Abstract: Process for obtaining nuclear fuel pellets, which produces no liquid effluent, for which the starting product is metallic uranium, which is oxidized to U.sub.3 O .sub.8, then crushed and either reduced to UO.sub.2 and activated (or vice versa) with the aid of at least one fine milling operation, or reduced to UO.sub.2 and activated using at least one oxidation-reduction cycle, the UO.sub.2 obtained then being shaped by pressing and fritting and the intermediate powders obtained are dense and pourable, no intermediate conditioning operation being required.

Abstract: A process for producing a sinterable doped superconductive powder which contains a sufficient amount of superconductive compound which is at least coated with sufficient uranium-235 and/or plutonium-239 dopant atoms to enable the sinterable powder to be formed into a sintered superconductive body which can be irradiated with thermal neutrons to produce from about 0.25.times.10.sup.14 to about 5.times.10.sup.14 fission events per cubic centimeter of the resulting irradiated superconductive body, said superconductive compound being selected from the group consisting of Bi.sub.2 CaSr.sub.2 Cu.sub.2 O.sub.8 .+-.x where x ranges from zero to 0.5, Bi.sub.2-y Pb.sub.y Ca.sub.2 Sr.sub.2 Cu.sub.3 O.sub.10.+-.z where y ranges from 0.1 to 0.5 and z ranges from zero to less than 1, and mixtures thereof.

Abstract: A method of passivating the surface of particulate uranium oxides is disclosed comprising a process of continuously contacting uranium oxide particles with an oxygen containing and cooling counter flowing gas stream. The treatment produces a protective surface which inhibits subsequent oxygen chemisorption of the particulate uranium oxides.

Abstract: A process for controlling the oxidation reaction of oxides of uranium and fixing the ratio of oxygen to uranium in uranium oxide compounds by means of a passification process, and the stabilized uranium oxide compounds produced therefrom. The method is especially useful in the production of uranium oxide fuel for nuclear reactors.

Abstract: Making uranium dioxide pellets of controlled grain size by treating 50-500 g/l UO.sub.2 F.sub.2 with NH.sub.3 in a first and a second stages to form (NH.sub.4).sub.2 U.sub.2 O.sub.7 precipitate, wherein the NH.sub.3 /U molar ratio is between 3-5 in the first stage and between 6-12 in the second stage. The precipitate is then formed into UO.sub.2 pellets having grain size within the range from 10 to 100 .mu.m.

Abstract: A polycrystalline Y-Ba-Cu-O superconductor doped with sufficient .sup.235 U and/or .sup.239 Pu atoms is irradiated with thermal neutrons to produce from about 2.5.times.10.sup.14 to about 50.times.10.sup.14 fission events per cubic centimeter of superconductor.

Abstract: Method for treating plutonium and/or uranyl nitrate by superheating and concentrating a nitric acid starting solution, characterized by the feature that nitric acid starting solution is concentrated to form a Pu(VI)-and/or U(VI)-containing nitrate melt, is poured into a mold, is solidified by cooling down, and is transported and/or stored as a solidified solid body until further processing.

Abstract: A process for producing large-grain, low-density UO.sub.2 micro-fuel spheres, the micro-fuel spheres having an average grain diameter of at least 5 .mu.m, which comprises:(a) isothermally heating one or more UO.sub.2 micro-fuel spheres in a stream consisting essentially of carbon dioxide for a time and at a temperature sufficient to produce grains having an increased average grain diameter of greater than 5 .mu.m, without substantially increasing the density of said micro-fuel spheres; and then(b) sintering said produced grains, in the absence of a sintering additive, in a reducing gas stream for a time and at a temperature sufficient to regulate both the ratio of oxygen to uranium in the produced grains, and the density of the grains.

Abstract: A process of fabricating UO.sub.2 pellets comprising the steps of filtering and drying a slurry of ammonium diuranate (ADU) including ammonium fluoride (NH.sub.4 F), in order to form ADU powder, and then subjecting the ADU powder to calcining, reducing, compacting and sintering treatments, to form UO.sub.2 pellets.

