Abstract: One-step solution combustion synthesis (SCS) methods for fabricating durable crystalline transuranic-doped rare earth zirconium pyrochlores are described. Methods are fast, amenable to upscaling, and present a simple strategy for producing crystalline ceramic materials that meet the minimum attractiveness criteria for special nuclear material. The methods include analysis of reactants and reaction conditions to select proper fuel as well as proper fuel content so as to encourage formation of the crystalline product in a single-step synthesis procedure.

Abstract: The present disclosure is directed towards methods of making titanium diboride products in various sizes. An aspect of the method provides (a) selecting a target average particle size for a target titanium diboride product; (b) selecting at least one processing variable from the group consisting of: an amount of sulfur, an inert gas flow rate, a soak time, and a reaction temperature; (c) selecting a condition of the processing variable based upon the target average particle size; and (d) producing an actual titanium diboride product having an actual average particle size using the at least one processing variable, wherein due to the at least one processing variable, the actual average particle size corresponds to the target average particle size.

Abstract: This invention relates to a method for making strontium-phosphate microparticles that incorporate radioisotopes for radiation therapy and imaging.

Abstract: The invention relates to a process for reprocessing a spent nuclear fuel and for preparing a mixed uranium-plutonium oxide, which process comprises: a) the separation of the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution resulting from the dissolution of the fuel in nitric acid, this step including at least one operation of coextracting the uranium and plutonium from said solution by a solvent phase; b) the partition of the coextracted uranium and plutonium to a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but no plutonium; c) the purification of the plutonium and uranium that are present in the first aqueous phase; and d) a step of coconverting the plutonium and uranium to a mixed uranium/plutonium oxide. Applications: reprocessing of nuclear fuels based on uranium oxide or on mixed uranium-plutonium oxide.

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

Abstract: A method of stabilizing nuclear material is disclosed. Oxides or halides of actinides and/or transuranics (TRUs) and/or hydrocarbons and/or acids contaminated with actinides and/or TRUs are treated by adjusting the pH of the nuclear material to not less than about 5 and adding sufficient MgO to convert fluorides present to MgF2; alumina is added in an amount sufficient to absorb substantially all hydrocarbon liquid present, after which a binder including MgO and KH2PO4 is added to the treated nuclear material to form a slurry. Additional MgO may be added. A crystalline radioactive material is also disclosed having a binder of the reaction product of calcined MgO and KH2PO4 and a radioactive material of the oxides and/or halides of actinides and/or transuranics (TRUs). Acids contaminated with actinides and/or TRUs, and/or actinides and/or TRUs with or without oils and/or greases may be encapsulated and stabilized by the binder.

Abstract: The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as carbides and transition-metal, lanthanide and actinide oxides, using an aerodynamic levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

Abstract: A method of producing metal chlorides is disclosed in which chlorine gas is introduced into liquid Cd. CdCl2 salt is floating on the liquid Cd and as more liquid CdCl2 is formed it separates from the liquid Cd metal and dissolves in the salt. The salt with the CdCl2 dissolved therein contacts a metal which reacts with CdCl2 to form a metal chloride, forming a mixture of metal chloride and CdCl2. After separation of bulk Cd from the salt, by gravitational means, the metal chloride is obtained by distillation which removes CdCl2 and any Cd dissolved in the metal chloride.

Abstract: A method for separation of actinide elements comprising feeding a solution containing actinide elements such as americium, curium, californium and the like, into a resin column in which a weakly basic primary, secondary or tertiary anion exchange resin obtained by resinifying pyridine, imidazole or alkylamine has been packed, and then feeding an eluent of a mixed solution of nitric acid and alkyl alcohol such as methanol, ethanol, propanol and the like into the resin column to chromatographically separate the actinide elements from each other. This method makes it possible to efficiently separate the actinide elements from each other by a unit operation at ordinary temperature and ordinary pressure while avoiding oxidation operation, and hence makes it possible to avoid generation of secondary wastes and operations difficult in terms of engineering, such as precipitation.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.

Abstract: A reprocessing process of spent nuclear fuels for roughly separating U and U—Pu from FP, TRU and the like in a nitric acid solution of spent nuclear fuels by utilizing phenomenon of cocrystallization of hexavalent U and Pu. For example, spent nuclear fuels are sheared and dissolved in nitric acid, and insoluble residues in the nitric acid solution are removed. Then, a nitric acid concentration in the solution is adjusted and a valence of Pu in the solution is adjusted to tetravalence. The solution is then cooled to crystallize uranyl nitrate hydrate crystals and separated into the crystals and a mother liquor, and the separated crystals are recovered as a U product. Then, a nitric acid concentration in the separated mother liquor is adjusted and a valance of U and Pu in the mother liquor is adjusted to hexavalence, and the mother liquor is cooled to crystallize uranyl-plutonyl nitrate hydrate crystals which are separated and recovered as a U—Pu mixed product.

