Abstract: Porous particles comprising an active ingredient and a coating exhibiting greater dissolution rate in aqueous media than in alcoholic media are disclosed. A process for the manufacture of the particles is also disclosed, as well as tamper-proof particles and solid dosage forms comprising the coated particles. The differential solubility characteristics of the particle coating allow the particles to be incorporated into abuse-deterrent medicaments.

Abstract: The present invention relates to a process for metallizing nonconductive plastics using etching solutions free of hexavalent chromium. The etching solutions are based on permanganate solutions. After the treatment of the plastics with the etching solutions, the plastics are metallized by means of known processes.

Abstract: An apparatus for uniform reactive thermal treatment of thin-film materials includes a chamber enclosing a tube shaped space filled with a work gas and heaters disposed outside the chamber. The apparatus further includes a loading configuration for subjecting a plurality of planar substrates to the work gas in the tube shaped space. Baffles are disposed above and below the loading configuration.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, but not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.

Abstract: This invention relates to the integration of ammonium carbonate leach processes with established acid and alkaline uranium leach processes as multifunctional industrial processes for the extraction, high degree purification and conversion of processed or semi-processed uranium as U3O8, UO2, or most tetra or hexa-valent forms of uranium, and where applicable, for the recovery of uranium from uranium ores, using advanced multiple stage membrane based technologies for the separation and concentration of uranium in solution from heavy metals and lighter elements that may be present in the solution, and the selective leach and precipitation properties of an ammonium carbonate leach.

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 reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: The present invention includes a composition of LiF—ThF4—UF4—PuF3 for use as a fuel in a nuclear engine.

Abstract: UI3(1,4-dioxane)1.5 and UI4(1,4-dioxane)2, were synthesized in high yield by reacting turnings of elemental uranium with iodine dissolved in 1,4-dioxane under mild conditions. These molecular compounds of uranium are thermally stable and excellent precursor materials for synthesizing other molecular compounds of uranium including alkoxide, amide, organometallic, and halide compounds.

Abstract: This disclosure relates generally to methods and rare earth-containing additives for removing target materials in the form of hydroxides, carbonates, hydrates, or oxyhydroxyls from, a typically aqueous, liquid medium.

Abstract: Systems for treating material are provided that can include a vessel defining a volume, at least one conduit coupled to the vessel and in fluid communication with the vessel, material within the vessel, and NF3 material within the conduit. Methods for fluorinating material are provided that can include exposing the material to NF3 to fluorinate at least a portion of the material. Methods for separating components of material are also provided that can include exposing the material to NF3 to at least partially fluorinate a portion of the material, and separating at least one fluorinated component of the fluorinated portion from the material. The materials exposed to the NF3 material can include but are not limited to one or more of U, Ru, Rh, Mo, Tc, Np, Pu, Sb, Ag, Am, Sn, Zr, Cs, Th, and/or Rb.

Abstract: Aqueous leachant compositions and processes for using the same comprising: (a) providing a metal-containing compound; and (b) subjecting the metal-containing compound to an acid digestion comprising contacting the metal-containing compound with an aqueous leachant; wherein the aqueous leachant comprises a mixture selected from the group consisting of: (i) sulfuric acid and one or more alkanesulfonic acids having alkane moieties selected from the group consisting of propyl, ethyl and methyl groups, at a weight ratio of alkanesulfonic acid to sulfuric acid of 1:1000 to 1:1; (ii) sulfuric acid and one or more salts of alkanesulfonic acids having alkane moieties selected from the group consisting of propyl, ethyl and methyl groups, at a weight ratio of salt of alkanesulfonic acid to sulfuric acid of 1:9 to 1:99.

Abstract: Catalyst for oxidation reactions which comprises at least one constituent active in the catalysis of hydrogen chloride oxidation and support therefor, which support is based on uranium oxide. The catalyst is notable for a high stability and activity.

Abstract: The present invention provides a nuclear fuel assembly, where a boron-containing compound is used as a burnable poison and is distributed in a majority of the rods in the assembly. The assembly comprises a plurality of fuel rods, each fuel rod containing a plurality of nuclear fuel pellets, wherein at least one fuel pellet in more than 50% of the fuel rods in the fuel assembly comprises a sintered admixture of a metal oxide, metal carbide or metal nitride and a boron-containing compound.

Abstract: A process of producing a ceramic powder including providing a plurality of precursor materials in solution, wherein each of the plurality of precursor materials in solution further comprises at least one constituent ionic species of a ceramic powder, combining the plurality of precursor materials in solution with an onium dicarboxylate precipitant solution to cause co-precipitation of the ceramic powder precursor in a combined solution; and separating the ceramic powder precursor from the combined solution. The process may further include calcining the ceramic powder precursor.

