Abstract: The known fully ceramic microencapsulated fuel (FCM) entrains fission products within a primary encapsulation that is the consolidated within a secondary ultra-high-temperature-ceramic of Silicon Carbide (SiC). In this way the potential for fission product release to the environment is significantly limited. In order to extend the performance of this fuel to higher temperature and more aggressive coolant environments, such as the hot-hydrogen of proposed nuclear rockets, a zirconium carbide matrix version of the FCM fuel has been invented. In addition to the novel nature to this very high temperature fuel, the ability to form these fragile TRISO microencapsulations within fully dense ZrC represent a significant achievement.

Abstract: What are described are a process for producing an insulating product for the construction materials industry or an insulating material as intermediate for production of such a product, and a corresponding insulating material/insulating product. Also described are the use of a matrix encapsulation method for production of composite particles in the production of an insulating product for the construction materials industry or of an insulating material as intermediate for production of such a product, and the corresponding use of the composite particles producible by means of a matrix encapsulation method.

Abstract: The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

Abstract: The present invention relates to a method for preparing a fully ceramic capsulated nuclear fuel material containing three-layer-structured isotropic nuclear fuel particles coated with a ceramic having a composition which has a higher shrinkage than a matrix in order to prevent cracking of ceramic nuclear fuel, wherein the three-layer-structured nuclear fuel particles before coating is included in the range of between 5 and 40 fractions by volume based on after sintering.

Abstract: The present disclosure provides systems and methods for fast molten salt reactor fuel-salt preparation. In one implementation, the method may comprise providing fuel assemblies having fuel pellets, removing the fuel pellets and spent fuel constituents from the fuel assemblies, granulating the removed fuel pellets or process feed to a chlorination process, processing the granular spent fuel salt into chloride salt by ultimate reduction and chlorination of the uranium and associated fuel constituents chloride salt solution, enriching the granular spent fuel salt, chlorinating the enriched granular spent fuel salt to yield molten chloride salt fuel, analyzing, adjusting, and certifying the molten chloride salt fuel for end use in a molten salt reactor, pumping the molten chloride salt fuel and cooling the molten chloride salt fuel, and milling the solidified molten chloride salt fuel to predetermined specifications.

Abstract: A process for dissolving nuclear fuel, in particular irradiated nuclear fuel, comprising immersion of the nuclear fuel in a nitric acid solution. This dissolution process further comprises mechanical milling of the nuclear fuel, this mechanical milling being performed in the nitric acid solution during the immersion. The disclosure also relates to the use of a mill equipped with mechanical milling structure to implement the dissolution process.

Abstract: The present disclosure generally relates to additive manufacturing systems and methods on a large-scale format. One aspect involves a build unit that can be moved around in three dimensions by a positioning system, building separate portions of a large object. The build unit has an energy directing device that directs, e.g., laser or e-beam irradiation onto a powder layer. In the case of laser irradiation, the build volume may have a gasflow device that provides laminar gas flow to a laminar flow zone above the layer of powder. This allows for efficient removal of the smoke, condensates, and other impurities produced by irradiating the powder (the “gas plume”) without excessively disturbing the powder layer. The build unit may also have a recoater that allows it to selectively deposit particular quantities of powder in specific locations over a work surface to build large, high quality, high precision objects.

Abstract: The invention provides a sintered rare earth metal oxide target for producing a radioisotope in an instrumentation tube of a nuclear power reactor, wherein the sintered target has a density of at least 90 percent of the theoretical density, and wherein the sintered target contains chromium in an amount of from 500 to 2000 ?g/g, and Mg and/or Ca in an amount of from 1000 to 6000 ?g/g. The sintered target is prepared by providing a rare earth metal oxide powder, blending the rare earth metal oxide powder with chromium oxide, dry granulating and consolidating the powder in a mold to form a spheroidal green body, and sintering the green body in solid phase to form a spheroidal ytterbia target.

