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: An apparatus for the removal of uranium from a body of material is provided. The apparatus has at least one ultrasonic extractor, having a bottom and a top. The at least one ultrasonic extractor is configured to accept solids at the bottom and acid at the top, and has a mixing screw and at least one source of ultrasonic energy. The mixing screw is configured to transport the solids in a direction countercurrent to the acid in the at least one ultrasonic extractor; and the source of ultrasonic energy is configured to impart ultrasonic energy into the solids and the acid, as the solids and the acid traverse the at least one ultrasonic extractor countercurrently.

Abstract: An apparatus and method to remove uranium from a body of material wherein the method includes the steps of depositing the body of solid material in an ultrasonic extractor and depositing an amount of acid in the ultrasonic extractor. The method also provides for the steps of heating a jacket of the ultrasonic extractor, transporting the body of solid material in the ultrasonic extractor and the amount of acid such that the body of solid material and the acid contact each other inside the heated ultrasonic extractor while the ultrasonic extractor provides ultrasonic energy to both the body of solid material and the amount of acid, wherein the amount of acid strips uranium from the body of solid material. The method further provides for collecting the amount of acid and the body of solid material in the ultrasonic extractor in different positions, transporting the amount of acid with the stripped uranium to an extraction mixer settler, and settling uranium product from the extraction mixer settler.

Abstract: The invention provides a method of separating uranium from at least fission products in irradiated nuclear fuel, said method comprising reacting said irradiated nuclear fuel with a solution of ammonium fluoride in hydrogen fluoride fluorinating said reacted irradiated nuclear fuel to form a volatile uranium fluoride compound and separating said volatile uranium fluoride compound from involatile fission products. The invention thus provides a reprocessing scheme for irradiated nuclear fuel. The method is also capable of reacting, and breaking down Zircaloy cladding and stainless steel assembly components. Thus, whole fuel elements may be dissolved as one thereby simplifying procedures over conventional Purex processes.

Abstract: A method of dissolving in an ionic liquid a metal in an initial oxidation state below its maximum oxidation state, characterized in that the ionic liquid reacts with the metal and oxidizes it to a higher oxidation state. The initial metal may be in the form of a compound thereof and may be irradiated nuclear fuel comprising UO2 and/or PuO2 as well as fission products. The ionic liquid typically is nitrate-based, for example a pyridinium or substituted imidazolium nitrate, and contains a Bronstead or Franklin acid to increase the oxidizing power of the nitrate. Suitable acids are HNO3, H2SO4 and [NO+]. Imidazolium nitrates and certain pyridinium nitrates form one aspect of the invention.

Abstract: Radioactively contaminated material is cleaned by contacting the material with a decontaminating liquid comprising an aqueous solution of nitric acid containing an NOx generating agent. The NOx generating agent may be a nitrite, for example, sodium nitrite, or a ferrous metal. The material to be cleaned may comprise a plastics material contaminated with uranium or other actinides. Cleaning is effected by placing the material in a rotatable, apertured vessel in which the material is subjected to a leaching cycle by contact with the decontaminating liquid and then a washing cycle in which the material is contacted with a washing liquid.

Abstract: The invention relates to a process for the processing of nuclear targets and/or fuels based on metallic aluminium by dissolving with the aid of aqueous tetramethylammonium hydroxide (TMAOH) solutions.Dissolving can consist of a total dissolving in a single stage of the core and the can of the nuclear fuel by TMAOH or a decanning of the can by TMAOH, followed by a nitric dissolving of the core. The use of TMAOH eliminates the disadvantages associated with dissolving in a concentrated nitric or sodium medium and decanning with soda and in particular permits the easy vitrification of the effluents produced by the process.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group colon hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: The "Purex" solvent extraction and separation process for recovering uranium from scrap material or spent fuel containing gadolinia is modified by treating uranium containing scrap or spent fuel with nitric acid to convert uranium oxide to soluble uranyl nitrate in an aqueous medium, separating the uranyl nitrate from the aqueous media with an organic medium containing tri-butyl phosphate, and then releasing the separated uranium from the organic medium with water. The improvement comprises monitoring the density of the organic medium and controlling the feed rate of the uranium in the aqueous medium in response to the density measurements.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably at least one substance from the group: hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on pressurized water reactors, boiling water reactors, metallic substances, high temperature alloys and brickworks as well. The used decontamination solution can, after regeneration, by recycled into the decontamination process. Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries. The decontamination agent (NBF.sub.