Abstract: The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as borides, carbides and transition-metal, lanthanide and actinide oxides, using an Aerodynamic Levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

Abstract: Apparatus and method for abatement of effluent from multi-component metal oxides deposited by CVD processes using metal source reagent liquid solutions which comprise at least one metal coordination complex including a metal to which is coordinatively bound at least one ligand in a stable complex and a suitable solvent medium for that metal coordination complex e.g., a metalorganic chemical vapor deposition (MOCVD) process for forming barium strontium titanate (BST) thin films on substrates. The effluent is sorptively treated to remove precursor species and MOCVD process by-products from the effluent. An endpoint detector such as a quartz microbalance detector may be employed to detect incipient breakthrough conditions in the sorptive treatment unit.

Abstract: The synthesis of actinide tetraborides including uranium tetraboride (UB4), plutonium tetraboride (PuB4) and thorium tetraboride (ThB4) by a solid-state metathesis reaction are demonstrated. The present method significantly lowers the temperature required to ≦850° C. As an example, when UCl4 is reacted with an excess of MgB2, at 850° C., crystalline UB4 is formed. Powder X-ray diffraction and ICP-AES data support the reduction of UCl3 as the initial step in the reaction. The UB4 product is purified by washing water and drying.

Abstract: The invention relates to a silicon carbide or metal carbide foam to be used as a catalyst or catalyst support for the chemical or petrochemical industry or for silencers, as well as the process for producing the same. The foam is in the form of a three-dimensional network of interconnected cages, whose edge length is between 50 and 500 micrometres, whose density is between 0.03 and 0.1 g/cm3 and whose BET surface is between 20 and 100 m2/g. The carbide foam contains no more than 0.1% by weight residual metal and the size of the carbide crystallites is between 40 and 400 Angstroms. The production process consists of starting with a carbon foam, increasing its specific surface by an activation treatment using carbon dioxide and then contacting the thus activated foam with a volatile compound of the metal, whose carbide it is wished to obtain.

Abstract: The steps for recovering uranium and transuranic elements are simplified, and the generation of waste solvent and waste materials is suppressed. Spent nuclear fuel is dissolved in nitric acid (S100) and the resulting solution is subjected to electrolytic oxidation so that U, Np, Pu, Am is oxidized to VI using Ce as oxidation catalyst. The solution is cooled, and nitrates of valence VI thereby deposit as crystals and are separated from the mother liquor (S104). The mother liquor is heated and concentrated (S114). The mixed crystalline deposit is dissolved in nitric acid (S106), uranyl nitrate is deposited alone by cooling (S108), and the crystals are separated from the U, Np, Pu, Am mixed solution (S110). The uranyl nitrate is dissolved in nitric acid (S112), and the heated and concentrated mother liquor is added to it to prepare another mixed solution.

Abstract: A method of reducing the concentration of neptunium and plutonium from alkaline radwastes containing plutonium and neptunium values along with other transuranic values produced during the course of plutonium production. The OH.sup.- concentration of the alkaline radwaste is adjusted to between about 0.1M and about 4M. [UO.sub.2 (O.sub.2).sub.3 ].sup.4- ion is added to the radwastes in the presence of catalytic amounts of Cu.sup.+2, Co.sup.+2 or Fe.sup.+2 with heating to a temperature in excess of about 60.degree. C. or 85.degree. C., depending on the catalyst, to coprecipitate plutonium and neptunium from the radwaste. Thereafter, the coprecipitate is separated from the alkaline radwaste.