Abstract: Solid immiscible alloys and methods for making the solid immiscible alloys are provided. The microstructure of the immiscible alloys is characterized by a minority phase comprising a plurality of particles of an inorganic material dispersed in a majority phase comprising a continuous matrix of another inorganic material. The methods utilize nanoparticles to control both the collisional growth and the diffusional growth of the minority phase particles in the matrix during the formation of the alloy microstructure.

Abstract: A sliding mechanism 1 includes a first sliding member 10 and a partner second sliding member 20 configured to slide relative to the first sliding member 10. The first sliding member 10 includes a matrix phase 11 and a hard phase 13 that is harder than the matrix phase 11, in which the hard phase 13 is embedded in the matrix phase 11 in a dispersed state. The second sliding member 20 includes a base 21 and a surface-treatment layer 23 that is formed on the sliding surface 20 of the base 21 and is harder than the matrix phase 11.

Abstract: A pyrolysis reactor includes a chamber having an inactive section and an active section. The inactive section is configured to hold an inactive pre-form capable of adhering carbon. The active section is configured to hold an active pre-form capable of adhering carbon. An induction coil is outside of and operatively adjacent the active section, and wherein the active section is configured to pyrolyze a hydrocarbon.

Abstract: A radiative recuperator preheats oxidant and/or fuel for combustion at one or more burners of a furnace. The recuperator includes a duct, at least portions of which comprise a material having a thermal conductivity of greater than 1 W/(m·K), preferably greater than 3 W/(m·K), that receives hot flue gas produced by the burner(s). The duct radiatively transfers heat to oxidant or fuel (for preheating) flowing through one or more metallic pipes disposed in between the duct and an insulating wall.

Abstract: Targetry coupled separation refers to enhancing the production of a predetermined radiation product through the selection of a target (including selection of the target material and the material's physical structure) and separation chemistry in order to optimize the recovery of the predetermined radiation product. This disclosure describes systems and methods for creating (through irradiation) and removing one or more desired radioisotopes from a target and further describes systems and methods that allow the same target to undergo multiple irradiations and separation operations without damage to the target. In contrast with the prior art that requires complete dissolution or destruction of a target before recovery of any irradiation products, the repeated reuse of the same physical target allowed by targetry coupled separation represents a significant increase in efficiency and decrease in cost over the prior art.

Abstract: Solid immiscible alloys and methods for making the solid immiscible alloys are provided. The microstructure of the immiscible alloys is characterized by a minority phase comprising a plurality of particles of an inorganic material dispersed in a majority phase comprising a continuous matrix of another inorganic material. The methods utilize nanoparticles to control both the collisional growth and the diffusional growth of the minority phase particles in the matrix during the formation of the alloy microstructure.

Abstract: A fuel pellet for a nuclear reactor includes a plurality of tristructural-isotropic fuel particles embedded in a structural silicon carbide matrix. A method of manufacturing a fuel pellet includes the steps of coating a plurality of tristructural-isotropic fuel particles with a coating slurry including silicon carbide powder to form a plurality of coated fuel particles; compacting the plurality of fuel particles; and sintering the compacted plurality of fuel particles to form the fuel pellet.

Abstract: A method for immobilizing liquid radioactive waste is provided, the method having the steps of mixing waste with polymer to form a non-liquid waste; contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to cause the polymer, waste, and filler to irreversibly bind in a solid phase, and compressing the solid phase into a monolith. The invention also provides a method for immobilizing liquid radioactive waste containing tritium, the method having the steps of mixing liquid waste with polymer to convert the liquid waste to a non-liquid waste, contacting the non-liquid waste with a solidifying agent to create a mixture, heating the mixture to form homogeneous, chemically stable solid phase, and compressing the chemically stable solid phase into a final waste form, wherein the polymer comprises approximately a 9:1 weight ratio mixture of styrene block co-polymers and cross linked co-polymers of acrylamides.

