Abstract: According to one aspect of the invention, a method to create a ceramic waste form from used nuclear fuel. An active metal salt waste, a rare earth metal waste, and raw materials are received. The active metal salt waste is combined with the rare earth metal waste, forming a waste salt. The waste salt is then heated to approximately 500° C. The raw materials are also heated to approximately 500° C. The waste salt and raw materials are then blended to form a homogenous waste mixture. The homogenous waste mixture is heated to a first predetermined temperature for a predetermined amount of time, creating a ceramic waste form. The ceramic waste form is cooled to a second predetermined temperature.

Abstract: This invention relates to the vitrification of radioactive waste products. According to this invention, a glass composition that is suitable for mixed waste products, which include flammable waste products, such as gloves, working clothes, plastic waste, and rubber, and low-level radioactive waste products, and a method of vitrifying the mixed waste products using the same are provided to significantly reduce the volume of radioactive waste products and to vitrify the mixed waste products using the glass composition, which is suitable for vitrifying the mixed waste products, thereby maximally delaying or completely preventing the leakage of radioactive materials from a glass solidified body.

Abstract: This invention relates to the vitrification of radioactive waste products. According to this invention, a glass composition, which is suitable for low-level radioactive waste resins, and a method of vitrifying the low-level radioactive waste resins using the same are provided to significantly reduce the volume of radioactive waste products and to vitrify low-level radioactive waste products using the glass composition, which is suitable for vitrifying the low-level radioactive waste resins, thereby maximally delaying or completely preventing the leakage of radioactive materials from a glass solidified body.

Abstract: Systems and methods for capturing waste are disclosed. The systems and methods provide for a high level of confinement and long term stability. The systems and methods include adsorbing waste into a metal-organic framework (MOF), and applying pressure to the MOF material's framework to crystallize or make amorphous the MOF material thereby changing the MOF's pore structure and sorption characteristics without collapsing the MOF framework.

Abstract: A radioactive waste (zeolite to which Cs-137 was adsorbed) in a waste tank and a glass raw material (soda lime glass) in a glass raw material tank are supplied into a solidifying vessel. Graphite in a graphite tank is also supplied into the solidifying vessel. The solidifying vessel is filled with a mixture of the radioactive waste, glass raw material, and graphite and is then disposed in an adiabatic vessel. The radioactive waste and glass raw material in the adiabatic vessel are heated by thermal energy generated due to radiation emitted from Cs-137. The heat is transferred to the peripheral portion of the solidifying vessel through the graphite, raising the temperature of the peripheral portion. The glass raw material is melted and enters clearances among the radioactive waste, producing a vitrified radioactive waste. This radioactive waste solidification method can shorten a time taken to produce a vitrified radioactive waste.

Abstract: A fabric phase sorptive extractor (FPSE) is a sampling device where a flexible fabric is coated with at least one sol-gel derived film that includes at least two of a metal oxide portion, a siloxy portion, and an organic portion, where the gel has at least some amorphous portions. The FPSE is flexible such that it can be used in an extended form or draped over a solid surface to contact a gaseous, liquid, or solid environment and can be manipulated for placement in a container where the absorbed analyte can be removed from the FPSE for instrumental analysis. The FPSE can have specific functionalities that bind to specific analytes or can provide a general sorbent medium for extraction of a wide range of analytes, such that the sampling device can be employed for sampling analytes with biological, environmental, food, pharmaceutical, bio-analytical, clinical, forensic, toxicological, national security, public health, and/or safety implications.

Abstract: Dispersions of superabsorbent polymers are described comprising blends of polyols and super absorbent polymer. Products comprising deposits formed from such dispersions are also described. Suitable polyols can include, for example, polyethylene glycol. The dispersions can be flowed onto an absorbent sheet to form a deposit (sheeted SAP), which in some embodiments can be along a fraction of the sheet surface. Generally, the deposits are dried through a wicking process in which the polyol is wicked away from the superabsorbent polymer. The sheeted SAP can be incorporated into a final product with the deposit secured within the product. The dispersion allows for selective placement of superabsorbent polymer onto an absorbent sheet and/or in a final product. Embodiments based on coated yarns and the like are also presented.

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: Provided is a vitrification process method of aluminum and filter radioactive wastes to produce high quality of glass solidification fit for legislations and rules as vitrification final product, comprising developing frit composition needed in vitrifying the aluminum and filter radioactive wastes, suitably mixing the aluminum and filter radioactive wastes with the frit and producing glass solidification having composition range of oxides of aluminum and filter radioactive wastes to maintain lower than 100 poise viscosity which is operating parameter of a melting furnace. The vitrification process method of aluminum and filter radioactive wastes comprise mixing the aluminum and filter radioactive wastes with the frit in an induction heating low temperature melting furnace and meting it at the temperature of 1,100˜1,200° C. to produce glass solidification.

