Abstract: The present invention relates to a process for sintering a compacted powder of at least one oxide of a metal selected from an actinide and a lanthanide, this process comprising the following successive steps, carried out in a furnace and under an atmosphere comprising an inert gas, dihydrogen and water: (a) a temperature increase from an initial temperature TI up to a hold temperature TP, (b) maintaining the temperature at the hold temperature TP, and (c) a temperature decrease from the hold temperature TP down to a final temperature TF, in which the P(H2)/P(H2O) ratio is such that: 500<P(H2)/P(H2O)?50 000, during step (a), from TI until a first intermediate temperature Ti1 between 1000° C. and TP is reached, and P(H2)/P(H2O)?500, at least during step (c), from a second intermediate temperature Ti2 between TP and 1000° C., until TF is reached.

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 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: Neutron source comprising a composite, said composite comprising crystals comprising BeO and AmBe13, and an excess of beryllium, wherein the crystals have an average size of less than 2 microns; the size distribution of the crystals is less than 2 microns; and the beryllium is present in a 7-fold to a 75-fold excess by weight of the amount of AmBe13; and methods of making thereof.

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: 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: Among fuel rods constituting a fuel assembly, Gd compound oxide is added to low Gd containing fuel rods that containing uranium dioxide of which enrichment exceeds 5 wt %. The Gd compound oxide is oxide of gadolinium and rare earth element A except for gadolinium and is expressed as a chemical formula A1-xGdxO2-0.5x or a chemical formula A1-xGdxO1.5. As the rare earth element A, cerium Ce, lanthanum La or erbium Er can be used.

Abstract: The invention relates to a method constituting an improvement of the PUREX method, which makes it possible to obtain separation of uranium from the other actinides (Pu, Np, Th, . . . ) in a single purification cycle. This method successively comprises: a) co-extracting the uranium(VI), plutonium(IV) and other actinides(IV) or (VI) from an aqueous nitric solution by using solvent phase and scrubbing the latter; b) back-extracting the plutonium in oxidation state (III) from the solvent phase by using an aqueous nitric solution; c) back-extracting the uranium in oxidation state (VI) from the solvent phase by using an aqueous nitric solution; d) concentrating the aqueous nitric solution resulting from step c) with respect to uranium(VI); and it is characterized in that some of the uranium(VI)-concentrated aqueous solution obtained in step d) is used for back-extracting the actinide(IV) or actinides(IV) from the solvent phase during step b) or between steps b) and c).

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

Abstract: Process for fabrication MOX fuel from weapon originating plutonium oxide (W—PuO2), including a process for fabricating MOX fuel pellets, a dry processing of W—PuO2 powder, and a dry processing of MOX fabrication scrap.

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

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 &mgr;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: 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: Disclosed is a method of manufacturing large-grained uranium dioxide pellets through the use of U3O8 single crystals. The method consists of two main steps; producing U3O8 single crystals and manufacturing large-grained UO2 pellets. The U3O8 single crystals are produced by annealing U3O8 powder to make U3O8 polycrystalline aggregates having a large crystal size, and by dividing the U3O8 polycrystalline aggregate into its constituent U3O8 single crystals. Large-grained UO2 pellets are produced by forming a mixture comprising UO2 powder and the U3O8 single crystals, making granules of the mixture, pressing the granules into green pellets, and sintering the green pellets above 1600° C. for more than one hour in a reducing gas. The U3O8 powder is produced preferably by heating defective UO2 pellets in air to oxidize UO2 to U3O8. The invention provides a method of manufacturing a large-grained UO2 pellet with defective pellets reused.

Abstract: This invention relates to a composite nuclear fuel material and method of manufacture of said material.The purpose of the invention is a composite nuclear fuel material comprising a matrix inert under irradiation and particles of nuclear fuel offering good resistance to fissure propagation under irradiation and high retention of volatile fission products.This purpose is achieved using a method which allows the creation of a clearance of a few microns between the particles and the matrix.