Abstract: The embodiments relate to a method for the production of a refractory metal component by casting. The method includes providing a slip that contains a powder including at least one refractory metal or a compound thereof, in addition to at least one binding agent. The method further includes processing the slip by casting, (e.g., film casting or slip casting), to form at least one slip coating, the slip being devoid of a metal binding agent. A component was produced by this method. The embodiments may be used, in particular, on X-ray tubes, accelerator targets, or fusion reactors, such as for a surface of an X-ray anode, or a wall of a fusion reactor.

Abstract: A nuclear fuel includes a volume of a nuclear fuel material defined by a surface, the nuclear fuel material including a plurality of grains, some of the plurality of grains having a characteristic length along at least one dimension that is smaller than or equal to a selected distance, wherein the selected distance is suitable for maintaining adequate diffusion of a fission product from a grain interior to a grain boundary in some of the grains, the nuclear fuel material including a boundary network configured to transport the fission product from at least one grain boundary of some of the grains to the surface of the volume of the nuclear fuel material.

Abstract: In a method for forming joints between MgB2 filaments of superconducting wires, the MgB2 filaments from the wires to be joined are exposed, and the exposed filaments are then exposed to a mixture of magnesium powder and boron powder in a furnace, and the MgB2 filaments and the magnesium and boron powders are pressed together in the furnace. The MgB2 filaments and the magnesium and the boron powders in the furnace are heated, and oxygen that is present within the furnace is preferentially trapped, and thus removed from the joint, by providing titanium within the furnace.

Abstract: Provided is a selective catalytic reduction (SCR) catalyst containing a carbon material loaded with vanadium and tungsten and a method of preparing the same, and relates to a method of loading vanadium and tungsten on a carbon material that exhibits excellent abrasion resistance and excellent strength and can be easily prepared.

Abstract: A method of making dispersion-strengthened alloy particles involves melting an alloy having a corrosion and/or oxidation resistance-imparting alloying element, a dispersoid-forming element, and a matrix metal wherein the dispersoid-forming element exhibits a greater tendency to react with a reactive species acquired from an atomizing gas than does the alloying element. The melted alloy is atomized with the atomizing gas including the reactive species to form atomized particles so that the reactive species is (a) dissolved in solid solution to a depth below the surface of atomized particles and/or (b) reacted with the dispersoid-forming element to form dispersoids in the atomized particles to a depth below the surface of said atomized particles. The atomized alloy particles are solidified as solidified alloy particles or as a solidified deposit of alloy particles.

Abstract: An apparatus for manufacturing an article from powder material including a first table, a second table rotatably mounted on the first table about a first axis and a third table rotatably mounted on the second table about a second axis. A hollow canister is supported by the third table. A vibrator is arranged to vibrate the canister. A first device is arranged to rotate the second table about the first axis and a second device is arranged to rotate the third table about the second axis. A hopper is arranged to supply powder material into the canister and a valve controls the flow of powder material from the hopper into the canister. A processor is arranged to control the valve, the vibrator, the first device and the second device to control the filling and packing density of the canister.

Abstract: A new interface between the cladding and the stack of pellets in a nuclear control rod. According to the invention, an interface joint made of a material transparent to neutrons, in the form of a structure with a high thermal conductivity and open pores, adapted to deform by compression across its thickness, is inserted between the cladding and the stack of pellets made of B4C neutron absorber material over at least the height of the stack. The invention also relates to associated production methods.

Abstract: This invention relates to a method of preparing a nuclear fuel including the step of depositing a coating which includes fluorine, or at least one compound thereof, around a kernel (12) of fissile material. The invention extends to a coated nuclear fuel particle (10).

Abstract: A method of making a multilayer article includes providing a mold defining a first cavity with a removable element therein; placing a first particulate component in the cavity; compressing the component along a first axis in a first compression step to form a first layer of the article; removing the element from the cavity after formation of the first layer; placing one or more additional particulate components in the mold; compressing the one or more additional particulate components along the first axis in one or more additional compression steps after removal of the element to form one or more additional layers of the article, wherein the first layer meets one of the one or more additional layers along a first interface substantially parallel to the first axis and the first layer meets one of the one or more additional layers along a second interface substantially perpendicular to the first axis.

