Abstract: Provided is a method of producing an Al—Sc based alloy suitable for production of an Al—Sc based alloy that: eliminates the needs for equipment for heating in an inert gas atmosphere or a vacuum atmosphere, a reducing agent such as metal Ca, and equipment and power for molten salt electrolysis; can be performed adequately by heating up to 1,050° C.; and enables continuous operation. The method of producing an Al—Sc based alloy includes: loading into a reaction vessel metal aluminum (Al), a metal fluoride salt, and a scandium compound; elevating a temperature of a reaction system to from 700 to 1,050° C. to form a molten metal layer including molten metal aluminum serving as a lower layer and a molten salt layer in which the metal fluoride salt and the scandium compound are melted serving as an upper layer; and transferring a scandium ion (Sc3+) generated in the molten salt layer side to the molten metal layer side.

Abstract: The invention relates to a process for manufacturing an oxychloride or oxide of actinide(s) and/or of lanthanide(s) from a chloride of actinide(s) and/or of lanthanide(s) present in a medium comprising at least one molten salt of chloride type comprising a step of bringing said chloride of actinide(s) and/or lanthanide(s) present in said medium comprising at least one molten salt of chloride type into contact with a wet inert gas.

Abstract: The invention relates to a process and a device for bringing two immiscible liquids into contact, without mixing and at high temperature, with heating and kneading by induction. In particular, the invention relates to a process and a device for bringing into contact metals and salts which are molten at high temperatures, for example as high as approximately 1,100 K.

Abstract: Used uranium oxide fuel is detoxified by extracting transuranic and reactive fission products into molten salt. By contacting declad and crushed used uranium oxide fuel with a molten halide salt containing a minor fraction of the respective uranium trihalide, transuranic and reactive fission products partition from the fuel to the molten salt phase, while uranium oxide and non-reactive, or noble metal, fission products remain in an insoluble solid phase. The salt is then separated from the fuel via draining and distillation. By this method, the bulk of the decay heat, fission poisoning capacity, and radiotoxicity are removed from the used fuel. The remaining radioactivity from the noble metal fission products in the detoxified fuel is primarily limited to soft beta emitters. The extracted transuranic and reactive fission products are amenable to existing technologies for group uranium/transuranic product recovery and fission product immobilization in engineered waste forms.

Abstract: The present invention provides a burner for use in a combustion-type waste gas treatment system for combusting waste gases emitted from semiconductor manufacturing system, particularly, a deposition gas containing SiH4 and a halogen-base gas, simultaneously at a high efficiency of destruction, making it difficult for a powder of SiO2 to be attached and deposited, performing a low-NOx combustion, and maintaining a desired level of safety. The combustion-type waste gas treatment system has a flame stabilizing zone (15), which is open toward a combustion chamber (11), surrounded by a peripheral wall (12), and closed by a plate (14) remotely from the combustion chamber. A waste gas, an auxiliary combustible agent, and air are introduced into and mixed with each other in the flame stabilizing zone (15), and the mixed gases are ejected toward the combustion chamber (11) perpendicularly to the plate (14).

Abstract: This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF4 with a eutectic melting point of 500° C. Prior to lowering the basket, the salt is heated to a temperature of between 550° C. and 700° C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF6. In addition, after dissolution, the basket contains PuO2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Abstract: Composites of at least two different plastics materials joined directly to one another, wherein

Abstract: In a method for reducing hydrogen chloride emissions from an asphalt air-blowing process, an asphalt is subjected to an air-blowing process where air is bubbled through hot asphalt to raise the softening point of the asphalt. The fumes from the air-blowing process are bubbled through a liquid seal in a knockout tank before going to an incinerator and finally being emitted to the atmosphere. The knockout tank normally operates to condense oil in the fume stream, and the liquid seal is composed of this oil, as well as some of the water evolved in the air-blowing process. When using ferric chloride or ferrous chloride as a catalyst in the air-blowing process, the fume stream contains significant levels of hydrogen chloride.