Abstract: The present invention provides a process for producing a uranium oxide by dissolving a yellow cake in sulfuric acid or hydrochloric acid, bringing the obtained solution into contact with a chelating resin of diaminocarboxylic acid type and subjecting the product to neutralizing precipitation followed by heat treatment. By the contact of the solution with the chelating resin, iron, copper, molybdenum and vanadium among the metallic impurities are removed and in the subsequent neutralizing precipitation step, other metallic impurities such as aluminum, calcium, magnesium, sodium and potassium are removed. This process can easily produce uranium oxide having a high purity using a simple apparatus.

Abstract: The present invention contemplates the conversion of uranium hexafluoride with superheated steam to produce a fine, submicron, uranyl fluoride powder which is agglomerated and densified in a fluidized bed of a uranium oxide material. Thereafter, the agglomerated and densified uranyl fluoride is defluorinated and reduced in the fluidized bed to yield a fluoride containing uranium oxide material having a uranium to oxygen ratio of from about 1:2.0 to about 1:2.67. The so-produced fluoride contains uranium oxide material treated with steam and hydrogen to produce ceramic grade uranium dioxide.

Abstract: Method for manufacturing sintered oxidic nuclear fuel bodies by molding uranium oxide starting powder, a mixture of uranium oxide and plutonium oxide starting powder or of uranium-plutonium oxide mixed-crystal starting powder to form blanks and by a heat treatment of these blanks with a U.sub.4 O.sub.9 or (U, Pu).sub.4 O.sub.9 crystal phase developed in them to a degree which can be crystallographically detected at a sintering temperature in the range of 1000.degree. to 1400.degree. C. in an oxidizing and subsequently in a reducing gas atmosphere. The starting powder and/or the blanks are preroasted at a roasting temperature below the sintering temperature in a roasting gas atmosphere with oxidizing action and with an oxygen potential in which the U.sub.4 O.sub.9 (U, Pu).sub.4 O.sub.9 crystal phase is developed and cooled down subsequently to a starting temperature below the roasting temperature in an inert or oxidizing cooling-down gas atmosphere.

Abstract: A process for converting UF.sub.6 gas into UO.sub.2 powder comprising blowing UF.sub.6 gas and steam into a fluid bed to produce UO.sub.2 F.sub.2 particle, hydrating and dehydrating the UO.sub.2 F.sub.2 particle to UO.sub.2 F.sub.2 powder, and defluorinating and reducing the UO.sub.2 F.sub.2 powder to UO.sub.2 powder. The UO.sub.2 powder is suitable for manufacturing a reactor fuel owing to its high activity, low remaining fluorine amount and excellent fluidity.

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 process for conversion of gaseous UF.sub.6 to UO.sub.2 powders by using a fluidized bed reaction apparatus comprising pyrohydrolizing gaseous UF.sub.6 and steam to obtain UO.sub.2 F.sub.2 particles, hydrating and dehydrating the UO.sub.2 F.sub.2 particles to UO.sub.2 F.sub.2 anhydride and reducing the UO.sub.2 F.sub.2 anhydride to UO.sub.2 powders. The obtained UO.sub.2 powders are suitable for production of nuclear fuels in power plant owing to its good ceramic properties, low fluorine contents and free flowability.

Abstract: A method for monitoring the stack gases of a purge cascade of a gaseous diffusion plant for uranium activity. A sample stream is taken from the stack gases and contacted with a volume of moisture-laden air for converting trace levels of uranium hexafluoride, if any, in the stack gases into particulate uranyl fluoride. A continuous strip of filter paper from a supply roll is passed through this sampling stream to intercept and gather any uranyl fluoride in the sampling stream. This filter paper is then passed by an alpha scintillation counting device where any radioactivity on the filter paper is sensed so as to provide a continuous monitoring of the gas stream for activity indicative of the uranium content in the stack gases.

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 uranium oxide product is formed in a gaseous reaction medium from uranium hexafluoride by interaction with a fuel gas such as hydrogen and oxygen. Oxide content and particle properties of the uranium oxide product are governed by the improved method and the apparatus for carrying out the method. An improvement is provided by applying a quenching gas to quench the conversion reaction of uranium hexafluoride to uranium oxide, and a combustion reactor apparatus which enables the application of the quenching gas is also provided.