Abstract: A method for fabricating porous UO2 sintered pellets to be fed into the electrolytic reduction process for the purpose of metallic nuclear fuel recovery is provided, which includes forming a powder containing U3O8 by oxidizing spent nuclear fuel containing uranium dioxide (UO2) (step 1), fabricating green pellets by compacting the powder formed in step 1 (step 2), fabricating UO2+x sintered pellets by sintering the porous U3O8 green pellets fabricated in step 2 at 1200 to 1600° C., in an atmospheric gas (step 3), and forming UO2 sintered pellets by cooling the UO2+x sintered pellets to room temperature, and reduction the same at 1000 to 1400° C., in a reducing atmosphere (step 4).

Abstract: A process of producing ceramic-ceramic composites, including but not limited to nuclear fuels, and composites capable of exhibiting increased thermal conductivities. The process includes milling a first ceramic material to produce a powder of spheroidized particles of the first ceramic material, and then co-milling particles of a second ceramic material with the spheroidized particles of the first ceramic material to cause the particles of the second ceramic material to form a coating on the spheroidized particles of the first material. The spheroidized particles coated with the particles of the second ceramic material are then compacted and sintered to form the ceramic-ceramic composite, in which the second ceramic material forms a continuous phase completely surrounding the spheroidized particles of the first ceramic material.

Abstract: A mixture of fine powder including thorium oxide was converted to granulated powder by forming a first-green-body and heat treating the first-green-body at a high temperature to strengthen the first-green-body followed by granulation by crushing or milling the heat-treated first-green-body. The granulated powder was achieved by screening through a combination of sieves to achieve the desired granule size distribution. The granulated powder relies on the thermal bonding to maintain its shape and structure. The granulated powder contains no organic binder and can be stored in a radioactive or other extreme environment. The granulated powder was pressed and sintered to form a dense compact with a higher density and more uniform pore size distribution.

Abstract: Process for the continuous production of chlorine dioxide comprising generating chlorine dioxide in an aqueous reaction medium in a reaction vessel (1) maintained at sub-atmospheric pressure, bringing gaseous chlorine dioxide from said reaction vessel to an absorption tower (7) and contacting it therein with a flow of water to form an aqueous solution containing chlorine dioxide, bringing said aqueous solution containing chlorine dioxide to a stripper (12), blowing a gas through said aqueous solution of chlorine dioxide in the stripper to strip off from 10 to 100% of the chlorine dioxide entering the stripper and form a gaseous chlorine dioxide product.

Abstract: Provided are a method of controlling the compositional gradient and solubility of doped-additives at grain boundaries during sintering of a uranium-based oxide green pellet including the additives, and a method of manufacturing a sintered nuclear fuel pellet having a large grain size using the same. The grain boundary solubility of the doped-additives is maintained at a certain level by stepwise varying of an oxygen partial pressure during isothermal sintering of a uranium-based oxide green pellet including the additives. The method of manufacturing a sintered nuclear fuel pellet having a large grain size includes preparing additive mixed uranium oxide powder, forming an additive mixed uranium oxide green pellet using the mixed powder, heating the green pellet to a sintering temperature in a gas atmosphere having a low oxygen partial pressure, and sintering while a sintering gas atmosphere is changed to stepwise increase an oxygen partial pressure at the isothermal sintering temperature.

Abstract: A high energy density fuel source that reduces hydride expansion during hydrogen release, including a rigid, thermally insulated container defining an internal volume, a heater mechanism disposed within the internal volume, and a metal hydride rod thermally connected to the heater mechanism, wherein the heater mechanism and metal hydride rod substantially occupying the entirety of the internal volume. The metal hydride rod preferably includes a malleable encapsulation compressed about, thermally coupled to, and substantially encapsulating a volume of compressed metal hydride powder, the malleable encapsulation defined by a first, thermally conductive, malleable cup inverted over a second, thermally conductive, malleable cup, the compressed malleable encapsulation defining a tortuous fluid flow path from the metal hydride to the internal volume of the container.