Abstract: The invention relates to a method for processing waste that contains contaminants. The contaminants may be metals modified by carbon, oxygen, phosphorus, or sulfur. The waste in the form of a powder is mixed with an ionic reducing agent in an inert liquid medium. The mixture is melted to give a first liquid phase and a second metal phase. The two phases are separated and solidified to enable disposal or temporary storage of the first liquid phase and to enable recycling of the second metal phase. The method is useful for inerting or reclaiming waste containing metal contaminants.

Abstract: A simple, cost effective method for reclaiming tungsten values from tungsten-thoria has been invented. The method involves reacting the tungsten-thoria with molten sodium hydroxide to form a soluble sodium tungstate and an insoluble residue of thoria, cooling the melt, dissolving the melt in water, and filtering to separate the insoluble residue of thoria from the dissolved sodium tungstate. The method resulting in substantially all of the tungsten values being reclaimed as sodium tungstate and acutely reducing the amount of radioactive material for disposal.

Abstract: A process for preparing uranium metal or alloy thereof suitable for use in a metal-based uranium enrichment plant or other use requiring a superdense metal comprising providing a molten metal bath containing the alloy metal and feeding uranium oxide and a reactive metal reductant into the molten metal bath so that the oxide is reduced to elemental uranium and alloying the thus formed uranium with the bath metal.

Abstract: A method for reprocessing metal parts that are radioactively contaminated with uranium includes smelting the metal parts so that a melt and a slag are formed. U.sub.235 -depleted uranium is admixed with the metal parts and/or the melt and/or the slag. It is contemplated for the U.sub.235 -depleted uranium to be admixed in the form of uranium glass.

Abstract: A method of melting treatment of radioactive miscellaneous solid wastes containing therein an electrically conductive substance and other waste components. This method comprises charging the radioactive miscellaneous solid wastes into a cold crucible induction melting furnace provided with a high-frequency coil; supplying a high-frequency current to the high-frequency coil of the melting furnace to thereby heat and melt the electrically conductive substance, e.g. a metal, in the miscellaneous solid wastes; indirectly heating the other components in the miscellaneous solid wasted by utilizing the electrically conductive substance as a starting source of heating and melting; and placing the whole of the radioactive miscellaneous solid wastes into a molten state.

Abstract: An apparatus and method for continuous production of liquid uranium alloys through the electrolytic reduction of uranium chlorides. The apparatus includes an electrochemical cell formed from an anode shaped to form an electrolyte reservoir, a cathode comprising a metal, such as iron, capable of forming a eutectic uranium alloy having a melting point less than the melting point of pure uranium, and molten electrolyte in the reservoir comprising a chlorine or fluorine containing salt and uranium chloride.

Abstract: An environmentally sound process is described for the remediation of waste materials that allows the separation, recovery and decontamination of metals. The method includes chemically reducing essentially all of a reducible toxic and potentially hazardous metal-containing component of a waste composition. The waste is directed into a molten metal bath, including a first reducing agent which, under the operating conditions of the molten metal bath, chemically reduces a metal of the metal-containing component to form a dissolved intermediate. A second reducing agent is directed into the molten metal bath. The second reducing agent, under the operations of the molten metal bath, chemically reduces the metal of the dissolved intermediate.

Abstract: Uranium chloride is reacted with either magnesium, sodium or calcium in the presence of a molten salt comprising light metal chlorides including lithium chloride. The temperature is maintained below the melting point of uranium. The magnesium may be in the form of magnesium-cadmium alloy, the temperature being maintained below the temperature at which magnesium and cadmium vaporize. The components of the molten salt may be first fused together so as to form the molten salt eutectic. Subsequently after separation of the uranium, products of the reaction may be recovered and recycled.