Abstract: The present invention provides a two-step process for producing nuclear grade, active uranium dioxide (UO2) powder in which the first step comprises reacting uranium hexafluoride (UF6) with steam in a flame reactor to yield uranyl fluoride (UO2F2); and the second step comprises removing fluoride and reducing UO2F2 to uranium dioxide (UO2) in a kiln under a steam/hydrogen atmosphere. The two-step process, each step separated by a positive sealed valve means to prevent gas, particularly H2 flow back, tightly controls the exothermicity of the reaction, which allows for a very tight temperature control which controls the growth of the particles and results in UO2 powder that is active and of consistent morphology.

Abstract: The invention relates to a process for the chemical beneficiation of raw material containing tantalum-niobium such as wastes, scoria, concentrates and ores.

Abstract: The object of the present invention is to provide a feedstock liquid, from which fuel kernels with good quality can be produced, and a method of preparing the feedstock liquid. The present invention provides a feedstock liquid with a viscosity from 4.0×10?2 to 6.5×10?2 Pa·s at 15° C., for the production of ammonium diuranate particles. The present invention also provides a method of preparing a feedstock liquid used for the production of ammonium diuranate particles, which includes mixing a uranyl nitrate solution and tetrahydrofurfuryl alcohol to produce a uranyl nitrate mixture, dissolving polyvinyl alcohol in water to produce an aqueous polyvinyl alcohol solution, mixing the aqueous polyvinyl alcohol solution with tetrahydrofurfuryl alcohol to produce a polyvinyl alcohol solution, and mixing the uranyl nitrate mixture with the polyvinyl alcohol solution.

Abstract: A substance in a condensed state, for example a powdered solid, is in continuous movement in the longitudinal direction (6) of a furnace (4, 5). A reactive gas mixture is brought into contact with the substance in the condensed state. A plurality of samples of the gaseous mixture are removed at a plurality of reference points (14) spaced apart from one another along the longitudinal direction (6) of the furnace (4, 5); each of the gas samples is analyzed outside the furnace to determine the composition of the gas mixture and for each point (14), the extent of a chemical reaction between the condensed substance and the reactive gas mixture is deduced from the composition of the gas mixture at each of the reference points (14). In particular, the apparatus comprises a sampling and injection rod (10) introduced into the furnace (4, 5) and disposed in its longitudinal direction (6).

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: Fluorine or a fluorine compound is subjected to a reaction with a spent oxide fuel to produce fluorides of uranium and plutonium, and recovering the fluorides using a difference in volatility behavior. The method includes steps of: subjecting a mixture of UO2 and PuO2 with hydrogen fluoride mixed with hydrogen to HF-fluorinate uranium and plutonium into UF4 and PuF3; subjecting UF4 and PuF3 with a fluorine gas to F2-fluorinate uranium and plutonium into UF6 and PuF6; and fractionating UF6 and PuF6 using a difference in phase change of obtained UF6 and PuF6, removing a part of UF6, and volatilizing the remaining UF6 and PuF6 at the same time. By such a reprocessing method, PuF4 hard to undergo a reaction is prevented from being formed as an intermediate fluoride, the material of a reactor is hard to be corroded, and a consumption of expensive fluorine gas is reduced.

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 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: It is an object to increase a reprocessing speed of spent nuclear fuel and to obtain uranium having a high purity and a plutonium mixture reusable as it is at a low cost through a simple procedure. The spent nuclear fuel 1 is subjected to fluorination using fluorine 2 in a fluorination step 3, and as a result, uranium, a mixture of uranium and plutonium and a fission product are separated and recovered independently of one another. The plutonium fluoride volatilized in the fluorination is recovered along with a fixing agent and then passed through an oxidative conversion step 8, thereby recovering a mixture of uranium and plutonium oxides 9. Since the uranium can be recovered in a high purity, it is managed very easily when reused or saved. Further, since the uranium and plutonium are recovered as a mixture thereof, fuel reproduction cost is decreased and prevention of proliferation is strengthened.

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 subject apparatus provides a means to produce UCl3 in large quantities without incurring corrosion of the containment vessel or associated apparatus. Gaseous Cl is injected into a lower layer of Cd where CdCl2 is formed. Due to is lower density, the CdCl2 rises through the Cd layer into a layer of molten LiCl—KCL salt where a rotatable basket containing uranium ingots is suspended. The CdCl2 reacts with the uranium to form UCl3 and Cd. Due to density differences, the Cd sinks down to the liquid Cd layer and is reused. The UCl3 combines with the molten salt. During production the temperature is maintained at about 600° C. while after the uranium has been depleted the salt temperature is lowered, the molten salt is pressure siphoned from the vessel, and the salt product LiCl—KCl-30 mol % UCl3 is solidified.

