Abstract: A method of preparing active, sinterable, finely-divided plutonium oxide (PuO.sub.2) powder from plutonium metal is disclosed. The process yields plutonium fissile material which can be easily blended to form a uniformly homogeneous powder for the fabrication of high-quality light water reactor ceramic fuel pellets. Such homogeneous fuels are required to prevent hot spots from developing in a reactor using the fuel.

Abstract: A method for producing UO.sub.2 or (U/Pu)O.sub.2 powder includes obtaining ammonium uranyl carbonate or ammonium uranyl plutonyl carbonate by treating a starting oxide selected from the group consisting of uranium oxide, plutonium oxide and uranium plutonium mixed oxide, with at least one solution selected from the group consisting of aqueous ammonium carbonate solution and aqueous ammonium hydrogen carbonate solution. The ammonium uranyl carbonate or ammonium uranyl plutonyl carbonate is then heated and in particular calcined.

Abstract: The invention is a specialized reaction cell for converting uranium metal to uranium oxide. In a preferred form, the reaction cell comprises a reaction chamber with increasing diameter along its length (e.g. a cylindrical chamber having a diameter of about 2 inches in a lower portion and having a diameter of from about 4 to about 12 inches in an upper portion). Such dimensions are important to achieve the necessary conversion while at the same time affording criticality control and transportability of the cell and product. The reaction chamber further comprises an upper port and a lower port, the lower port allowing for the entry of reactant gasses into the reaction chamber, the upper port allowing for the exit of gasses from the reaction chamber. A diffuser plate is attached to the lower port of the reaction chamber and serves to shape the flow of gas into the reaction chamber.

Abstract: A multichamber type fluid bed reaction apparatus has a series of chambers formed by dividing the fluid bed part and the wind box part thereof with particular walls. Each reaction chamber is supplied with a reaction gas whose composition, flow amount and temperature can be selected arbitrarily. A fluid bed, a moving bed or a fixed bed may be used in each divided chamber. Regulation of particle formation, gas-particle reaction and amount of particles transferring between the chambers results in an improvement of efficiency and a simplification of the apparatus.

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: A process of preparing an actinide compound of the formula An.sub.x Z.sub.y wherein An is an actinide metal atom selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, x is selected from the group consisting of one, two or three, Z is a main group element atom selected from the group consisting of nitrogen, phosphorus, oxygen and sulfur and y is selected from the group consisting of one, two, three or four, by admixing an actinide organometallic precursor wherein said actinide is selected from the group consisting of thorium, uranium, plutonium, neptunium, and americium, a suitable solvent and a protic Lewis base selected from the group consisting of ammonia, phosphine, hydrogen sulfide and water, at temperatures and for time sufficient to form an intermediate actinide complex, heating said intermediate actinide complex at temperatures and for time sufficient to form the actinide compound, and a process of depositing a thin film of such an actinide compound, e.g.

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 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: Ceramic uranium dioxide is produced from uranium metal by steam oxidation of the metal, followed by oxidation in air to produce U.sub.3 O.sub.8, and subsequent reduction in hydrogen of the U.sub.3 O.sub.8 to produce UO.sub.2. The steam environment may be between 250.degree. C. and 500.degree. C., the oxygen-containing environment at 700.degree. C. or below, and the reducing environment at 600.degree. C. or below.

Abstract: Uranium dioxide nuclear fuel pellets are produced using a ceramic grade uranium dioxide powder derived from the reaction of uranium hexafluoride with steam and hydrogen. The powder is pelletized and the green pellets are transported through a furnace and subjected to fast-fired sintering typically at temperature of the order of 2000-2200.degree. C. and for a time interval between 50 to 500 seconds. It has been found that fast-fired sintering of such powders under these conditions enables densities of the order of 95% theoretical density to be achieved, with resinter density changes of the order of 2% and lower, without requiring any mechanical compression of the hot pellets. The inclusion of seed crystals of uranium dioxide in the pelletized powder increases the grain size of the fuel pellets.