Abstract: The invention is a process for the removal of rare earths from molten chloride electrolyte salts used in the reprocessing of integrated fast reactor fuel (IFR). The process can be used either continuously during normal operation of the electrorefiner or as a batch process. The process consists of first separating the actinide values from the salt before purification by removal of the rare earths. After replacement of the actinides removed in the first step, the now-purified salt electrolyte has the same uranium and plutonium concentration and ratio as when the salt was removed from the electrorefiner.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a Cu--Mg alloy containing not less than about 25% by weight Mg at a temperature in the range of from about 750.degree. C. to about 850.degree. C. to precipitate uranium metal and some of the noble metal fission products leaving the Cu--Mg alloy having transuranium actinide metals and rare earth fission product metals and some of the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels which contain rare earth and noble metal fission products. The oxide fuel is reduced with Ca metal in the presence of CaCl.sub.2 and a U-Fe alloy containing not less than about 84% by weight uranium at a temperature in the range of from about 800.degree. C. to about 850.degree. C. to produce additional uranium metal which dissolves in the U-Fe alloy raising the uranium concentration and having transuranium actinide metals and rare earth fission product metals and the noble metal fission products dissolved therein. The CaCl.sub.2 having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein is separated and electrolytically treated with a carbon electrode to reduce the CaO to Ca metal while converting the carbon electrode to CO and CO.sub.2. The Ca metal and CaCl.sub.2 is recycled to reduce additional oxide fuel.

Abstract: A process of separating transuranium actinide values from uranium values present in spent nuclear oxide fuels containing rare earth and noble metal fission products as well as other fission products is disclosed. The oxide fuel is reduced with Ca metal in the presence of Ca chloride and a U-Fe alloy which is liquid at about 800.degree. C. to dissolve uranium metal and the noble metal fission product metals and transuranium actinide metals and rare earth fission product metals leaving Ca chloride having CaO and fission products of alkali metals and the alkali earth metals and iodine dissolved therein. The Ca chloride and CaO and the fission products contained therein are separated from the U-Fe alloy and the metal values dissolved therein.

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: A process of preparing anhydrous actinide metal trichlorides of plutonium or neptunium by reacting an aqueous solution of an actinide metal trichloride selected from the group consisting of plutonium trichloride or neptunium trichloride with a reducing agent capable of converting the actinide metal from an oxidation state of +4 to +3 in a resultant solution, evaporating essentially all the solvent from the resultant solution to yield an actinide trichloride hydrate material, dehydrating the actinide trichloride hydrate material by heating the material in admixture with excess thionyl chloride, and recovering anhydrous actinide trichloride is provided.

Abstract: A process for separating uranium values and transuranic values from fission products containing rare earth values when the values are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride.

Abstract: A particulate mixture of Fe.sub.2 O.sub.3 and RE.sub.2 O.sub.3, where RE is a rare earth element, is reacted with an excess of HF acid to form an insoluble fluoride compound (salt) comprising REF.sub.3 and FeF.sub.3 present in solid solution in the REF.sub.3 crystal lattice. The REF.sub.3 /FeF.sub.3 compound is dried to render it usable as a reactant in the thermite reduction process as well as other processes which require an REF.sub.3 /FeF.sub.3 mixture. The dried REF.sub.3 /FeF.sub.3 compound comprises about 5 weight % to about 40 weight % of FeF.sub.3 and the balance REF.sub.3 to this end.

Abstract: A process for the extraction of uranium and plutonium from spent nuclear fuels or breeder reactor materials. The spent nuclear fuels or breeder reactor materials are dissolved in nitric acid to provide an aqueous acid solution containing uranium, plutonium, neptunium, other transuranium elements, fission products, corrosion products, activation products and other contamination products. This aqueous acid solution is fed, as an aqueous phase, at a controllable flow rate into a liquid-liquid extraction apparatus also having an organic solvant phase flowing at a controllable rate. The organic phase contains an extraction agent. The temperature of solutions in the extraction apparatus and/or the concentration of the aqueous acid solution before the said aqueous acid solution is fed into the extraction apparatus, is adjusted to satisfy the following inequality: ##EQU1## where T.sub.E =the temperature of the solutions in the extractor (.degree.C.);U.sub.f =the uranium concentration of the feed solution (g/l);Pu.

Abstract: A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

Abstract: Method for recovery of actinides from nuclear waste material containing sintered and other oxides thereof using O.sub.2 F.sub.2 to generate the hexafluorides of the actinides present therein. The fluorinating agent, O.sub.2 F.sub.2, has been observed to perform the above-described tasks at sufficiently low temperatures that there is virtually no damage to the containment vessels. Moreover, the resulting actinide hexafluorides are not destroyed by high temperature reactions with the walls of the reaction vessel. Dioxygen difluoride is readily prepared, stored and transferred to the place of reaction.

Abstract: Method for removal of plutonium impurity from americium oxides and fluorides. AmF.sub.4 is not further oxidized to AmF.sub.6 by the application of O.sub.2 F at room temperature, while plutonium compounds present in the americium sample are fluorinated to volatile PuF.sub.6, which can readily be separated therefrom, leaving the purified americium oxides and/or fluorides as the solid tetrafluoride.