Abstract: To provide an inorganic fiber that suppresses adverse effects on a human body and living environments, exhibits high biosolubility, and also exhibits excellent heat resistance as a constituent material for a filter material, a sealing material, or the like. The inorganic fiber comprises 30 mass % or more and less than 81 mass % of Al2O3, more than 19 mass % and 65 mass % or less of MgO and 0 mass % to 40 mass % of SiO2, wherein the total content of Al2O3, MgO and SiO2 relative to the entire fiber is 98 mass % or more.

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: 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: A method for fabricating a sintered annular nuclear fuel pellet includes molding nuclear fuel powder or granule, an oxide of a fissile element (M), to fabricate an annular nuclear fuel green body. A rod-like shaped structure is inserted into the annular nuclear fuel green body and sintered in a slight oxidizing gas atmosphere such that the oxide of the fissile element has a balanced O/M ratio higher than a desired O/M ratio (oxygen/fissile element) of a final sintered annular nuclear fuel pellet, while being maintained in a cubic phase. The sintered annular nuclear fuel pellet is then reduced in a reductive gas atmosphere so as to have the desired O/M ratio in the state that the rod-like shaped structure is inserted.

Abstract: A method for fabricating a sintered annular nuclear fuel pellet includes: molding nuclear fuel powder or granules to fabricate an annular nuclear fuel green body; inserting a rod-like shaped structure into the annular nuclear fuel green body; sintering the rod-like shaped structure-inserted annular nuclear fuel green body in a reductive gas atmosphere; and separating the sintered annular nuclear fuel pellet from the rod-like shaped structure.

Abstract: A method for the production of spherical combustible or fertile material particles from an oxide of the group of the heavy metals uranium, plutonium, or mixtures thereof. For this purpose, the process steps of producing a base solution of the nitrates of the heavy metal(s), adding at least one first reagent in order to adjust the viscosity of the solution, dropping the solution to form microspheres, at least superficially solidifying the microspheres in an atmosphere containing ammonia, collecting the microspheres in a solution containing ammonia, and subsequent washing, drying and thermal treatment are carried out, where at least one of urea, ammonium carbonate, ammonium hydrogen carbonate, ammonium cyanate, and biuret are added to the base solution before adding the first reagent. The solution thus prepared is heated to a temperature T where 80° C.?T<Ts and where Ts=boiling temperature of the solution, and is maintained at the temperature over a time period t where 2 h?t?8 h.

Abstract: A metal carbide ceramic fiber having improved mechanical properties and characteristics and improved processes and chemical routes for manufacturing metal carbide ceramic fiber. Metal carbide ceramic fibers may be formed via reaction bonding of a metal-based material (e.g. boron) with the inherent carbon of a carrier medium. One embodiment includes a method of making a metal carbide ceramic fiber using VSSP to produce high yield boron carbide fiber. Embodiments of the improved method allow high volume production of high density boron carbide fiber. The chemical routes may include a direct production of boron carbide fiber from boron carbide powder (B4C) and precursor (e.g. rayon fiber) having a carbon component to form a B4C/rayon fiber that may be processed at high temperature to form boron carbide fiber, and that may be subsequently undergo a hot isostatic pressing to improve fiber purity. Another route may include a carbothermal method comprising combining boron powder (B) with a precursor (e.g.

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 method for manufacturing an annular nuclear fuel pellet is provided. In the method, an annular nuclear fuel green compact whose lateral cross-section is a trapezoid is prepared. The thickness of the annular nuclear fuel green compact reduces along one direction of the central axis, and a green density of the nuclear fuel green compact increases along one direction of the central axis. The annular nuclear fuel green compact is sintered under a reducing gas atmosphere so that the annular nuclear fuel pellet is obtained. According to this method, the annular pellet which has uniform inner and outer diameters and small diametric tolerances along the pellet height is fabricated without grinding the pellet surfaces.