Abstract: An environmentally sound process is described for the remediation of waste materials that allows the separation, recovery and decontamination of metals. The method includes chemically reducing essentially all of a reducible toxic and potentially hazardous metal-containing component of a waste composition. The waste is directed into a molten metal bath, including a first reducing agent which, under the operating conditions of the molten metal bath, chemically reduces a metal of the metal-containing component to form a dissolved intermediate. A reagent is directed into the molten metal bath for metal-ligand exchange with the dissolved intermediate to form a metal-ligand exchange product that includes the metal of the dissolved intermediate. A second reducing agent is directed into the molten metal bath. The second reducing agent, under the operations of the molten metal bath, chemically reduces the metal of the metal-ligand exchange product.

Abstract: In the pyrochemical reduction of uranium dioxide or other actinide metal oxides by reaction with magnesium, magnesium oxide byproduct is produced. The use of a salt flux comprising magnesium chloride and a rare earth element trichloride such as neodymium chloride is disclosed. The neodymium chloride reacts with magnesium oxide to form magnesium chloride and neodymium oxychloride. The resulting magnesium chloride-neodymium oxychloride salt mixture can readily be subjected to electrolysis to regenerate magnesium and neodymium chloride for reuse in the pyrochemical reduction process. Other uses of the magnesium chloride-neodymium chloride salt flux are also proposed.

Abstract: A method is disclosed for reducing the volume of a radioactive composition by separating a radioactive first component from a second component of the radioactive composition. The method includes directing the radioactive composition into a reaction zone. The reaction zone includes a molten bath, wherein oxidation of a component of the radioactive composition in the molten bath will cause separation of the radioactive first component from the second component. An oxidizing agent is directed into the molten bath, which oxidizes a component of the radioactive composition, whereby the radioactive first component is separated from the second component. The net volume of the radioactive composition is thereby reduced.

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 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 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: A method is provided for cleaning depleted uranium and its associated radioactivity from target sands, for reusing many of the reactants used in the cleaning, and for recovering the depleted uranium content as uranyl nitrate. The method involves roasting and tumbling target sands with molten nitrate mixtures, followed by aqueous extraction to remove the nitrates, then nitric acid extraction to remove uranium oxides as uranyl nitrates which can be solvent extracted into organic phases.

Abstract: A method is described for decontaminating magnesium fluoride resulting from the reduction of uranium fluoride to the metal by reaction with magnesium. This decontamination employs reactions with magnesium and carbon to remove radioactive components from the said magnesium fluoride in its molten state.

Abstract: A process for recovering non-ferrous metal values from their ores, minerals, concentrates, oxidic roasting products, or slags by sulphating said starting material using a mixture comprising iron (III) sulphate and alkali metal- or ammonium sulphate as a reagent.

Abstract: Niobium is recovered from scrap uranium-niobium alloy by melting the scrap with tin, solidifying the billet thus formed, heating the billet to combine niobium with tin therein, placing the billet in hydrochloric acid to dissolve the uranium and leave an insoluble residue of niobium stannide, then separating the niobium stannide from the acid.

Abstract: Spent nuclear fuels, including actinide fuels, volatile and non-volatile fission products, are reprocessed and separated in a molten metal solvent housed in a separation vessel made of a carbon-containing material. A first catalyst, which promotes the solubility and permeability of carbon in the metal solvent, is included. By increasing the solubility and permeability of the carbon in the solvent, the rate at which actinide oxides are reduced (carbothermic reduction) is greatly increased. A second catalyst, included to increase the affinity for nitrogen in the metal solvent, is added to increase the rate at which actinide nitrides form after carbothermic reduction is complete.

Abstract: A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Abstract: In a method for reprocessing irradiated nuclear fuels, the fuel to be reprocessed is dissolved in a fused-salt bath while absolute sulfuric acid is added dropwise to said bath, plutonium sulfate is thermally converted into the corresponding oxide, the uranium oxide is electrolytically deposited from the fused-salt bath as the electrolyte, the melted salts are recycled to the starting end of the process and the fission products and the melted salts are conditioned for final disposal.