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: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group, hexafluorosilicate acid, fluoroboric acid, and the salts of both these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid. The HBF.sub.

Abstract: A method of dissolving nuclear material from a relatively long length cut from a nuclear fuel cartridge and having a liquid permeable passageway extending from one end to the other end of the cut length. The method comprises holding the cut length upright in a dissolvent for the nuclear material in the cut length, so that gas generated by the dissolvent in reacting with the nuclear material produces an effect analogous to that of an air lift pump and circulates the dissolvent through the passageway.

Abstract: A method for preparing highly hydrogen-reactive surfaces on metals which normally require substantial heating, high pressures, or an extended induction period, which involves pretreatment of said surfaces with either a non-oxidizing acid or hydrogen gas to form a hydrogen-bearing coating on said surfaces, and subsequently heating said coated metal in the absence of moisture and oxygen for a period sufficient to decompose said coating and cooling said metal to room temperature. Surfaces so treated will react almost instantaneously with hydrogen gas at room temperature and low pressure. The method is particularly applicable to uranium, thorium, and lanthanide metals.

Abstract: A treating process for the separation of coated nuclear fuel particles from a graphitic matrix in which the nuclear fuel particles are embedded, which is employed in an installation for the conditioning of graphitic fuel elements of high-temperature nuclear reactors. The graphite which encompasses the nuclear fuel particles is conducted away through the action of a brush which isolates the nuclear fuel particles together with their coatings or the nuclear particles themselves, and wherein the nuclear fuel particles which are contained in the brushed product are then separated from the comminuted graphite. The nuclear fuel particles are worked out of the graphitic matrix through the action of the bristles of the brush which conduct the graphite away, but in which the coated nuclear fuel particles when they are contained within hard coatings or the nuclear fuel particles themselves, remain preserved.

Abstract: Apparatus for agitating and circulating a liquid 20 comprising a vessel having two upright chambers 10, 12 and two vertically displaced ducts 14, 18 connecting the chambers 10, 12. In use of the apparatus, a steady flow of gas into the top of one chamber 10 provides both an oscillation and a circulation of the liquid 20 in the apparatus, as the gas is periodically vented through the upper duct 18.The apparatus has one application in the dissolution of material from nuclear fuel pins during reprocessing.

Abstract: Reprocessing a nuclear reactor fuel rod which contains nuclear fuel in a cladding tube, by removing the cladding tube from the nuclear fuel. The cladding tube is uniformly heated in hermetically sealed condition together with the nuclear fuel contained thereon to permanently expand the diameter of the cladding tube without the formation of cracks in the cladding tube, increasing the distance between the nuclear fuel and the cladding tube. Subsequently the expanded cladding tube is opened at one end. The nuclear fuel is removed from the opened cladding tube and is processed further, separate from the cladding tube.

Abstract: Removal of the graphite structure from the nuclear fuel material of fuel elements of gas-cooled high-temperature reactors is performed by heat-treatment of the fuel elements in oxygen-containing gas at temperatures below 700.degree. C. until the carbon or graphite structure is loosened up and converted into a mechanically removable material. During heat-treatment, mechanical forces are preferably applied for continually removing the outer layers that are most heavily attacked and transport of the dust into cooler temperature zones. For this purpose, the fuel elements are agitated during heat-treatment, as by an oscillating sieve or by brushes. Pre-impregnation with a material catalyzing the combustion is useful if it is important to have the heat-treatment temperature as low as possible.

Abstract: Disclosed is a method of comminuting irradiated ferritic steel by placing the steel in a solution of a compound selected from the group consisting of sulfamic acid, bisulfate, and mixtures thereof. The ferritic steel is used as cladding on nuclear fuel rods or other irradiated components.

Abstract: A neclear fuel rod assembly used in a nuclear reactor and deposited with radioactive scale is taken out of a fuel pool and disposed in a substantially sealed vertical casing, while the scale is still wetted by water. Gas in the casing is heated to a temperature higher than room temperature by the heat of nuclear fission or an electric heater to dry and peel off the scale. A mixture of the scale and water collected in a bottom portion of the casing is sent to a waterscale separator. To rapidly and perfectly remove the scale, after completing the step described above, water is sprayed onto the fuel rod assembly to wet remaining scale. Then, the drying of the scale with hot gas is repeated again.