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 method of controlling the crystal grain size of UO.sub.2 pellets is disclosed, which comprises(1) providing an aqueous uranyl solution which is free of hydrofluoric acid and nitric acid and which contains uranyl fluoride (UO.sub.2 F.sub.2) and/or uranyl nitrate (UO.sub.2 (NO.sub.3).sub.2) as a uranium component,(2) reacting with ammonia said aqueous uranyl solution to precipitate ammonium diuranate (ADU), while adjusting the ratio of said uranium components to a predetermined value within the range varying from the ratio 100% of uranyl fluoride and 0% of uranyl nitrate to the ratio 0% of uranyl fluoride and 100% of uranyl nitrate, adjusting the uranium concentration of the reaction mixture to 50 to 1,000 gU/liter and also adjusting the rate of contact between said uranium component in said aqueous uranyl solution with ammonia to at least 2 moles NH.sub.3 /min/mole U,(3) calcining and reducing said ADU precipitate to form UO.sub.2 powder, and(4) molding and sintering said UO.sub.2 powder, thus producing UO.

Abstract: A mixture of uranium dioxide and additive of aluminosilicate composition is formed into a compact and sintered to produce a nuclear fuel wherein the uranium dioxide grains have an average grain size of at least about 20 microns and wherein substantially all of the grains are each enveloped by glassy aluminosilicate phase.

Abstract: A particulate mixture of uranium dioxide and additive of magnesium aluminosilicate composition is formed into a compact and sintered to produce a nuclear fuel wherein the uranium dioxide grains have an average grain size of at least about 20 microns and wherein substantially all of the grains are each enveloped with glassy magnesium aluminosilicate phase.

Abstract: A particulate mixture of uranium dioxide and additive of magnesium silicate composition is formed into a compact and sintered to produce a nuclear fuel wherein the uranium dioxide grains have an average grain size of at least about 20 microns and wherein substantially all of the grains are each enveloped with glassy magnesium silicate phase.

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: Matter is floated to the surface of a liquid by bonding ions to the surface of the matter to give the matter a charge, and forming a froth with the aid of a frothing agent having groups of opposite charge to the ions so that the frothing agent bonds to the matter and is carried in the froth to the surface of the liquid. By removing the froth the matter can be separated from any inert matter present in the liquid. The oxidation state of the surface of the matter may be changed before bonding takes place with the ions to one which facilitates that bonding. The matter can be particulate or dissolved ions. For example, uranium dioxide particles are oxidized with hydrogen peroxide, sodium carbonate added to produce a negatively charged uranyl carbonate complex and a froth formed with the aid of cetyl trimethylammonium bromide. Cationic groups in the latter bond to the uranyl carbonate complex causing the uranyl carbonate complex to be concentrated in the froth at the surface of the liquid.

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 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: 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: 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: 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: 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 and apparatus for separating uranium-containing particulate matter from the off-gas stream of a calciner, without condensation of ammonium fluoride from the gas stream along with the particulate matter, wherein the off-gas is passed through a filter while maintained at a temperature in excess of 340.degree. C. Preferably, the off-gas stream is forced through the filter by an aspirator, located downstream of the filter, and the filter comprises a plurality of filter units in side-by-side relationship, each of said filter units being purged by a nitrogen gas stream while the off-gas stream continues to flow through the remainder of the filter units for a continuous filtering operation.

Abstract: A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide.

Abstract: An improved process for producing powdered uranium dioxide from a solution of uranyl nitrate which is suitable for the manufacture of fuel for nuclear reactors. The process is continuous and comprises an incremental precipitation of soluble uranyl nitrate with ammonium hydroxide which is interrupted with an intermediate aging period. The precipitate of ammonium uranate solids is dried and thermally converted to a powdered oxide of uranium.

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: Method for the manufacture of oxidic nuclear fuel bodies by a heat treatment of blanks obtained from UO.sub.2 starting power or a mixture of UO.sub.2 and PuO.sub.2 starting powder at a treatment temperature in the range of 1000.degree. C. to 1400.degree. initially in a gas atmosphere with oxidizing action and subsequently in a gas atmosphere with reducing action. The oxygen potential of the gas atmosphere with oxidizing action is kept in a range in which a crystallographically demonstrable U.sub.4 O.sub.9 or (U, Pu).sub.4 O.sub.9 crystal phase is generated in the blanks during the heating to the treatment temperature in this gas atmosphere with oxidizing action.

Abstract: An improved method for compression molding typically brittle, particulate ceramic materials comprising uranium dioxide in the manufacture of nuclear fuel products.

Abstract: A process for manufacturing uranium dioxide powder which comprises forming fine uranium dioxide powder having a high sinterability and coarse uranium dioxide powder having a low sinterability continuously in one process by changing periodically the precipitation condition of the ammonium diuranate under the same condition of calcining and reducing of the ammonium diuranate. The thus obtained mixture of these uranium dioxide powders is suitable for uranium dioxide pellet which is a fuel of nuclear power reactor.