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: A method for purifying metallic alloys of uranium for use as nuclear reactor fuels in which the metal alloy is first converted to an oxide and then dissolved in nitric acid. Initial removal of metal oxide impurities not soluble in nitric acid is accomplished by filtration or other physical means. Further purification can be accomplished by carbonate leaching of uranyl ions from the partially purified solution or using traditional methods such as solvent extraction.

Abstract: The present invention relates, in general, to a method of producing mixed oxide nuclear fuel and, in particular, to a method of reducing the amount of residual gas absorbed on fuel pellets after sintering.

Abstract: A method and an apparatus for producing microspherical ceramic particles such as particulate ceramic fuels for use in nuclear reactors. To produce gel particles by performing microwave heating on the small droplets of a stock solution or sol that contains uranium and other elements for nuclear reactor fuels, the small droplets are allowed to fall down through a cavity resonator using microwaves having an engineering frequency of 2.45 GHz and a quartz rod is inserted into or withdrawn out of the resonator to adjust the resonant frequency so that it will lie within the spectrum width of the microwave frequency.

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.

Abstract: A process for producing a UO.sub.2 pellet comprising the steps of producing UO.sub.2 powder in accordance with the ADU (ammonium diuranate) method or the AUC (ammonium uranyl carbonate) method, forming a compact of said UO.sub.2 powder, and sintering the compact, wherein UO.sub.2 powder having a specific surface area of 5-50 m.sup.2 /g is used as a raw material in the compact forming step. At least one of chlorine or a chlorine compound (or bromine or a bromine compound) is added, in one or more of the UO.sub.2 powder producing step, compact forming step, or compact sintering step, in an amount such that the chlorine content (or bromine content) in the UO.sub.2 pellet amounts to 3-25 ppm chlorine (or 6-50 ppm bromine).

Abstract: Neptunium of minor actinide nuclides separated from spent fuel is added to fuel of reactor cores (inner reactor cores and/or outer reactor cores) of a fast reactor and americium of the separated minor actinide nuclides and rare earth elements are added to either or both of radial and axial blankets of the fast reactor for burning. Thus, the minor actinide nuclides with long half-lives can be burnt with the fast reactor core with the minimized effects of the rare earth elements. For a burner reactor, americium and rare earth elements may be added to shields for burning. Curium may be added together with americium and rare earth elements. Neptunium is added in amount of 2% to 5% by weight based on the weight of the fuel and the rare earth elements are added in an amount of 50% by weight or less based on the weight of the fuel. A Purex process is used to separate neptunium and a Truex process is used to separate americium and curium.

Abstract: The present invention relates to a process of decontaminating zirconium-based alloy claddings tubes used in nuclear reactor fuel rods. The process involved the use of a permanganate in a dilute acid solution. The process renders the alloy suitable for uncontrolled release into a non-radioactive environment.

Abstract: The improved oxide, once-through plutonium fuel compound that can be used for nuclear fission in currently operating light-water reactors and fast reactors has a composition in the range defined by the lines that connect the three compositional points of a three-component system consisting of plutonium dioxide (PuO.sub.2), a plutonium host phase and alumina (Al.sub.2 O.sub.3). The compound also has such a phase structure that two phases, the plutonium host phase having plutonium dioxide dissolved therein and the alumina phase, are in equilibrium.

Abstract: A continuous denitration apparatus by microwave heating comprises an oven into which microwaves are applied and a cylindrical rotary drum disposed horizontally and rotatably disposed inside the oven. The rotary drum has the outer peripheral surface composed of a honeycomb or porous sheet-like microwave dielectric member. Inside the oven are disposed a trough for storing a nitrate solution of a nuclear fuel material below the rotary drum so as to immerse a lower portion of the outer peripheral surface of the rotary drum in the solution stored inside the trough, a heater above the rotary drum so as to heat the outer peripheral surface of the rotary drum opposing the heater, and a scraper for scraping off a denitrated product formed on the outer peripheral surface of the roatray drum between the heater and the trough.