Abstract: Process for the preparation of pulverulent metallic oxides with predetermined reactivity which is adjusted by the specific surface area, which are intended to be subjected to subsequent conversions, by thermal treatment of corresponding nitrates in the form of an aqueous solution or of a solids material, consisting of hydrated uranyl nitrate, on its own or in admixture with at least one of thorium nitrate, cerium nitrate and plutonium nitrate, which process is characterized in that it consists of two treatment stages:(a) the first stage essentially consisting of incompletely dehydrating the hydrated uranyl nitrate.(b) the second stage essentially consisting of decomposing the product resulting from the first stage by calcination under controlled steam pressure.The process is used for obtaining uranium oxide on its own or in admixture with thorium oxide, cerium oxide and plutonium oxide, with a view to subsequent chemical conversions and/or sintering.

Abstract: A reproducible method and stable binder composition for preserving the induced plasticity of a particulate ceramic material admixture comprising uranium dioxide and a fugitive binder, and product thereof, for subsequent compaction by compressing molding in the manufacture of nuclear fuel pellets.

Abstract: A process for manufacturing uranium oxide powder from UF.sub.6 which comprises converting UF.sub.6 to UO.sub.2 F.sub.2 by its reaction with excess alcohol in gas phase and further converting the formed UO.sub.2 F.sub.2 to uranium oxide by combusting hydrocarbon formed in the gas phase reaction and the excessive part of alcohol with oxygen containing gas supplied separately and supplying a regulated amount of steam separately to the combustion reaction zone.

Abstract: A process is described for the recovery or uranium from wet-process phosphoric acid utilizing an alkyl pyrophosphoric extractant. The extractant also contains a modifier for retaining the alkyl pyrophosphoric acid ester in solution in an essentially water-immiscible organic diluent during stripping. After extracting the uranium from wet-process acid, the APPA extractant is treated with an oxidizing agent such as hydrogen peroxide and the uranium stripped into an alkaline stripping solution. The alkaline stripping solution is an aqueous solution of an alkali metal or ammonium carbonate unsaturated in uranium. The barren extractant is recycled for contacting with fresh wet-process acid. Any ferric iron present is precipitated in the stripping solution and then separated from the solution. The uranium is precipitated, separated, dried and calcined to a U.sub.3 O.sub.8 product.

Abstract: A process is described for the recovery of uranium from wet-process phosphoric acid utilizing an alkyl pyrophosphoric extractant. The extractant also contains a modifier for retaining the alkyl pyrophosphoric acid in solution in an essentially water-immiscible organic diluent during stripping. After extracting the uranium from wet-process acid, the APPA extractant is treated with an oxidizing agent such as hydrogen peroxide and the uranium stripped into an alkaline stripping solution. The alkaline stripping solution is an aqueous solution of an alkali metal or ammonium carbonate or hydroxide. The barren extractant is recycled for contacting with fresh wet-process acid. The uranium and any ferric iron present is precipitated in the stripping solution. The uranium is dissolved in an aqueous solution such as an aqueous solution of ammonium uranyl tricarbonate (AUT), any precipitated ferric solids are separated from the solution, and the uranium is reprecipitated, separated, dried and calcined to a U.sub.3 O.sub.

Abstract: Uranium dioxide powder produced by a gas phase process in which uranium hexafluoride is reacted with dry steam and then with steam and/or hydrogen at a higher temperature is subjected to mechanical treatment, e.g. milling, to break down its structure and increase its packing density. Other powders may be included with the uranium dioxide. The treated powder is mixed with a limited quantity (e.g. 0.5% by weight) of binder, preferably a high strength adhesive, to produce a free flowing powder and formed into pellets by pressing. The pellets are then sintered. Optionally the free flowing powder is spheroidised by tumbling prior to pressing into pellets.

Abstract: A process for the production of uranium trioxide having a large specific surface by the thermal denitration of hydrated uranyl nitrate corresponding to the formula UO.sub.2 (NO.sub.3).sub.2 .multidot.xH.sub.2 O, in which 2.ltoreq.x.ltoreq.6, characterized in that, in a first stage, the said nitrate is treated in the form of a liquid at room temperature in the range from 160.degree. C. to 260.degree. C. until a solid phase containing at least 55% by weight of uranium is obtained and, in a second stage, the said solid phase is progressively heated to a temperature of at most 600.degree. C. at a heating rate of at most 1000.degree. C. per hour, and is kept at this temperature of at most 600.degree. C. until a second solid phase of UO.sub.3 having a large specific surface is obtained. The UO.sub.3 product produced has a specific surface of at least 10 m..sup.2 g.sup.-1.