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: A method of decladding an assembly comprising an element selected from the group consisting of uranium, thorium and mixtures thereof, clad in stainless steel, zirconium, or an alloy consisting essentially of zirconium and containing minor amounts of nickel, chromium, tin, iron or combinations thereof. In a first step the cladding is scored or perforated to expose the selected element. Thereafter, the assembly is exposed to a hydrogen atmosphere at an elevated temperature for a time sufficient for the hydrogen and selected element to react and form a hydride. The temperature then is further increased to decompose the hydride back to gaseous hydrogen and the selected element. The hydriding-dehydriding preferably are repeated at least two additional times to ensure complete release of any volatile gases present.

Abstract: A method of leaching a material from a cavity having a closed or a partially closed top. The method consists of subjecting the liquid to a number of pressure cycles, each pressure cycle involving a decrease in pressure to cause boiling of the liquid, followed by a rise in pressure to inhibit the boiling. The method may include the step of heating the liquid to a temperature near to its boiling point.

Abstract: A method and apparatus for processing spent nuclear reactor fuel wherein plutonium is continuously contaminated with radioactive fission products and diluted with uranium. Plutonium of sufficient purity to fabricate nuclear weapons cannot be produced by the process or in the disclosed reprocessing plant. Diversion of plutonium is prevented by radiation hazards and ease of detection.

Abstract: Irradiated nuclear fuel is separated from the protective metal sheath in which it is enclosed by treating the sheath with an embrittling agent which effects embrittlement of the sheath, breaking up the sheath and separating the nuclear fuel material from the embrittled and broken-up sheath.

Abstract: Process for recovering compounds of enriched nuclear fuel from scrap materials is disclosed. The process yields an acid solution with the nuclear fuel dissolved therein. Except for spent filter media, the scrap materials are calcined to produce an oxidized material, which is reduced to a particulate material and leached in an acid bath that contains some recycled acid to yield an acid solution of the nuclear fuel material. The scrap materials comprised of spent filter media are mulched and contacted with the acid solution. The insoluble materials are separated from the acid solution, and at least a portion of the acid solution is recycled to the leaching step, while the remainder of the acid solution is collected for subsequent treatment to recover the dissolved nuclear fuel material. Apparatus for performing the foregoing process is also disclosed.

Abstract: A graphite can be easily and completely gasified by soaking the graphite in a catalyst solution of a nitrate of a gasification catalyst selected from the group consisting of cobalt, nickel and iron and nitric acid to impregnate the graphite with the gasification catalyst and then gasifying the thus catalyst-impregnated graphite at temperatures above 1000.degree. C. And further a graphite can be easily and completely gasified by soaking the graphite in a catalyst solution of cobalt nitrate, sodium nitrate and nitric acid to add the mixing catalyst of cobalt and sodium to the graphite uniformly and then gasifying the thus catalyst-added graphite at temperatures below 900.degree. C.

Abstract: Irradiated nuclear fuel is dissolved in nitric acid under a carbon dioxide atmosphere which is treated to remove iodine, krypton and xenon and recycled.

Abstract: Process for the extraction of fission products contained in irradiated nuclear fuel elements which have been subject to a temperature of at least 1200.degree. C. during their irradiation prior to dissolving the fuel by the wet process. After mechanically treating the elements in order to decan and/or cut them they are brought into contact with water in order to pass the fission products into aqueous solution. The treated elements are then separated from the thus obtained aqueous solution. At least one of the fission products is then recovered from the aqueous solution. The fission products are iodine, cesium, rubidium and tritium.

Abstract: A process for recovering compounds of enriched nuclear fuel from scrap materials is disclosed. The process yields an acid solution with the nuclear fuel dissolved therein. Except for spent filter media, the scrap materials are calcined to produce an oxidized material, which is reduced to a particulate material and leached in an acid bath that contains some recycled acid to yield an acid solution of the nuclear fuel material. The scrap materials comprised of spent filter media are mulched and contacted with the acid solution. The insoluble materials are separated from the acid solution, and at least a portion of the acid solution is recycled to the leaching step, while the remainder of the acid solution is collected for subsequent treatment to recover the dissolved nuclear fuel material. Apparatus for performing the foregoing process is also disclosed.

Abstract: Method for removing fuel pellets from a nuclear fuel element without damaging the fuel pellets or fuel element sheath so that both may be reused. The method comprises holding the fuel element while a high pressure stream internally pressurizes the fuel element to expand the fuel element sheath away from the fuel pellets therein so that the fuel pellets may be easily removed.

Abstract: A process is provided involving the use of hydrogen for the separation of uranium and plutonium and mixtures thereof, from composite substances and assemblies or mixtures containing metallic and ceramic components where the metallic component is zirconium or a zirconium alloy and the ceramic component contains uranium and/or plutonium.

