Abstract: A method for treatment of spent ion-exchange resins for disposal includes feeding a mixture of spent ion-exchange resins to the a loading tank, separating the ion-exchange resins, feeding separated ion-exchange resins into the a drying chamber, vacuum drying the ion-exchange resins and subjecting the resins to additional heat treatment in a high-temperature furnace, and unloading the treated ion-exchange resins into a transport container. A device for treatment of spent ion-exchange resins includes a loading tank, a metering device connected to a drying chamber, an inclined feed screw located between the loading tank and the metering device, a vacuum pump, a heated gas filter, a high-temperature furnace equipped with a vacuum drying and gas purification system, and a feeding device located between the drying chamber and the high-temperature furnace. A docking unit is connected to a lower part of the high-temperature furnace.

Abstract: Devices for eliminating radioactive contamination of radioactive waste by providing adaptive processing of the decontamination solution for reuse. The plant for electrochemical decontamination of metal radioactive waste includes a pipe equipped with shut-off valves, a radioactive waste processing module that comprises a unit for electrochemical decontamination connected by a ventilation channel to the ventilation module and pipe for decontamination solution supply and discharge equipped with shut-off valves.

Abstract: A canister for interim storage and subsequent consolidation of waste materials via hot pressing and comprising at least one ion exchange material. The canister is configured to house the ion exchange material after it is exchanged with a contaminating ion without releasing the contaminating ion and to consolidate waste materials via hot-isostatic pressing. A method comprising contacting a fluid waste with an ion exchange material.

Abstract: A container for the consolidation of waste materials including radioactive containing waste, and a method of consolidating such materials. The container comprises an outer cylinder and an inner cylinder comprising internal compression plates that are designed to resist collapse during consolidation, and therefore control the size of the consolidated container to a predictable shape and dimension. The container is sufficient to hold a variety of materials, including hazardous, toxic, or radioactive waste, and the container is configured to hold such waste without releasing it to the environment.

Abstract: A method for vitrifying waste to prevent the formation of molybdate secondary phases includes forming a feed mixture that includes the waste, a source of vanadium, and at least one of glass frit or glass forming chemicals and vitrifying the feed mixture in a melter to produce a glass product that includes the waste.

Abstract: A cement curing formulation and curing method for high-level radioactive boron waste resins from a nuclear reactor. The curing formulation comprises the following raw materials: cement, lime, water, curing aids and additives. The curing method comprises: (1) weighing the raw materials and the high-level radioactive boron waste resins, and adding lime into a curing container; (2) then adding the high-level radioactive boron waste resins; (3) feeding other raw materials under stirring; (4) adding the cement and supplementing water depending on the moisture state of the cement, and stirring until uniform; and (5) standing and maintaining after stirring until uniform. The curing formulation has the features of a high curing containment rate, high strength of the cured body, better water resistance, better freeze-thaw resistance, and low radioactive leakage.

Abstract: Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix.

Abstract: Methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials are described. For example, a method of capturing and immobilizing radioactive nuclei includes flowing a gas stream through an exhaust apparatus. The exhaust apparatus includes a metal fluorite-based inorganic material. The gas stream includes a radioactive species. The radioactive species is removed from the gas stream by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus.

Abstract: A method and device for limiting the degassing of tritiated waste issued from the nuclear industry are provided. The method reduces an amount of generated tritiated hydrogen (T2 or HT) and/or tritiated water (HTO or T2O) including at least one piece of tritiated waste from the nuclear industry. The method includes placing the package in contact with a mixture including manganese dioxide (MnO2) combined with a component that includes silver; and placing the package in contact with a molecular sieve.

Abstract: Methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials are described. For example, a method of capturing and immobilizing radioactive nuclei includes flowing a gas stream through an exhaust apparatus. The exhaust apparatus includes a metal fluorite-based inorganic material. The gas stream includes a radioactive species. The radioactive species is removed from the gas stream by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus.

Abstract: Disclosed herein is a method of treating radioactive metal waste using melt decontamination, wherein radioactive metal waste, which is generated from nuclear fuel processing facilities or nuclear fuel production facilities, and which cannot be easily treated by surface decontamination because it has a complicated geometric shape, and the surface contamination of which cannot be measured, can be treated by melt decontamination. The method is advantageous in that radioactive metal waste, which cannot be treated by conventional surface decontamination, can be treated, so that radioactive metal waste can be recycled, thereby obtaining economic profits, and further in that a large storage space necessary for cutting and then storing radioactive metal waste is not required, and in that excessive manpower and cost are not required.