Abstract: Metal oxides having a perovskite or perovskite-like crystal structure are prepared by a process comprising subjecting a mixture of starring powders to a high energy milling sufficient to induce chemical reaction of the components and thereby directly mechanosynthesize said metal oxide in the form of a perovskite or a perovskite-like material having a nanocrystalline structure as determined by X-ray diffractometry. The process according to the present invention is simple, efficient, not expensive and does not require any heating step for producing a perovskite that may easily show a very high specific surface area. Another advantage is that the perovskite obtained according to the present invention also has a high density of lattice defects thereby showing a higher catalytic activity, a characteristic which is highly desirable in their eventual application as catalysts and electronic conductors.

Abstract: Described are preferred polymerized organic-inorganic processes for producing mixed metal oxide powders suitable for use in the ceramics and related industries. The preferred processes employ a non-chelating polymer such as polyvinyl alcohol or polyalkylene glycol as a carrier and can provide single-phase, mixed oxide powders in high yields at relatively low temperatures.

Abstract: A method of removing from a metal salt ionic species contained therein involves contacting the metal salt with an ionic liquid to dissolve the metal salt, the ionic species or both. At least in the case where both the metal salt and the ionic species are dissolved, the resultant ionic liquid composition is treated to separate the ionic species therefrom and subsequently processed to recover the metal salt.

Abstract: Uniformly sized and shaped particles of metal salts are provided comprised of one or more metal cations in combination with one or more simple oxoacid anions and a general method for the controlled precipitation of said metal salts from aqueous solutions. The methods proceed via the in situ homogeneous production of simple or complex oxoacid anions in which one or more of the nonmetallic elements e.g. Group 5B and 6B (chalcogenides), and 7B (halides) comprising the first oxoacid anion undergo oxidation to generate the precipitant anionic species along with concurrent reduction of the nonmetallic element of a second, dissimilar oxoacid anion. The oxoacid anions are initially present in solution with one or more metal cations known to form insoluble salts with the precipitant anion.

Abstract: A process for converting UF6 to a solid uranium compound such as UO2 and CaF. The UF6 vapor form is contacted with an aqueous solution of NH4OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH4OH and NH4F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH4OH and NH4F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH)2 to precipitate CaF2 leaving dilute NH4OH.

Abstract: A process for the destructive oxidation of organic compounds containing one or more optionally substituted hydrocarbon groups, comprising the step of carrying out the oxidation in the presence of a catalyst comprising uranium oxide in which the average oxidation state of the uranium is greater than four.

Abstract: A method has been developed for the solution-based metal exchange of carboxylato-alumoxanes [Al(O)x(OH)y(O2CR)z]n with a wide range of metal cations. Metal-exchanged carboxylato-alumoxanes are new, particularly those in which about 10% to about 50% or more of the Al ions are exchanged for other metal ions. Additionally, the carboxylic acid ligands can be stripped from the boehmite core of metal-exchanged carboxylato-alumoxanes at low temperature leading to the formation of metal-exchanged boebmite particles. These new material phases can be used as intermediates for preparation of mixed metal aluminum oxide materials. Thermolysis of the metal-exchanged carboxylato-alumoxanes or metal-exchanged boehmite particles results in doped aluminas (M/Al2O3), binary (MAlOx), ternary (MM′AlOx) and even more complex metal aluminum oxide compounds, where M and M′ are metal ions other than those of aluminum and are preferably those of Lanthanide metals or transition metals.

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 multicomponent fluid feed apparatus is disclosed that independently preheats and then mixes two or more fluid streams being introduced into a high temperature chemical reactor to promote more rigorous and complete reactions using assemblies of inert tubular elements and an integral mixing orifice plate. The design allows use of ceramic and speciality alloy materials for high temperature service with particularly corrosive halide feeds such as UF.sub.6 and HF. Radiant heat transfer to the tubular elements from external means gives the necessary system high temperatures without excessive temperatures to cause material failure. Preheating of the gaseous reactants in a separate step prior to mixing and injecting the gaseous reactants into a high temperature chemical reactor was found to provide an improved thermal conversion of UF.sub.6 to uranium oxides.

Abstract: The present invention is a method of removing an impurity of plutonium, lead or a combination thereof from a mixture of radionuclides that contains the impurity and at least one parent radionuclide. The method has the steps of (a) insuring that the mixture is a hydrochloric acid mixture; (b) oxidizing the acidic mixture and specifically oxidizing the impurity to its highest oxidation state; and (c) passing the oxidized mixture through a chloride form anion exchange column whereupon the oxidized impurity absorbs to the chloride form anion exchange column and the 22.sup.9 Th or 2.sup.27 Ac "cow" radionuclide passes through the chloride form anion exchange column. The plutonium is removed for the purpose of obtaining other alpha emitting radionuclides in a highly purified form suitable for medical therapy.