Abstract: A method for treating process waste consisting of uranium hexafluoride, UF.sub.6, molybdenum hexafluoride, MoF.sub.6 and molybdenum oxytetrafluoride, MoOF.sub.4 is provided wherein the foregoing mixture is hydrolyzed in an aqueous solution of ammonium carbonate and ammonium hydroxide. The resulting mother liquor is digested at 60.degree.-80.degree. C. to drive off carbon dioxide and convert the ammonium uranyl carbonate, (NH.sub.4).sub.4 UO.sub.2 (CO.sub.3).sub.3 to ammonium diuranate, (NH.sub.4).sub.2 U.sub.2 O.sub.7. The pH is maintained above 9 to prevent premature precipitation of the molybdenum values as molybdic acid or uranyl molybdate.The ammonium diuranate which forms as a yellow solid is filtered, slurried in aqueous ammonium hydroxide, filtered and dried. The yield is quantitative. Pyrolysis results in conversion of the diuranate salt to uranium oxide, U.sub.3 O.sub.8, of suitable purity to be recycled in the uranium hexafluoride production process.

Abstract: Hydroxymethane diphosphonic acid and alkali metal or ammonium salt of such acid are prepared. They are useful in detergent compositions and in sequestering and chelating polyvalent metals.

Abstract: UO.sub.2 particles are brought into contact with liquid water of low pH in an oxidizing environment to dissolve uranyl hydroxide from the particle surfaces and cause more uranyl hydroxide to form and be dissolved. When sufficient uranyl hydroxide has entered the solution, the pH is raised to cause uranyl hydroxide to leave the solution and deposit on the UO.sub.2 particles. The particles are then ready for contact with an appropriate reactant to form a fugitive binder.

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: 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: For the operation of material-test-reactors, recently there have been considered fuels having lower degrees of enrichment in U-235. For this purpose, there is needed U.sub.3 O.sub.8 powder having a crystal size of 40 to 90 .mu.m, which is producible economically and in whose production there is developed as little secondary waste as possible and a pore-free product. For this purpose, uranium compounds of the stage of oxidation +4 to +6 are calcined in an oxygen containing atmosphere whereby the calcining is carried out above 1250.degree. C. in an atmosphere containing more than 30 volume % oxygen.

Abstract: A recovery process of uranium comprising:(1) extracting uranium ions with an organic solvent containing one or more compounds selected from the group consisting of alkyl phosphoric acid, alkyl-aryl phosphoric acid, alkyl dithio phosphoric acid, aryl dithio phosphoric acid, neutral phosphoric acid ester and alkyl amine together with a petroleum hydrocarbon as a diluent; and(2) stripping the uranium ions in the resultant organic solvent from the step (1) to an aqueous phase with contact of an aqueous solution containing one or more compounds selected from the groups of NH.sub.4 F, NH.sub.4 HF.sub.2, KF or KHF.sub.2.

Abstract: A method for improving the physical properties of pressed bodies or pellets formed of particulate nuclear fuel material containing uranium dioxide. The green or unfired pressed bodies comprise a fugitive binder dispersed through the particulate fuel material.

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: For the production of plate shaped fuel elements for material testing and research reactors with highly enriched uranium recently there has been needed U.sub.3 O.sub.8 fuels which have a high density, high strength and a small open porosity. Such fuels are obtained if U.sub.3 O.sub.8 powder produced in known manner is first compressed mechanically to molded bodies of any shape, then processed to a granulate having a size of fuel grains below 200.mu. and subsequently sintered to high density particles, preferably at 1370.degree..+-.50.degree. C.

Abstract: In a process for the conversion of uranium hexafluoride to an uranium oxide by injecting uranium hexafluoride and dry steam into a first region of a vessel so as to form a plume of particles of uranyl fluoride and reacting the uranyl fluoride in a second region of the vessel with a countercurrent flow of steam and/or hydrogen the operation of the process is such that a major proportion of the uranyl fluoride is caused to circulate within the first region so that the original uranyl fluoride particles are able to grow and agglomerate in a dendritic manner.