Abstract: There is provided a process for decontaminating and converting volumetrically contaminated radioactive metals, especially nickel, and recovering a decontaminated metal hydroxide or metal carbonate. The process includes the use of hydrogen peroxide to oxidize and remove nucleotides.

Abstract: Disclosed is a method of disposing of radioactive metal waste using melting decontamination, including sorting radioactive metal waste generated in nuclear fuel processing or production facilities by predetermined sorting criteria, and charging sorted metal waste into a melting furnace so as to be melted; adding a impurity remover to the melt of the melting furnace to remove generated slag; pouring the melt having no slag into a mold to form an ingot; subjecting the ingot to gamma spectroscopy using a gamma spectrometer to measure gamma rays of U-235 (185.72 keV, 57.2%) among uranium isotopes, performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using MCNP computer code, and calculating total radioactivity of the ingot from the quantified radioactivity and mass of U-235; and efficiently and rapidly determining whether the ingot subjected to radioactivity measurement satisfies a clearance limit.

Abstract: The present invention relates to a radioactive cesium adsorbent, a method for producing the same, and a method for decontaminating the environment from radioactive cesium with the adsorbent. The radioactive cesium adsorbent of the present invention includes a hydrophilic fiber substrate supporting a Prussian blue analogue, in particular, Prussian blue, and the Prussian blue analogue is immobilized in the inside of the fibers.

Abstract: Disclosed herein is a method of treating radioactive metal waste using melt decontamination, wherein radioactive metal waste, which is generated from nuclear fuel processing facilities or nuclear fuel production facilities, and which cannot be easily treated by surface decontamination because it has a complicated geometric shape, and the surface contamination of which cannot be measured, can be treated by melt decontamination. The method is advantageous in that radioactive metal waste, which cannot be treated by conventional surface decontamination, can be treated, so that radioactive metal waste can be recycled, thereby obtaining economic profits, and further in that a large storage space necessary for cutting and then storing radioactive metal waste is not required, and in that excessive manpower and cost are not required.

Abstract: Disclosed is a method of disposing of radioactive metal waste using melting decontamination, including sorting radioactive metal waste generated in nuclear fuel processing or production facilities by predetermined sorting criteria, and charging sorted metal waste into a melting furnace so as to be melted; adding a impurity remover to the melt of the melting furnace to remove generated slag; pouring the melt having no slag into a mold to form an ingot; subjecting the ingot to gamma spectroscopy using a gamma spectrometer to measure gamma rays of U-235 (185.72 keV, 57.2%) among uranium isotopes, performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using MCNP computer code, and calculating total radioactivity of the ingot from the quantified radioactivity and mass of U-235; and efficiently and rapidly determining whether the ingot subjected to radioactivity measurement satisfies a clearance limit.

Abstract: The present invention provides processes to immobilize high alkaline radioactive and/or hazardous waste in a silicate-based glass, the waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds. The invention also provides silicate-based glass compositions for use in immobilizing radioactive and/or hazardous waste.

Abstract: Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

Abstract: A method of rendering hazardous materials less dangerous comprising trapping the hazardous material in nanopores of a nanoporous composite material, reacting the trapped hazardous material to render it less volatile/soluble, sealing the trapped hazardous material, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

Abstract: Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI2, CuI, or Bi5O7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425° C. to 550° C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500° C. (below the silver iodide sublimation temperature of 500° C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

Abstract: A detoxicant solution includes 0.1 to 50 parts by weight of at least one Friedel-Crafts reagent (Lewis acids) (I); 0.5 to 80 parts by weight of one or more of the following specific solvents (II): dimethyl sulfoxide, 1,3-dimethyl-2-imidazolidinone or octamethylcyclotetrasiloxane; and 1 to 80 parts by weight of an alcoholic solvent (III). The term “parts by weight” relates in each case to 100 parts by weight of the complete detoxicant solution.

Abstract: The present invention provides processes to immobilize high alkaline radioactive and/or hazardous waste in a silicate-based glass, the waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds. The invention also provides silicate-based glass compositions for use in immobilizing radioactive and/or hazardous waste.