Abstract: The process for converting feed materials of high mineral content containing primary metal values and fluorine values to the primary metal or useful compounds thereof and to fluorine values or useful compounds thereof, wherein the feed materials constitutes a difficultly soluble matrix, the process having the steps of contacting the feed materials in a reactor with a humidified, gaseous system at from about 200.degree. C. to about 1600.degree. C., the contacting being carried out such as to convert the primary metal values to oxide residues at commercially acceptable rates and to evolve gaseous fluoride from the feed, digesting said oxide residues in an acidic digest medium and separating the primary metal values from the resulting digest liquor and from other components of the residues.

Abstract: The present invention provides a process for the oxidation of uranium hexafluoride by injecting the uranium hexafluoride and an oxidant gas together into a reaction vessel to form a plume characterized in that a plurality of the said plumes are formed together in the same vessel, the plumes mutually contributing to a circulating product formation stream in the said vessel. The process may include the establishment of three or more plumes simultaneously contributing to the reaction between the gases in the reaction vessel. The oxidant gas may comprise steam. The process may be one in which the product is formed as a particulate solid. The product particles may initially be formed as dentritic particles which may be recirculated in the reaction vessel to promote seeding, growth, agglomeration and aggregation of the required product particles, the plumes thereby contributing to the product formation process in the vessel.

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 of purifying a UF.sub.6 gas stream which comprises irradiating the UF.sub.6 gas stream with laser radiation in a vessel in order to selectively convert fluoride impurities in the gas stream to involatile products, removing the purified UF.sub.6 gas stream from the vessel and separately removing the impurities from the vessel.

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

Abstract: A process for reducing the concentration of fissionable uranium containing components of a feed material to below a target concentration level, the process having the steps of providing the feed material as an aqueous system of high-enriched uranyl nitrate hydrate (HE-UNH) wherein at least about 18% of the uranium therein is the .sup.235 U isotope, providing an aqueous dilution medium containing natural uranyl nitrate hydrate (NUNH) wherein at least about 90% of the uranium therein is the .sup.238 U isotope, forcing a diluent stream of the dilution medium through a jet pump to provide a suction region within the pump coextensive with the diluent stream, providing a feed fluid connection of the HE-UNH within the first vessel means with the suction region to cause a HE-UNH stream to flow into the suction region and intermix therein with the diluent stream to form a product stream of markedly reduced .sup.235 U concentration.

Abstract: The process for converting feed materials of high mineral content and substantial radioactivity levels to concentrated radionuclide products of high radioactivity levels and to other products of discard or very low radioactivity levels, wherein the feed materials consists of a difficultly soluble matrix contain substantial metal, fluorine, and radionuclide values assaying above about 30 pCi/g, the process having the steps of contacting the feed materials with high temperature steam of from about 200.degree. C. to about 1500.degree. C.

Abstract: A controlled multi-stage process for the stripping of uranium includes the introduction at one end of a solvent extraction device a uranium loaded organic solution. A concentrated stripping acidic aqueous solution is introduced at the other end of the solvent extraction device so that the aqueous solution and the organic solution are contacted counter-currently in the solvent extraction device at a temperature not substantially exceeding 35.degree. C.

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: 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: An off-gas quencher and solid recovery scrubber unit includes a wet flue gas scrubber which has the dual responsibilities of lowering the temperature of the inlet hot gas entering through the scrubber and trapping contaminants from the gas stream into the liquid stream. The hot exhaust gases are first cooled by evaporating the liquid scrubber solution. The contaminants of the exhaust gas are neutralized by a suitable reagent such as sodium hydroxide and the product is collected in the scrubbing solution. Since the solution is continuously recycled, the concentration of the scrubbing agent will be diminished as the scrubbing proceeds, while the concentration of the scrubbing product in the solution will rise to the solubility limit of the product. The scrubbing products start to precipitate and are collected at the bottom of the scrubber and are withdrawn. The scrubbing reagents are continuously replenished to the scrubber.

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 for the treatment of a material which is or is suspected to contain or carry one or more actinides or compounds thereof to dissolve such actinides or compounds comprises contacting the material with an aqueous solution having a pH in the range 5.5 to 10.5 which is free of heavy metal ions and comprises ingredients which are naturally degradable to non-toxic products with or without mild physical assistance such as heat or ultra-violet radiation, said solution comprising:(a) carbonated water;(b) a conditioning agent;and (c) a complexing agent which comprises the anion of a carboxylic acid having from 2 to 6 carbon atoms.The process may be employed to separate spent nuclear fuel from its metal containment or it may be employed to decontaminate surface, e.g. concrete or soil or pipes carrying traces of actinides, or bulk materials such as soil or rubble.