Abstract: ADU (ammonium diuranate) is prepared in particle form directly by reacting ammonium gas with liquid droplets of atomized uranyl compound solutions. Generation of liquid filtrate is prevented by using concentrated solutions of uranyl compounds as feed solutions, or drying the wet ADU particles formed before their settlement when a feed of low concentration is used. The ADU particle thus prepared is finely divided and easy-handling. No filtration operation is necessary in the preparation. The UO.sub.2 powder consequently obtained after calcining and reduction has consistent quality from batch to batch and has good pelletizing and sintering properties. Uranium dioxide with low fluorine content can be prepared from uranyl fluoride solution. Gadolinium-uranium oxide can also be prepared with the present method using an aqueous mixture of gadolinium nitrate and uranyl nitrate as a feed solution.

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: The disclosed invention consists of a ceramic form of fuel for a nuclear reactor comprising an oxide of a fissionable element and containing therein a consumable neutron absorbing agent. The ceramic fissionable fuel product provides longer fuel performance in nuclear reactor service.

Abstract: A method of fabricating uranium dioxide (UO.sub.2) powder from uranium hexafluoride (UF.sub.6) is disclosed, which comprises(1) reacting UF.sub.6 gas with steam with controlling the temperature of reaction between said UF.sub.6 gas and said steam at a predetermined temperature within the range of 200.degree. to 700.degree. C., to form solid uranyl fluoride (UO.sub.2 F.sub.2) and/or uranium oxide with an O/U ratio (oxygen-to-uranium atomic ratio) of 2.7 to 3,(2) dissolving said UO.sub.2 F.sub.2 and/or uranium oxide in water or nitric acid to form an aqueous uranyl solution containing UO.sub.2 F.sub.2 and/or uranyl nitrate (UO.sub.2 (NO.sub.3).sub.2),(3) reacting said aqueous uranyl solution with ammonia to precipitate ammonium diuranate (ADU),(4) filtering said precipitate,(5) drying said precipitate,(6) calcining said dry precipitate, and(7) reducing said calcined precipitate, whereby controlling the characteristics of said UO.sub.2 powder.

Abstract: For manufacturing uranium oxide based nuclear fuel pellets, a fine and reactive U.sub.3 O.sub.8 powder is mixed with a fine UO.sub.2 powder obtained by dry conversion. The U.sub.3 O.sub.8 is obtained by oxidation in air of UO.sub.2 obtained by a dry process, at a temperture less than 800.degree. C.

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: A process is disclosed for producing sintered mixed metal oxide nuclear f pellets containing UO.sub.2 and the oxide of at least one other fissionable or rare earth element M, the pellets being directly soluble in nitric acid without nitric acid additive or prior treatment of the pellets. The process comprises the steps of mixing together nitrate solutions of the elements, concentrating the mixture of solutions, thermally denitrating the concentrated nitrate mixture without additives, to obtain an intermediate mixed oxide powder, calcining the intermediate mixed oxide powder, reducing the calcined mixture, stabilizing the uranium oxide UO.sub.2 in the reduced oxide mixture, shaping and pressing the resulting stabilized, reduced oxide mixture to obtain pellets of green material, sintering the pellets of green material and grinding the sintered pellets.

Abstract: The invention relates to an ammonium plutonate uranate, its preparation process and its use for producing the mixed oxide (U,Pu)O.sub.2.The ammonium plutonate uranate of formula [Cu,Pu)O.sub.3 ].sub.2 [NH.sub.3 ].sub.x [H.sub.2 O].sub.y in which x and y have a value from 0 to 5 with x+y=5, is prepared by neutralization with ammonia of an aqueous solution of uranium VI and plutonium VI salts, e.g. uranyl and plutonyl nitrates. This solution can be prepared by oxidizing a solution of plutonium IV and uranyl nitrate. By calcining under a reducing atmosphere the ammonium plutonate uranate precipitate formed, a mixed oxide (U,Pu)O.sub.2 is obtained having a solubility in 11N HN O.sub.3 without fluoride of at least 99.9%.

Abstract: Not readily dissolvable nuclear fuel is dissolved in an airtight closed vessel or autoclave. Nitric acid is evaporated and condensed in the vessel and the nitric acid condensate trickles over the nuclear fuel disposed outside of the nitric acid undergoing evaporation.