Abstract: An actinide nitride-fueled nuclear reactor and a method of operation therefor, including continuous in situ removal of fission products and optional addition of fuel-forming actinide material as the reaction proceeds. The reactor employed has a fuel system comprising a critical mass of a nitride of an actinide metal in contact with a non-critical solution of the actinide metal in a molten metal solvent of low neutron adsorption cross section such as tin, said fuel system being maintained under a nitrogen atmosphere in an inert, refractory vessel such as graphite which is non-conducive to the formation of actinide oxides. Fission products formed are continuously exchanged with the actinide metal dissolved in the molten metal solvent as the nuclear reaction proceeds, with equivalent amounts of actinide nitride being formed and precipitated into the critical mass as fission products are dissolved in the molten metal solution.

Abstract: An organic waste containing at least one element selected from the group consisting of strontium, cesium, iodine and ruthenium is treated to achieve a substantial reduction in the volume of the waste and provide for fixation of the selected element in an inert salt. The method of treatment comprises introducing the organic waste and a source of oxygen into a molten salt bath maintained at an elevated temperature to produce solid and gaseous reaction products. The gaseous reaction products comprise carbon dioxide and water vapor, and the solid reaction products comprise the inorganic ash constituents of the organic waste and the selected element which is retained in the molten salt. The molten salt bath comprises one or more alkali metal carbonates, and may optionally include from 1 to about 25 wt.% of an alkali metal sulfate.

Abstract: In the recovery of plutonium from irradiated nuclear fuel elements especially those coming from fast nuclear reactors, the improvement consisting in that the spent nuclear fuel elements are subjected to the action of a molten nitrate bath (mixed nitrates of alkali metals or alkaline earth metals), whereafter the plutonium thus obtained is further decomposed in a nitrate bath at a higher temperature, then the plutonium is recovered and a further thermal decomposition of the remaining material at a still higher temperature enables the uranium to be recovered. The recovery of plutonium requires the action of the nitric vapors, the recovery of uranium does not. Molten nitrates can likewise be recovered and recycled.

Abstract: The reduction of UF.sub.5 to UF.sub.4 in a molten fluoride salt by sparging with hydrogen is catalyzed by metallic platinum. The reaction is also catalyzed by platinum alloyed with gold reaction equipment.

Abstract: There is described a method for sodium-deactivating and/or stocking irradiated nuclear fuel elements, said elements being contacted with a fluid, in which use is made as fluid, of a molten salt mixture which is uncorrosive for the element stainless steels, inert relative to uranium and plutonium oxides by the deactivating temperature and reactive with sodium metal without producing a detonating gas.

Abstract: Uranium is separated from mixed oxides of thorium and uranium by a pyrometallurgical process in which the oxides are mixed with a molten chloride salt containing thorium tetrachloride and thorium metal which reduces the uranium oxide to uranium metal which can then be recovered from the molten salt. The process is particularly useful for the recovery of uranium from generally insoluble high-density sol-gel thoria-urania nuclear reactor fuel pellets.

Abstract: Ceramic nuclear fuel is reprocessed through a method wherein the fuel is dispersed in a molten eutectic mixture of at least two alkali metal nitrates and heated to a temperature in the range between 200.degree.C and 300.degree.C. That heated mixture is then subjected to the action of a gaseous stream containing nitric acid vapors, preferably in the presence of a catalyst such as sodium fluoride. Dissolved fuel can then be precipitated out of solution in crystalline form by cooling the solution to a temperature only slightly above the melting point of the bath.

Abstract: An electrolytic process for making uranium from uranium oxide using Cl.sub.2 anode product from an electrolytic cell to react with UO.sub.2 to form uranium chlorides. The chlorides are used in low concentrations in a melt comprising fluorides and chlorides of potassium, sodium and barium in the electrolytic cell. The electrolysis produces Cl.sub.2 at the anode that reacts with UO.sub.2 in the feed reactor to form soluble UCl.sub.4, available for a continuous process in the electrolytic cell, rather than having insoluble UO.sub.2 fouling the cell.