Abstract: A nuclear fuel conversion system comprises a unit for heating and denitrifying a nitric acid solution of uranyl nitrate, plutonium nitrate, or mixture thereof, a unit for grinding denitrified products, a unit connected to the grinding unit through a first conveying unit for roasting and reducing ground products, a unit connected to the roast-reduction unit through a second conveying unit for further grinding and then sieving the roasted and reduced products, and a unit for disposing of waste gases and waste liquor. The denitrification unit includes a microwave generator for heating and denitrifying the nitric acid solution and for cooling the denitrified product, and the first and second conveying units comprise pneumatic carrier means each of which consists of a suction nozzle, cyclone connected to the nozzle through a pipe, and a blower for creating a reduced pressure condition in the pipe.

Abstract: A method of recovering uranium dioxide from a sodium uranyl carbonate solution obtained by the alkaline carbonate leaching of uranium ore in which a solution is reacted at a temperature above 130.degree. C. and at superatmospheric pressure with particular metallic iron. The precipitated UO.sub.2 is recovered from the solution.

Abstract: A method is disclosed for producing a dimensionally stable UO.sub.2 fuel pellet of large grain mole %, and relatively large pore size. A dopant containing an element selected from the group consisting of aluminum, calcium, magnesium, titanium, zirconium, vanadium, niobium, and mixtures thereof is added to a highly sinterable UO.sub.2 powder, which is a UO.sub.2 powder that is sinterable to at least 97% theoretical density at 1600.degree. C. in one hour, and the UO.sub.2 powder can then be formed into fuel pellets. Alternatively, the dopant can be added at a step in the process of producing ammonium diuranate. The ammonium diuranate is collected with about 0.05 to about 1.7 mole%, based on UO.sub.2, of a compound containing said element. That mixture is then calcined to produce UO.sub.2 and the UO.sub.2 is formed into a fuel pellet. The addition of the dopant can also be made at the hydrolysis stage in the manufacture of UO.sub.2 by a dry conversion process.

Abstract: A nuclear fuel material green body of density from about 30 to 70% of theoretical density having tensile strenght and plasticity adequate to maintain the integrity of the body during processing leading to ultimate sintered condition is produced by adding an amine carbonate or carbamate or mixture thereof to a particulate mass of the nuclear fuel material under conditions resulting in reaction with the amine compound to form a water-soluble compound effective as a binder for the particulate material.

Abstract: The grain size of particulate uranium oxides is increased and adjusted by the application thereto of hydrogen peroxide. The procedure is useful in the manufacture of nuclear reactor fuel from powdered fissionable materials.

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: A nuclear fuel material green body of density from about 30 to 70% of theoretical density having tensile strength and plasticity adequate to maintain the integrity of the body during processing leading to ultimate sintered condition is produced by adding one or more amines to a particulate mass of the nuclear fuel containing about five percent of ammonium uranyl carbonate under conditions resulting in reaction of the amine with the ammonium uranyl carbonate, liberation of ammonia and formation of a water-soluble uranyl compound more effective as a binder than the ammonium uranyl carbonate.

Abstract: In a process for converting UF.sub.6 into UO.sub.2, the UF.sub.6 is brought into contact with an aqueous aluminum nitrate solution. The resultant product is solvent extracted with tributyl phosphate to remove uranyl nitrate. The raffinate has a fluorine/aluminum (F/Al) weight ratio within the range of from about 0.5 to about 1.2. A sufficient quantity of hydrofluoric acid is added to the raffinate to minimize the solubility of aluminum fluoride (AlF.sub.3) therein and thereby maximize the precipitation potential of AlF.sub.3. Generally this occurs when sufficient hydrofluoric acid has been added to cause the F/Al weight ratio to be within the range of from about 1.8 to about 2.2. As a result of this treatment, the raffinate is divided into an uranium-containing aqueous solution and an AlF.sub.3 precipitate which contains substantially no uranium.

Abstract: A method for the heat treatment of a radioactive substance by microwave power comprising preparing a nitrate solution of uranium, thorium, plutonium or a mixture thereof, and applying to the nitrate solution microwave energy sufficient to directly convert the nitrate solution into an oxide powder of uranium, thorium, plutonium or a mixture thereof, respectively. Such oxide powder is suitable for the manufacture of nuclear fuel pellets.