Abstract: There is provided a process for recovering block shaped graphite fuel elements made of a graphite block with a parallel cooling channel and which block contains the fuel arranged in separate zones. The process is especially useful for such fuel elements in which the cooling channels and the fuel zones are distributed in a screenlike manner adjacent to each other and wherein these fuel elements are first comminuted mechanically and then are reconditioned by burning and wet chemical separation into fissile material and waste material. For the mechanical comminution of the graphite block conical pins (or mandrels) are simultaneously pressed into several cooling channels until the block is broken apart along the entire length of these cooling channels.

Abstract: A process for separately recovering uranium, plutonium and neptunium substantially free of fission products from irradiated nuclear fuel is presented in which the fuel is dissolved in a strong mineral acid forming an aqueous dissolved nuclear fuel solution and treated to separate the uranium, plutonium and neptunium therefrom substantially free of said fission products by the sequential steps of solvent extraction, ion exchange and fluorination. The process has an improvement comprising the addition of a sufficient quantity of an additive of a stable metallic complex to the aqueous dissolved nuclear fuel solution prior to solvent extraction. This achieves improved purity of the separated uranium, plutonium and neptunium.

Abstract: There are prepared fuel elements for high temperature reactors from which the fuel zone can be removed from the structural graphite after the burnup of the fissile material has taken place so that the fuel element can be filled with new fuel and again placed in the reactor by having the strength of the matrix in the fuel zone sufficient for binding the embedded coated fuel particles but substantially less than the strength of the structural graphite whereby by the action of force it can be easily split up without destroying the particles.

Abstract: A method is disclosed for eliminating carbon from nuclear fuel elements in the reprocessing thereof in a closed system wherein such carbon is caused to combine with hydrogen to form methane in an attack zone, and the methane is then caused to flow from the attack zone to a regeneration zone where it is cracked into carbon and hydrogen.

Abstract: The present invention provides an improved dissolution process for ZrO.sub.2 -UO.sub.2 -CaO-type pressurized water reactor fuels. The zirconium cladding is dissolved with hydrofluoric acid, immersing the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers in the resulting zirconium-dissolver-product in the dissolver vessel, and nitric acid is added to the dissolver vessel to facilitate dissolution of the uranium from the ZrO.sub.2 -UO.sub.2 -CaO fuel wafers.

Abstract: An improved bottom assembly is provided for a nuclear reactor fuel reprocessing dissolver vessel wherein fuel elements are dissolved as the initial step in recovering fissile material from spent fuel rods. A shock-absorbing crash plate with a convex upper surface is disposed at the bottom of the dissolver vessel so as to provide an annular space between the crash plate and the dissolver vessel wall. A sparging ring is disposed within the annular space to enable a fluid discharged from the sparging ring to agitate the solids which deposit on the bottom of the dissolver vessel and accumulate in the annular space. An inlet tangential to the annular space permits a fluid pumped into the annular space through the inlet to flush these solids from the dissolver vessel through tangential outlets oppositely facing the inlet.

Abstract: After a first extraction of an aqueous nitric acid solution of spent nuclear fuel by a suitable organic solvent and a first washing of the resulting organic phase by an aqueous solution to remove tritiated water, the organic phase is submitted to a second washing with a volume of dilute aqueous solution of nitric acid free from tritium, said volume being substantially smaller than that of the organic phase, whereby the organic phase is rendered substantially tritium-free and contamination by tritium in the following stages of the process is avoided.

Abstract: Mercury can be recovered from nitric acid-containing fluids by reacting the fluid with aluminum metal to produce mercury metal, and then quenching the reactivity of the nitric acid prior to nitration of the mercury metal.

Abstract: Contaminated surface layers are decontaminated by treatment with an aqueous fluorine base-containing decontamination solution. The aqueous decontamination solution contains 0.05 to 50 Mol of decontamination agent per liter, and the decontamination agent preferably comprises at least one substance from the group.Iadd.: .Iaddend..[.colon.]. hexafluorosilicate acid, fluoroboric acid, and the salts of both of these. The decontamination solution produces the required high decontamination factors on metallic substances and brickworks as well. The used decontamination solution can, after regeneration, be recycled into the decontamination process.Release of decontaminated material by dissolution of the surface layer of the decontaminated objects provides decontamination of objects having complicated and hard-to-measure geometries.The decontamination agent (HBF.sub.4 -acid) is advantageously produced from contaminated boric acid from pressurized water reactor wastes by reaction with fluoride or hydrofluoric acid.