Abstract: An improved method for preparing nuclear fuel pellets and recycling uranium dioxide utilizing microwave radiation wherein microwave induction furnaces replace conventional refractory-type sintering and shaker-air furnaces.

Abstract: Sodium uranate which is easy to handle, characterized by the spherical form of the grains, the average diameter thereof, between 40 and 150 .mu.m, their uncompacted apparent bulk density of 2.5 to 3.5 g/cm.sup.3 and a flow value higher than or equal to 95, measured on the CARR scale produced by diluting the sodium uranate-containing solution introduced into a fluid bed reactor to between about 0.5 and 5 g/l within a period of less than about 2 seconds before the uranium-containing solution enters the reactor.

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 the neutralization of the uraniferous liquor then the fixation of the uranium and the SO.sub.4.sup.= ions over a strong anion-exchange resin, the elution thereof with an ammonium carbonate and/or bicarbonate solution in the form of an ammonium uranyl carbonate which is subsequently decomposed and ammonium sulphate, the precipitation and the calcination of the precipitated uranates and/or diuranates with regeneration of the eluent solution of ammonium carbonate and/or bicarbonate, the causticiation of the liquor freed from uranium which leaves the anion-exchange resin to eliminate the SO.sub.4.sup.= ions in calcium form, recycling a carbonated and/or bicarbonated liquor which is free from SO.sub.4.sup.

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: This information relates to the recovery of uranium from uranium peroxide yellowcake produced by precipitation with hydrogen peroxide. The yellowcake is calcined at an elevated temperature to effect decomposition of the yellowcake to uranium oxide with the attendant evolution of free oxygen. The calcination step is carried out in the presence of a reducing agent which reacts with the free oxygen, thus retarding the evolution of chlorine gas from sodium chloride in the yellowcake. Suitable reducing agents include ammonia producing compounds such as ammonium carbonate and ammonium bicarbonate. Ammonium carbonate and/or ammonium bicarbonate may be provided in the eluant used to desorb the uranium from an ion exchange column.

Abstract: High purity ammonium uranyl tricarbonate is produced from crude plant yellow cake by treatment of the yellow cake with a mineral acid, neutralization of the acidic solution with a conversion of the uranium therein to ammonium uranyl tricarbonate, precipitation of the ammonium uranyl tricarbonate with a non-solvent, followed by separation. Optionally, the ammonium uranyl tricarbonate can be calcined to produce high purity uranium oxides.

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: 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: According to one embodiment of the present invention, the oxygen gas is introduced into a horizontal, unitary and totally-closed furnace shell, and nuclear fuel to be reprocessed is charged onto an oxidation and separation plate which is horizontally disposed for vibration within the furnace shell along the axis, so that the nuclear fuel may be reprocessed by oxidation in the oxygen atmosphere while the fuel is vibrated and transported on the oxidation and separation plate and the reprocessed nuclear fuel in the form of particles may be discharged. At the same time released volatile fission products from the pulverized fuel are isolated by means of a vacuum blower. According to another embodiment of the present invention, a reduction furnace which is substantially similar in construction to the oxidation furnace described above except that a reducing gas is supplied by an independent source.

Abstract: A method is described for the preparation of U.sub.3 O.sub.8 nuclear fuel material by direct precipitation of uranyl formate monohydrate from uranyl nitrate solution. The uranyl formate monohydrate precipitate is removed, dried and calcined to produce U.sub.3 O.sub.8 having a controlled particle size distribution.

Abstract: A method of making metallic oxide or carbide particles of uranium, which prises fuels or breeder materials for nuclear reactors. An aqueous solution of uranyl nitrate or chloride and, if necessary, colloidal carbon is added dropwise into an organic ketone or ketone mixture phase which is located above an aqueous ammonia solution. The thereupon formed particles are sintered.

Abstract: A reductive stripping flow sheet for recovery of uranium from wet-process phosphoric acid is described. Uranium is stripped from a uranium-loaded organic phase by a redox reaction converting the uranyl to uranous ion. The uranous ion is reoxidized to the uranyl oxidation state to form an aqueous feed solution highly concentrated in uranium. Processing of this feed through a second solvent extraction cycle requires far less stripping reagent as compared to a flow sheet which does not include the reductive stripping reaction.