutonium, from a load of totally or partially sulfurated UO2 powder or UO2 powder and PuO2 powder, by lubrication, pelletizing and sintering, in which: the load of powder subjected to the lubrication, pelletizing and sintering is prepared by the following successive steps: sulphidation of a UO2 powder by the above sulphidation method; optionally mixing, said sulfurated powder in a matrix comprising a UO2 powder, or of a UO2 powder and a PuO2 powder; and, subjecting said load, formed from said sulfurated powder or said mixture, to lubrication, pelletizing and sintering operations.

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 depleted UF6 processing plant including a first fluidized bed reactor configured to react depleted UF6 with steam to produce UO2F2 and hydrogen fluoride, a second fluidized bed reactor connected to the first fluidized bed reactor and configured to react the UO2F2 with steam to produce U3O8, hydrogen fluoride and oxygen, a gas cooler configured to cool the hydrogen fluoride generated in the first and second fluidized bed reactors down to 150 to 300° C., and a fluorine fixing reactor containing granular calcium carbonate and connected to the gas cooler to receive the hydrogen fluoride cooled down to 150 to 300° C. from the gas cooler. The fluorine fixing reactor is configured to form granular calcium fluoride from the granular calcium carbonate and the hydrogen fluoride passing through the fluorine fixing reactor.

Abstract: The invention relates to a production process of a composite material composed of aggregates of a blend of UO2 and of PuO2 dispersed in a UO2 matrix comprising the steps of dry co-grinding of a UO2 powder and of a PuO2 powder in order to obtain a homogenous primary blend, of consolidating the primary blend in order to obtain cohesive aggregates, of sieving the aggregates in a range of 20 to 350 ?m, of diluting the sieved aggregates in a UO2 matrix in order to obtain a powder blend, of pelletising the powder blend and of sintering the pellets obtained in order to obtain the composite.

Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: Sulphidation method for a UO2 powder, in which said powder is sulfurated by bringing it into contact with a gaseous sulphidation agent.

Abstract: A method for producing mixed oxide nuclear fuel pellets comprises the steps of preparing an U-Pu oxide blend powder having a Pu content in excess of the finally desired value, preparing uranium oxide powder, mixing adequate quantities of both powders in order to achieve the desired plutonium content and compacting and sintering the mixture for obtaining the pellets. The step of preparing the uranium oxide powder involves the following sequence of substeps: a) preparing an aqueous solution of uranyl nitrate to which between 0.5 and 2 wt % of organic thickeners are added such that the viscosity of the solution is adjusted to values between 20 and 100 centipoise, b) dispersing of the solution into droplets, c) introducing the droplets into a hydroxide bath, d) washing the resulting beads, e) drying the beads by azeotropic distillation using an immiscible organic solvent, f) thermally treating the beads in an oxidizing atmosphere and g) thermally treating in a reducing atmosphere.

Abstract: The invention relates to a process for preparing a castable powder of uranium dioxide UO2, for use in the manufacture of MOX fuel. This process comprises the following stages: 1) to prepare an aqueous suspension of a powder of UO2 obtained by dry process from uranium hexafluoride, said suspension comprising 50 to 80% by weight of UO2 and at least one additive chosen among deflocculation agents, organic binders, hydrogen peroxide H2O2 and a powder of U3O8, in such a quantity that the viscosity of the suspension does not exceed 250 mPa.sec, and 2) to atomise this suspension and dry it in a hot gas, at a temperature of 150 to 300° C., to obtain a castable powder of UO2 with an average particle size of 20 to 100 &mgr;m.

Abstract: The object of the present invention is to simplify the depleted UF6 processing plant and processing method and to prevent calcium fluoride to be a fine powder, wherein the processing plant comprises a first fluidized bed reactor for forming UO2F2 and hydrogen fluoride by allowing depleted UF6 to react with steam, a second fluidized bed reactor for forming U3Oa, hydrogen fluoride and oxygen by allowing UO2F2 to react with steam, a gas cooler for cooling hydrogen fluoride at 150 to 300° C., and a fluoride fixing reactor for forming calcium fluoride by allowing cooled hydrogen fluoride to contact calcium carbonate; and wherein the processing process comprises a dry vapor-phase reaction step for forcing UO2F2 and hydrogen fluoride by allowing depleted UF6 to react with steam at a temperature of 230 to 280° C.

Abstract: A method of dissolving in an ionic liquid a metal in an initial oxidation state below its maximum oxidation state, characterized in that the ionic liquid reacts with the metal and oxidizes it to a higher oxidation state. The initial metal may be in the form of a compound thereof and may be irradiated nuclear fuel comprising UO2 and/or PuO2 as well as fission products. The ionic liquid typically is nitrate-based, for example a pyridinium or substituted imidazolium nitrate, and contains a Bronstead or Franklin acid to increase the oxidizing power of the nitrate. Suitable acids are HNO3, H2SO4 and [NO+]. Imidazolium nitrates and certain pyridinium nitrates form one aspect of the invention.