Abstract: A process for the separation of neptunium from an organic phase, which is developed in the recovery of irradiated nuclear fuel and/or fertile material. The organic phase contains uranium-, plutonium- and neptunium ions, tritium in the form of tritiated water and fission products in ionic form, as well as an organic extraction agent dissolved in diluent. After a first wash step, and before the organic phase is further fed to a uranium-plutonium separation or to a uranium-plutonium coreextraction, the organic phase is brought into contact with an aqueous solution containing diluted HNO.sub.3, butyraldehyde and a low concentration of sulfamic acid in countercurrent flow as a second wash step for the selective reduction of Np (VI) to Np (V) and for selective stripping of Np (V), with respect to U and Pu, from the organic into an aqueous phase.

Abstract: In the process for treating irradiated nuclear fuel to effect separation of uranium plutonium other higher actinides, and fission products, in which nitric acid treatment, followed by solvent extraction, then backwashing the reduction of tetra- and hexa-valent plutonium to the tri-valent form, then a second solvent extraction by which the tri-valent plutonium remains in the aqueous phase while uranium goes into the solvent phase, the reduction step is performed by hydrazine with or without tetra-valent uranium nitrate and catalyzed by technetium in the tetra-valent form with or without technetium in one or more higher valency states. The technetium can be present in the system as an irradiation product or be added to the process stream in a combined form.

Abstract: A process for the purification of uranium hexafluoride containing traces of neptunium fluoride and/or plutonium fluoride, wherein the uranium hexafluoride to be purified is contacted with a metal fluoride chosen from the group including lead fluoride PbF.sub.2, uranium fluorides of UF.sub.4+x in which x has a value between 0 and 1 and chromium trifluoride CrF.sub.3, at a temperature such that the plutonium and/or neptunium fluorides are reduced, and wherein the thus purified uranium hexafluoride is recovered.

Abstract: Metallic phosphates are prepared by heating mixtures of BPO.sub.4 and a metallic oxide or salt.

Abstract: Process for the separation of americium from curium contained in an aqueous nitric solution, wherein the initial solution undergoes oxidation at ambient temperature in the presence of phosphoric acid or phosphate ions at a concentration at the most equal to 0.1 mol per liter in such a way as to pass the americium at valency VI, the thus treated solution is brought into contact with a slightly reducing organic solvent having a high affinity for americium at valency VI in such a way as to extract the americium in the organic phase, the organic phase containing the americium at valency VI is then washed, followed by the re-extraction of the americium in the organic phase by means of an aqueous solution.

Abstract: During the reprocessing of irradiated nuclear fuel by solvent extraction techniques a primary separation to give a uranium containing product stream and a plutonium containing product stream occurs. The plutonium in the plutonium containing stream is separated from neptunium and uranium by bringing a solution containing plutonium, neptunium and uranium in an organic solvent into contact first with an aqueous solution of a hydroxylamine and/or a hydrazine salt at 30.degree. to 35.degree. C. to preferentially reduce the neptunium and to extract it into the aqueous phase and then bringing the organic solution containing plutonium and uranium into contact with an aqueous phase containing a hydroxylamine and a hydrazine salt at about 50.degree. C. to preferentially reduce the plutonium and to extract it into the aqueous phase leaving the uranium in the organic solvent.

Abstract: This invention is a process for decomposing ammonium nitrate and/or selected metal nitrates in an aqueous solution at an elevated temperature and pressure. Where the compound to be decomposed is a metal nitrate (e.g., a nuclear-fuel metal nitrate), a hydroxylated organic reducing agent therefor is provided in the solution. In accordance with the invention, an effective proportion of both nitromethane and nitric acid is incorporated in the solution to accelerate decomposition of the ammonium nitrate and/or selected metal nitrate. As a result, decomposition can be effected at significantly lower temperatures and pressures, permitting the use of system components composed of off-the-shelf materials, such as stainless steel, rather than more costly materials of construction. Preferably, the process is conducted on a continuous basis. Fluid can be automatically vented from the reaction zone as required to maintain the operating temperature at a moderate value--e.g., at a value in the range of from about 130.

Abstract: An actinide dioxide, e.g. uranium dioxide, plutonium dioxide, neptunium dioxide, etc., is prepared by reacting the actinide nitrate hexahydrate with sodium dithionite as a first step; the reaction product from this first step is a novel composition of matter comprising the actinide sulfite tetrahydrate. The reaction product resulting from this first step is then converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g. nitrogen) to a temperature of about 500.degree. to about 950.degree. C. for about 15 to about 135 minutes. If the reaction product resulting from the first step is, prior to carrying out the second heating step, exposed to an oxygen-containing atmosphere such as air, the resultant product is a novel composition of matter comprising the actinide oxysulfite tetrahydrate which can also be readily converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g. nitrogen) at a temperature of about 400.degree.