Abstract: A method for treating a nitric aqueous liquid effluent containing nitrates of metals or metalloids, comprising a step for calcination of the effluent in order to convert the nitrates of metals or metalloids into oxides of said metals or metalloids, at least one compound selected from the nitrates of metals or metalloids and the other compounds of the effluent leading upon calcination to a tacky oxide, and a dilution adjuvant leading upon calcination to a non-tacky oxide being added to the effluent prior to the calcination step, a method wherein the dilution adjuvant comprises aluminium nitrate and at least one nitrate selected from iron nitrate and rare earth nitrates.

Abstract: The invention relates to a canister (1) for final repository of spent fuel elements from a nuclear reactor, comprising an insert (2) that contains said spent fuel elements, an inner copper casing (4a, 4b, 4c) that encloses the insert (2), and at least one outer casing (5a, 5b, 5c) that encloses the copper casting and that consists of a passive-film-forming metal or metal alloy, the passive film on the casing being constituted by an essentially oxidic film that is rich in one or more of the metals in the group of metals that consist of the metals zirconium, chromium and titanium.

Abstract: Systems and processes for reducing the volume of radioactive waste materials through pyrolysis and vitrification carried out by microwave heating and, in some instances, a combination of microwave heating and inductive heating. In some embodiments, the microwave-enhanced vitrification system comprises a microwave system for treating waste material combined with a modular vitrification system that uses inductive heating to vitrify waste material. The final product of the microwave-enhanced vitrification system is a denser, compacted radioactive waste product.

Abstract: A radioactive containment composition may be created for containing radionuclides from a radioactive material by mixing a clay mineral with water. This mixture may form an aqueous clay suspension. The mixture can be refined by filtering to remove coarse material. The aqueous clay suspension may be applied to a radioactive material, allowing the radionuclides to be exchanged with cations in the aqueous clay suspension. The resulting aqueous slurry, a silver-based solution may be added to produce a suspension. The suspension may be collected, heated and analyzed.

Abstract: Process for confinement of waste containing at least one chemical species to be confined, by in-can vitrification in a hot metal can into which waste and a vitrification additive are added, the waste and the vitrification additive are melted to obtain a glass melt which is then cooled, characterised in that at least one oxidising agent is also added into the metal can and in that the concentration of oxidising agent(s) expressed as oxide(s) in the glass melt is between 0.1 and 20% by mass, preferably 4 and 20% by mass, even more preferably 5 and 15% by mass, and even more preferably 10 and 13% by mass of the glass melt mass.

Abstract: The present invention provides processes to immobilize radioactive and/or hazardous waste in a borosilicate glass, the waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds. The invention also provides borosilicate glass compositions for use in immobilizing radioactive and/or hazardous waste.

Abstract: Example embodiments and methods may provide segmented waste rods capable of containing and disposing of waste generated from spent nuclear fuel, including elements left over from fuel that has been harvested for desired isotopes produced in the fuel. Example methods may provide methods for forming and using example embodiment segmented waste rods.

Abstract: A radioactive containment composition may be created for containing radionuclides from a radioactive material by mixing a clay mineral with water. This mixture may form an aqueous clay suspension. The mixture can be refined by filtering to remove coarse material. The aqueous clay suspension may be applied to a radioactive material, allowing the radionuclides to be exchanged with cations in the aqueous clay suspension. The resulting aqueous slurry, a silver-based solution may be added to produce a suspension. The suspension may be collected, heated and analyzed.

Abstract: Briefly described, compositions, materials including the compositions, methods of using the compositions, and methods of degrading contaminants, are described herein. The composition can include a polyoxometalate/ cationic silica material. In addition, the compositions can be made of a polyoxometalate/cationic silica material, a copper (II) salt having a weakly bound anion, and a nitrate salts. Further, the compositions can be made of a polyoxometalate/cationic silica material, a copper (II) salt having a weakly bound anion, a compound selected from tetraethylammonium (TEA) nitrate, tetra-n-butylammonium (TBA) nitrate, and combinations thereof.

Abstract: Process for the stabilization of mercury metal by reaction of the mercury metal with sulphur in the solid state, in which the mercury and the sulphur are brought into contact, at an Hg/S molar ratio of 1/1 to 1/3, in a reactor integral with a hollow pipe in fluid communication with the interior space of the said reactor, the said hollow pipe comprising a first end connected to the wall of the said reactor and a second end distant from the said reactor; the said hollow pipe and the said reactor being hermetically sealed, the said hollow pipe being provided with rotating means external to the said pipe and to the said reactor for rotating the said reactor and the said pipe around the axis of the said pipe, and the said hollow pipe being provided, at its end distant from the reactor, with means for introducing the sulphur and the mercury inside the reactor and discharging the reaction products.