Abstract: A process for converting UF6 to a solid uranium compound such as UO2 and CaF. The UF6 vapor form is contacted with an aqueous solution of NH4OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH4OH and NH4F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH4OH and NH4F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH)2 to precipitate CaF2 leaving dilute NH4OH.

Abstract: Uranium trioxide is reduced to uranium dioxide using microwave radiation or radiofrequency radiation directed in such a way that the radiation encounters an interface between uranium trioxide and the uranium-containing reduction product without first having passed through that product. By this method, and also using a reducing gas, it is possible to obtain UO2 with an O:U ratio less than 2.04:1.

Abstract: Uranium trioxide Is reduced to uranium dioxide using microwave radiation or radiofrequency radiation directed in such a way that the radiation encounters an Interface between uranium trioxide and the uranium-containing reduction product without first having passed through that product. By this method, and also using a reducing gas, it is possible to obtain UO2 with an O:U ratio less than 2.04:1.

Abstract: A method for producing uranium oxide includes combining uranium tetrafluoride and a phyllosilicate mineral containing a solid oxidizing agent within the mineral's structure having a lower thermodynamic stability than the uranium oxide; heating the combination below the vapor point of the uranium tetrafluoride to sufficiently react the uranium tetrafluoride and the oxidizing agent to produce uranium oxide and a non-radioactive fluorine compound; and removing the fluorine compound.

Abstract: The conversion apparatus comprises in succession: a reactor provided with injectors of UF.sub.6, steam, and nitrogen so as to cause UO.sub.2 F.sub.2 to be formed by hydrolysis; a rotary tubular pyrohydrolysis furnace for transforming UO.sub.2 F.sub.2 into uranium oxide, and provided with heaters distributed in at least five zones; and a tail end for conditioning the oxide powder. The injectors comprise three concentric nozzles connected respectively to inlets for UF.sub.6, nitrogen, and steam, UF.sub.6 being fed to the central nozzle and nitrogen being injected between UF.sub.6 and steam.

Abstract: A thermal decomposition method useful in the nuclear industry for preparing a powdered mixture of metal oxides having suitable reactivity from nitrates thereof in the form of an aqueous solution or a mixture of solids. According to the method, the solution or the mixture of solids is thermomechanically contacted with a gaseous fluid in the contact area of a reaction chamber, said gaseous fluid being fed into the reaction chamber at the same time as the solution or mixture at a temperature no lower than the decomposition temperature of the nitrates, and having a mechanical energy high enough to generate a fine spray of the solution or a fine dispersion of the solid mixture, and instantly decompose the nitrates. The resulting oxide mixtures may be used to prepare nuclear fuels.

Abstract: A method for producing uranium oxide includes combining uranium oxyfluoride and a solid oxidizing agent having a lower thermodynamic stability than the uranium oxide after "oxide"; heating the combination below the vapor point of the uranium oxyfluoride to sufficiently react the uranium oxyfluoride and the oxidizing agent to produce uranium oxide and a non-radioactive fluorine compound; and removing the fluorine compound after "compound".

Abstract: A method for producing uranium oxide includes combining uranium oxyfluoride and silicon and heating the combination below the vapor point of the uranium oxyfluoride to sufficiently react the uranium oxyfluoride and silicon to produce uranium oxide and a non-radioactive fluorine compound; and removing the fluorine compound, e.g. silicon tetrafluoride.

Abstract: A method for producing silicon tetrafluoride includes combining uranium oxyfluoride and silicon dioxide; heating the combination below the melting point of the uranium oxyfluoride to sufficiently react the uranium oxyfluoride and the silicon dioxide to produce non-radioactive silicon tetrafluoride and an oxide of uranium; and removing the silicon tetrafluoride.

Abstract: A rotary kiln in which reaction is to occur between counterflowing reactants and an injector for a reactant extends into a reaction zone in the kiln. The injector is provided with means for adjusting the temperature of the injected reactant to a temperature in the desired range for that zone. Further means are provided for maintaining a temperature in the desired range throughout the zone. The arrangement be used in the production of uranium oxides from uranium hexafluoride.

Abstract: A multicomponent fluid feed apparatus is disclosed that independently preheats and then mixes two or more fluid streams being introduced into a high temperature chemical reactor to promote more rigorous and complete reactions using assemblies of inert tubular elements and an integral mixing orifice plate. The design allows use of ceramic and speciality alloy materials for high temperature service with particularly corrosive halide feeds such as UF.sub.6 and HF. Radiant heat transfer to the tubular elements from external means gives the necessary system high temperatures without excessive temperatures to cause material failure. Preheating of the gaseous reactants in a separate step prior to mixing and injecting the gaseous reactants into a high temperature chemical reactor was found to provide an improved thermal conversion of UF.sub.6 to uranium oxides.

Abstract: A method for producing uranium oxide includes combining uranium tetrafluoride and a solid oxidizing agent having a lower thermodynamic stability than the uranium oxide; heating the combination below the vapor point of the uranium tetrafluoride to sufficiently react the uranium tetrafluoride and the oxidizing agent to produce uranium oxide and a non-radioactive fluorine compound; and removing the fluorine compound.