Abstract: This invention relates to a novel method and a novel generator, or source, for providing gaseous negative ions of selected metal hexafluorides. The method is summarized as follows: in an evacuated zone, reacting gaseous fluorine with an actinide-metal body selected from the group consisting of uranium, plutonium, neptunium, and americium to convert at least part of the metal to the hexafluoride state, thus producing gaseous negatively charged metal-hexafluoride ions in the evacuated zone, and applying an electric field to the zone to remove the ions therefrom. The ion source comprises a chamber defining a reaction zone; means for evacuating the zone; an actinide-metal body in the zone, the metal being uranium, plutonium, neptunium, or americium; means for contacting the body with gaseous fluorine to convert at least a part thereof to the hexafluoride state; and means for applying an electric field to the evacuated zone to extract gaseous, negatively charged metal-hexafluoride ions therefrom.

Abstract: By thoroughly mixing a compound of a base metal with a metal of the eighth subgroup of the Periodic Table, heating the resultant to a temperature in excess of 800.degree. C at which it is subjected to a stream of hydrogen and treating thus-obtained intermetallic compound at a still higher temperature and under a high vacuum, a base metal having a purity of at least 98 percent results. When the base metal is an actinide, an alloy or a mixture of alloy and intermetallic compound may be obtained in lieu of the indicated intermetallic compound.

Abstract: An actinide dioxide, e.g., uranium dioxide, plutonium dioxide, neptunium dioxide, etc., is prepared by reacting the actinide nitrate hexahydrate with sodium dithionite as a first step; the reaction product from this first step is a novel composition of matter comprising the actinide sulfite tetrahydrate. The reaction product resulting from this first step is then converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) to a temperature of about 500.degree. to about 950.degree. C. for about 15 to about 135 minutes. If the reaction product resulting from the first step is, prior to carrying out the second heating step, exposed to an oxygen-containing atmosphere such as air, the resultant product is a novel composition of matter comprising the actinide oxysulfite tetrahydrate which can also be readily converted to the actinide dioxide by heating it in the absence of an oxygen-containing atmosphere (e.g., nitrogen) at a temperature of about 400.degree.

Abstract: Salts or materials containing plutonium and americium are dissolved in hydrochloric acid, heated, and contacted with an alkali metal carbonate solution to precipitate plutonium and americium carbonates which are thereafter readily separable from the solution.

Abstract: The preparation of metal and metalloid carbides, borides, nitrides silicides and sulfides by reaction in the vapor phase of the corresponding vaporous metal halide, e.g., metal chloride, with a source of carbon, boron, nitrogen, silicon or sulfur respectively in a reactor is described. Reactants can be introduced into the reactor through a reactant inlet nozzle assembly. Inhibition and often substantial elimination of product growth on exposed surfaces of such assembly is accomplished by introducing the corresponding substantially anhydrous hydrogen halide, e.g., hydrogen chloride, into the principal reactant mixing zone.

Abstract: Pure monocarbides, or pure mononitrides or carbonitrides of metals are prepared by first forming a mixture of carbon with an oxalate of the metals and thermally decomposing the metal oxalate in the presence of the carbon by a stream of hydrogen. The hydrogen is removed and monocarbides are then formed by heating the decomposition products in vacuo to carbothermally reduce them. Mononitrides and carbonitrides can be formed by replacing the hydrogen with nitrogen and heating the decomposition products in the nitrogen.

Abstract: An improved Purex wet recovery process including the step of extracting and separating uranium and plutonium simultaneously from the fission products in the presence of nitric acid and nitrous acid by using a multistage extractor unit having an extracting section and a washing section is provided for separating and recovering neptunium simultaneously with uranium and plutonium contained in spent nuclear fuel. The improved method comprises the steps of maintaining the nitrous acid concentration in said extracting section at a level suited for effecting oxidation of neptunium from (V) to (VI) valence, while lowering the nitrous acid concentration in said washing section so as to suppress reduction of neptunium from (VI) to (V) valence, and maintaining the nitric acid concentration in said washing section at a high level.

Abstract: A process for preparing actinide-nitrides from massive actinide metal which is suitable for sintering into low density fuel shapes by partially hydriding the massive metal and simultaneously dehydriding and nitriding the dehydrided portion. The process is repeated until all of the massive metal is converted to a nitride.