Abstract: The present invention provides processes to immobilize radioactive and/or hazardous waste in a borosilicate glass, the waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds. The invention also provides borosilicate glass compositions for use in immobilizing radioactive and/or hazardous waste.

Abstract: Radiation shields and techniques for radiation shielding are provided. Bitumen substances, such as asphalt or tar, are mixed with radioactive waste, leaded glass, or a radioactive waste and leaded glass composite. In embodiments where the bitumen substance is mixed with leaded glass, the resulting mixture can be coated onto containers that house radioactive waste or the resulting mixture can be coated onto the outer surface of the radioactive waste.

Abstract: A target material (60) to be treated in a liquid reactant metal is loaded into a containment area defined within a liquid reactant metal treatment vessel (11). The containment area is then placed below the level (L) of the liquid reactant metal in the treatment vessel (11). This places the target material (60) in contact with the liquid reactant metal and allows the desired reactions to occur. Reaction products are then removed from the treatment vessel (11). Placing the containment area below the level (L) of liquid reactant metal in the treatment vessel (11) may be accomplished by pivoting the vessel from a loading position to a treating position to shift the level of liquid reactant metal in the vessel.

Abstract: A hydrogen-trapping compound is provided, along with a process for manufacturing the compound, and its uses, wherein the hydrogen-trapping compound is characterized in that it contains at least one metal salt of formula MX(OH), in which M represents a divalent transition element, for example Co or Ni; O represents an oxygen atom; X represents an atom of group 16 of the Periodic Table of the Elements, excluding O, for example a sulphur atom; and H represents a hydrogen atom, and wherein the hydrogen-trapping compound is effective for trapping hydrogen, hydrogen within a material and free hydrogen and is applicable in situations in which hydrogen is evolved and in which it has to be trapped, especially for safety reasons.

Abstract: A method of controlling or containing radioactive contamination by providing a neutron absorbing material to a radioactive contamination site. Preferably the neutron absorbing material is present as a powder, granule, slurry or suspension, allowing the neutron absorbing material to blanket cover the radioactive contamination site. Suitable neutron absorbing materials include lanthanide elements having a cross section of 100 Barns or greater, as well as hafnium, zirconium, tantalum, silver, indium, and hydrogen.

Abstract: This invention deals with multi-component composite materials and techniques for improved shielding of neutron and gamma radiation emitting from transuranic, high-level and low-level radioactive wastes. Selective naturally occurring mineral materials are utilized to formulate, in various proportions, multi-component composite materials. Such materials are enriched with atoms that provide a substantial cumulative absorptive capacity to absorb or shield neutron and gamma radiation of variable fluxes and energies. The use of naturally occurring minerals in synergistic combination with formulated modified cement grout matrix, polymer modified asphaltene and maltene grout matrix, and polymer modified polyurethane foam grout matrix provide the radiation shielding product. These grout matrices are used as carriers for the radiation shielding composite materials and offer desired engineering and thermal attributes for various radiation management applications.

Abstract: The present invention relates to an immobilizing medium for the encapsulation of radioactive waste. The waste immobilizing medium has a sodium silicate based glass matrix in which there is contained radioactive waste wherein the waste comprises one or more inert metal components and one or more fission products. At least a portion of the inert metal components are dissolved in the glass matrix and increase its durability. As a result, the waste immobilising medium is highly durable and leach resistant and is suitable for long term storage of radioactive waste. The inert metal components preferably comprise iron, nickel and chromium.

Abstract: A process for immobilizing metallic sodium in glass form. The process comprises: (A) introducing in a dispersed state, into a reactor, an amount of a vitrified matrix precursor, metallic sodium and iron oxide (Fe2O3) sufficient to ensure oxidation of the metallic sodium; (B) producing a homogeneous mixture of these constituents; (C) heating the mixture to a temperature between 1000–1600° C. to form a molten homogeneous mixture in which the sodium introduced in (A) is converted to sodium oxide; and (D) recovering and cooling the molten mixture to obtain a vitrified matrix having a homogeneous composition, which matrix incorporates the sodium introduced in (A) as a constituent oxide. In a particular embodiment, the process may be used for the containment of metallic sodium containing radioactive elements.