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: Production device including a press, a conveyor (4) intended to transport pellets from the press to a sintering area, means (26) for transferring pellets from the press to the conveyor (4), and means for inspecting at least one nuclear fuel pellet at the exit of a press, said inspection means including means (20) for identifying each pellet according to its die of origin.

Abstract: The object of the present invention is mainly a powder dispenser including a casing capable of impulsing the powder in a reciprocal movement on a plane along a determined displacement direction (X), and means (19,23) for grouping the powder along distinct axes, substantially parallel to the X direction. The object of the present invention is also a method for making pellets, notably nuclear fuel pellets, applying a dispenser according to the present invention.

Abstract: The present invention relates to the use of a mixed carbonate of formula AB(CO3)2, in which A and B are different and chosen from alkali metals, alkaline-earth metals and rare earths, for the containment of radioactive carbon. This use may for example involve a process comprising: mixing CO2 having a radioactive carbon to be contained, or a simple carbonate of an alkali, alkaline-earth or rare-earth metal having a radioactive carbon to be contained, with an aqueous solution of a mixture of ACln and BClm or with an aqueous solution of a mixture of A(OH)n, and B(OH)m in order to obtain a precipitate of AB(CO3)2, where n and m are integers sufficient to compensate for the charge of A and B respectively; recovery of the AB(CO3)2 precipitate in powder form; and then pressing and sintering of the powder at a 20 temperature below the decarbonation temperature of the mixed carbonate manufactured in order to obtain sintered pellets of mixed carbonates for the containment of the radioactive carbon.

Abstract: Microstructured nuclear fuel adapted for nuclear power system use includes fissile material structures of micrometer-scale dimension dispersed in a matrix material. In one method of production, fissile material particles are processed in a chemical vapor deposition (CVD) fluidized-bed reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. At least one pneumatic or electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. An exhaust associated with the particle coating chamber and can provide a port for placement and removal of particles and powder. During use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and come to rest in the matrix material.

Abstract: A method for fabricating a sintered duplex nuclear fuel pellet includes the steps of: preparing a first powder composed of a material selected from the group consisting of UO2 and UO2—Er2O3, and a second powder composed of UO2—Gd2O3 and a sintering additive; producing a duplex compact consisting of an annular outer portion composed of the first powder and a cylindrical inner portion composed of the second powder; and sintering the duplex compact under a reducing gas atmosphere, wherein the sintering additive contains manganese of 0.001% to 2% by weight based on the total weight of the cylindrical inner portion.

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: To furnish stably a sintered nuclear fuel compact of uranium dioxide with a large grain diameter, a fabrication method therefor requires sintering a starting material at two stages in an oxidizing atmosphere and in a reducing atmosphere at relatively low temperatures. The starting material is either a powder mixture of uranium dioxide fresh material powder and triuranium octaoxide in a weight ratio of 75/25 ˜ 55/45 or pulverulent mixed uranium oxides yielding a final O/U ratio equal to that of the powder mixture. The sintering process in the oxidizing atmosphere involves at least one time period of a temperature rise to a higher temperature of 1200 to 1100° C. and at least one time period of a temperature drop to a lower temperature of up to 1080° C. A temperature difference between the higher and lower temperatures is preferred to be 50° C. or more.

Abstract: A nuclear fuel sintered body is produced from a powder which contains at least one fissile heavy metal oxide. During the further treatment of the powder over the course of the process preceding the sintering operation, a dopant that contains at least 100 ppm of an iron oxide compound is added to the powder. The powder is a UO2-containing powder obtained from a dry-chemical conversion process, and if appropriate, a powder which contains further fissile heavy metal oxide (U3O8, PuO2, inter alia). As a result, the sintered body is provided with high plasticity combined, at the same time, with a large grain size. This advantageously reduces an interaction between the nuclear fuel sintered body and a fuel rod cladding tube during an operation of the reactor.