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 sintered nuclear fuel compact of UO.sub.2 or the mixed oxides (U, Pu)O.sub.2 and (U, Th)O.sub.2, with which reactivity losses in a nuclear reactor having relatively long fuel element cycles are avoided, has in its sintered matrix neutron poison in the chemical compound form UB.sub.x with x=2; 4 and/or 12 and/or B.sub.4 C. A sintered nuclear fuel compact of this kind is produced by sintering from a compact comprising a mixture of at least one of the mixture components UO.sub.2, PuO.sub.2, (U, Pu)O.sub.2 and (U, Th)O.sub.2 powder with UB.sub.x powder, where x=2; 4 and/or 12 and/or B.sub.4 C powder.

Abstract: Sintered nuclear fuel bodies are manufactured by pressing a powder of UO.sub.2 containing 1-20 percent by weight Gd.sub.2 O.sub.3 and up to 20 percent by weight U.sub.3 O.sub.8 into a pressed body and sintering the pressed body, the Gd.sub.2 O.sub.3 at least partly consisting of monoclinic Gd.sub.2 O.sub.3. The monoclinic Gd.sub.2 O.sub.3 used can be manufactured by heat treatment of cubic Gd.sub.2 O.sub.3 at a temperature of at least 1250.degree. C. By employing monoclinic Gd.sub.2 O.sub.3 a considerably increased homogeneity with respect to solid solution of gadolinium in the UO.sub.2 structure is obtained. Particularly favorable results are obtained if the sintering is carried out in a sintering atmosphere which, at 1750.degree. C., has an oxygen partial pressure which is higher than 10.sup.-7 atm.

Abstract: An automated fuel rod production system includes a radioactive powder fabrication and processing stage, a pellet fabrication stage, a pellet processing stage, a tube preparation stage and a fuel rod fabrication and inspection stage, all of which provide a continuous (paced) mode of operation from the conversion of a radioactive gas to powder, through the fabrication of the powder into pellets, to completion of the assembly of the fuel rods. Extra capacity is designed into the system at critical points in the powder processing and pellet fabrication and processing stages to facilitate the continuous, paced mode of operation.

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: Production of mixed-oxide nuclear fuel pellets of uranium dioxide (UO.sub.2+x) and plutonium dioxide (PuO.sub.2) which are soluble in nitric acid. The powder mixture of oxides is added to a halogen-free milling aid, such as propane diol, then is milled and subsequently granulated, pressed and sintered. The milling aid is expelled by subsequent drying processes or heat treatments (sintering). The milling process is controlled to obtain a primary grain size of less than two micrometers.

Abstract: UO.sub.2 base powder exhibiting any specific surface and crystallite diameter properties is mixed with rare earth (SE) oxide-containing powder, the particles of which exhibit at least in one surface layer, a crystal lattice of the fluorite type, with the stoichiometric composition (SE.sub.0.5, U.sub.0.5) 0.sub.2.00 and/or form it in sintering; and is compacted to form compacts which are sintered in a gas atmosphere with reducing action at 1500.degree. C. to 1750.degree. C. to form high-density sintered bodies.

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 simplified internal gelation process for the preparation of gel spheres of nuclear fuels. The process utilizes perchloroethylene as a gelation medium. Gelation is accomplished by directing droplets of a nuclear fuel broth into a moving volume of hot perchloroethylene (about 85.degree. C.) in a trough. Gelation takes place as the droplets float on the surface of the perchloroethylene and the resultant gel spheres are carried directly into an ager column which is attached to the trough. The aged spheres are disengaged from the perchloroethylene on a moving screen and are deposited in an aqueous wash column.

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: The invention relates to a process for the treatment of a metal oxide powder, particularly a UO.sub.2 or PuO.sub.2 powder, with a view to its fritting.This process consists of oxidizing the powder, so as to form on its surface a hydrated oxide layer of the same metal. When the oxide is UO.sub.2 or PuO.sub.2 and mixtures thereof, as well as those prepared with ThO.sub.2, the treated powder can be used for producing nuclear fuel pellets by cold compression, followed by fritting.