Abstract: A method for producing uranium oxide includes combining uranium tetrafluoride, silicon and a gaseous anhydrous oxidizing agent having a lower thermodynamic stability than any oxide of uranium produced; heating the combination below the vapor point of the uranium tetrafluoride to sufficiently react the uranium tetrafluoride, silicon and the oxidizing agent to produce uranium oxide and a non-radioactive fluorine compound; and removing the fluorine compound.

Abstract: A method for producing silicon tetrafluoride includes combining uranium tetrafluoride and silicon dioxide; heating the combination below the melting point of the uranium tetrafluoride to sufficiently react the uranium tetrafluoride and the silicon dioxide to produce non-radioactive silicon tetrafluoride and an oxide of uranium; and removing the silicon tetrafluoride.

Abstract: The present invention provides a method for varying and controlling the chemical composition and physical properties of the uranium oxide solids produced by the thermal conversion of UF.sub.6. The method allows the production of predominantly UO.sub.2, U.sub.3 O.sub.8, or UO.sub.3 interchangeably from the same reactor simply by controlling the hydrogen and oxygen contents of the feed relative to uranium. The temperature profile of the thermal reactor is established by specifying the preheat of the feed prior to mixing, the feed composition, and the reactor wall temperature to thus vary and control the physical properties of the resulting solids according to the end use of the uranium product.

Abstract: The present invention provides a process for the oxidation of uranium hexafluoride by injecting the uranium hexafluoride and an oxidant gas together into a reaction vessel to form a plume characterized in that a plurality of the said plumes are formed together in the same vessel, the plumes mutually contributing to a circulating product formation stream in the said vessel. The process may include the establishment of three or more plumes simultaneously contributing to the reaction between the gases in the reaction vessel. The oxidant gas may comprise steam. The process may be one in which the product is formed as a particulate solid. The product particles may initially be formed as dentritic particles which may be recirculated in the reaction vessel to promote seeding, growth, agglomeration and aggregation of the required product particles, the plumes thereby contributing to the product formation process in the vessel.

Abstract: A single-step process for producing solid uranium oxide and gaseous HF from UF.sub.6 which comprises bringing together two gaseous reactant streams, one of said streams comprising UF.sub.6 optionally admixed with oxygen as O.sub.2, and the second reactant stream comprising a mixture of hydrogen as H.sub.2 or as a hydrogen-containing compound and oxygen as an oxygen-containing compound, said gaseous reactant streams being brought together at a temperature and composition such that the UF.sub.6 is converted rapidly by flame reaction into readily separable solid uranium oxide and a gaseous HF product.

Abstract: A process for the production of a uranium oxide in which UF.sub.6 is converted by reaction with steam into UO.sub.2 F.sub.2 in a first step and by further reaction of UO.sub.2 F.sub.2 with steam and/or hydrogen in a second step to produce an oxide of uranium, the process being carried out in an apparatus comprising a single kiln vessel having a first region, having UF.sub.

Abstract: This invention provides an improved process of preparing UO.sub.2 powder from poor quality, partially oxidized powder containing organic and inorganic impurities. The process is illustrated in the flow chart of FIG. 1 which includes the steps of (a) oxidizing a uranium-containing scrap also containing inorganic and cationic organic impurities; (b) solubilizing, typically with nitric acid, the uranium contained in the oxidized scrap to produce uranyl nitrate; (c) solvent extracting the solubilized product of step (b) to remove cation impurities to provide a purified uranyl nitrate solution; (d) precipitating the purified uranyl nitrate with ammonia to form ammonium diuranate powder; and (e) calcining and passivating the ammonium diuranate powder to produce UO.sub.2 powder; and optionally (f) forming the UO.sub.2 powder of step (e) into pellets and sintering the formed pellets to produce sintered UO.sub.2 pellets.

Abstract: Finely powdered urania scrap materials are subjected to sintering conditions (greater than 1600.degree. C.) in a hydrogen atmosphere for about 4 to 8 hours in order to remove impurities (such as Si, Fe, Ni, Sn, Cu, Na and Pb). The process upgrades the quality of the urania powder to where it can be used directly as clean scrap makeup, thereby avoiding expensive decontamination and recovery steps like solvent extraction. The novelty of the process is in the use of sintering conditions (greater than 1600.degree. C.) in a hydrogen atmosphere on finely divided powder to decontaminate urania materials.

Abstract: A process for reclaiming scrap UO.sub.2 materials yields a high-sinter-density pellet. The scrap is oxidized in a high-temperature furnace to produce U.sub.3 O.sub.8. The U.sub.3 O.sub.8 particles from the oxidation furnace are reacted with nitric acid to produce a solution of uranyl nitrate that meets the concentration and free acid requirements of the ADU precipitation process. A controlled two-stage ADU precipitation process is carried out to produce ADU particles with a size and morphology that leads to high-surface-area UO.sub.2 powder with excellent sintered pellet ceramic characteristics. After calcination and hydrogen reduction to UO.sub.2, the high-surface-area UO.sub.2 powder is passivated.