Abstract: The present invention utilizes one or more processes to immobilize a waste that contains one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds present in the waste. Each of the processes create a barrier against leaching and diffusion of the wastes. The first barrier is created by integrating the waste with an immobilizing mineral. The second barrier is a layer of non-radioactive or non-hazardous material that covers the first barrier. The second barrier may be created using an overgrowth procedure or by sintering. The third barrier is created by a rock or glass matrix that surrounds the first and/or second barriers. The fourth barrier is created by ensuring that the rock or glass has the same or similar composition as the indigenous rock at the disposal site. The resultant rock or glass matrix is in equilibrium with the groundwater or local hydrothermal solutions that are saturated with components of the indigenous rock of the disposal area.

Abstract: A liquid reactant metal alloy includes at least one chemically active metal for reacting with non-radioactive material in a mixed waste stream being treated. The reactant alloy also includes at least one radiation absorbing metal. Radioactive isotopes in the waste stream alloy with, or disperse in, the chemically active and radiation absorbing metals such that the radiation absorbing metals are able to absorb a significant portion of the radioactive emissions associated with the isotopes. Non-radioactive constituents in the waste material are broken down into harmless and useful constituents, leaving the alloyed radioactive isotopes in the liquid reactant alloy. The reactant alloy may then be cooled to form one or more ingots in which the radioactive isotopes are effectively isolated and surrounded by the radiation absorbing metals. These ingots comprise the storage products for the radioactive isotopes. The ingots may be encapsulated in one or more layers of radiation absorbing material and then stored.

Abstract: The present invention provides processes to immobilize radioactive and/or hazardous waste in a borosilicate glass, the waste containing one or more of radionuclides, hazardous elements, hazardous compounds, and/or other compounds. The invention also provides borosilicate glass compositions for use in immobilizing radioactive and/or hazardous waste.

Abstract: A decontamination system uses magnetic molecules having ferritin cores to selectively remove target contaminant ions from a solution. The magnetic molecules are based upon a ferritin protein structure and have a very small magnetic ferritin core and a selective ion exchange function attached to its surface. Various types of ion exchange functions can be attached to the magnetic molecules, each of which is designed to remove a specific contaminant such as radioactive ions. The ion exchange functions allow the magnetic molecules to selectively absorb the contaminant ions from a solution while being inert to other non-target ions. The magnetic properties of the magnetic molecule allow the magnetic molecules and the absorbed contaminant ions to be removed from solution by magnetic filtration.

Abstract: The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods.

Abstract: Apparatus for containing, transporting, and storing or disposing of radioactive machinery, including decommissioned nuclear reactor pressure vessels. An improved, economically-produced container allows easier handling and packaging of machinery within plants where the machinery has been installed, and offers improved shock absorption and attenuation characteristics, especially when packaging is complete, through the provision, for example, of integral fenders. The invention also includes methods of making the container.

Abstract: A process intended to prevent sedimentation of gas hydrates in petroleum production installations comprises adding to the petroleum effluent at least one additive making it possible to modify the rheological behaviour of the petroleum effluent to impart thereto a flow threshold fluid behaviour.

Abstract: Encapsulating calcined radioactive waste in strong, corrosion-resistant spheres of dimensions such that heat from the radiation melts the ice at a rate which brings the spheres to the bottom of the permanent icefield in a relatively short time, with the resulting waste ultimately being no more hazardous than natural uranium ore.

Abstract: This invention relates to a method for conditioning a waste constituted of an aqueous solution of sodium hydroxide NaOH of 3 to 10 M, possibly radioactive. The method is as follows: a) a metakaolin powder is added to the aqueous solution such that a suspension is obtained capable of solidifying and forming a crystalline phase of the zeolite A type; b) the suspension is introduced into a mould; c) the suspension is left to solidify in the mould in order to obtain a moulded solid product based on zeolite A; d) the moulded product is dried; and e) the zeolite A phase is converted into a nepheline type phase by heat treatment at a temperature of 1000° C. to 1500° C.

Abstract: The invention relates to methods of vitrifying waste and for lowering the melting point of glass forming systems by including lithia formers in the glass forming composition in significant amounts, typically from about 0.16 wt % to about 11 wt %, based on the total glass forming oxides. The lithia is typically included as a replacement for alkali oxide glass formers that would normally be present in a particular glass forming system. Replacement can occur on a mole percent or weight percent basis, and typically results in a composition wherein lithia forms about 10 wt % to about 100 wt % of the alkali oxide glass formers present in the composition. The present invention also relates to the high lithia glass compositions formed by these methods.