Abstract: The present invention provides systems, methods and devices for storing and/or disposing of hazardous waste material such as calcined material. In certain embodiments, the system comprises a filling nozzle having a valve body having a distal end and an outer surface, the outer surface proximate the distal end being configured to sealingly and removeably couple to an inner surface of a filling port of the container. In certain embodiments, the method comprises (a) coupling an outer surface of a filling nozzle with an inner surface of a filling port of a container to form a first seal (b) adding hazardous waste material into the container (c) decoupling the filling port from the filling nozzle and (d) inserting a fill plug into the filling port, the fill plug forming a second seal with the inner surface of the filling port, the second seal being distally spaced from at least a portion of the first seal with respect to the container.

Abstract: A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

Abstract: The present disclosure relates to a method of consolidating a calcine comprising radioactive material, the method comprising mixing 60-80% (by weight) of a radionuclide containing calcine with at least one non-radioactive additive, such as an oxide, and hot isostatic pressing the mixture to form a stable monolith of glass/ceramic. In one embodiment, the ratio of radionuclide containing calcine to additives is about 80:20 by weight, wherein the non-radioactive additive comprises oxides such as BaO, CaO, Al2O3, TiO2, SiO2 and others, that combine with the waste elements and compounds to form a ceramic mineral or glass/ceramic material, after hot isostatic pressing. Non-limiting examples of mineral phases that may be formed are: hollandite (BaAl2Ti6O16), zirconolite (CaZrThO7), and perovskite (CaTiO3).

Abstract: Disclosed herein is a method for preparing a ceramic waste form containing radioactive rare-earth and transuranic oxide, and the ceramic waste form with enhanced density, heat-stability, and leach resistance prepared by the same.

Abstract: A method for vitrification of waste to reduce the formation of persistent secondary phases comprising separating at least one glass frit constituent from an initial glass frit to form a modified glass frit. The waste, modified glass frit, and the at least one glass frit constituent are mixed together with the modified glass frit and the at least one glass frit constituent being added as separate components. The resulting mixture is vitrified.

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: 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 vitrification of high level waste to reduce the formation of persistent secondary phases comprising the steps of providing a high level waste for vitrification; providing a glass frit additive for mixing with said high level waste; redistributing selected constituents of said glass frit for mixing separately as raw chemicals with said high level waste; and, feeding said high level waste, said glass frit additive, and the redistributed glass frit constituents to a melter for vitrification of said high level waste so that formation of secondary phases is suppressed.

Abstract: A matrix material for safe temporary and/or ultimate disposal of radioactive wastes suitable for the embedment of radioactive wastes, contains graphite and at least inorganic binder which can be glass, aluminosilicate, silicate, borate and lead sulfide.

Abstract: The present invention relates to a process for packaging radioactive wastes, in which the following successive steps are carried out: a/ radioactive wastes, the solids content of which comprises at least 90% of compounds selected from CaCO2, Fe2O3, SiO2, Al2O3 and B2O3, are supplemented, so as to achieve a target composition of said supplemented wastes after calcination, and b/ said supplemented radioactive wastes are melted and c/ said melt is poured into a container, so as to obtain, after cooling, a product comprising a vitreous or vitro-crystalline synthetic rock, having said target composition, characterized in that said target composition corresponds to the following definition, in a CaO, SiO2 and X2O3 ternary system, in which X2O3 is a trivalent oxide or a mixture of trivalent oxides selected from Al2O3, Fe2O3 or B2O3, PC and PS being the mass percentages of CaO and SiO2: PC is from 35 to 60%, and PS is from 10 to 45%.

Abstract: The present disclosure relates to a method of consolidating a calcine comprising radioactive material, the method comprising mixing 60-80% (by weight) of a radionuclide containing calcine with at least one non-radioactive additive, such as an oxide, and hot isostatic pressing the mixture to form a stable monolith of glass/ceramic. In one embodiment, the ratio of radionuclide containing calcine to additives is about 80:20 by weight, wherein the non-radioactive additive comprises oxides such as BaO, CaO, Al2O3, TiO2, SiO2 and others, that combine with the waste elements and compounds to form a ceramic mineral or glass/ceramic material, after hot isostatic pressing. Non-limiting examples of mineral phases that may be formed are: hollandite (BaAl2Ti6O16), zirconolite (CaZrThO7), and perovskite (CaTiO3).

Abstract: A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

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: The present invention relates to a process for the manufacture of a glass frit for the containment by vitrification of a material comprising at least one oxidizable or reducible chemical species, and also to a process for the containment of said material by vitrification. The process for the manufacture of the glass frit comprises a step of incorporating into a raw glass frit at least one redox couple, the nature and the amount of which make it possible to maintain said at least one chemical species in its oxidized or reduced state. The containment process comprises mixing and hot melting the resulting glass frit and the material to be contained. The present invention makes it possible to optimize the containment of pollutants such as radionucleides, metals and metalloids. The material may be nuclear waste or a material derived from the incineration of household refuse.

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: 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: The present invention relates to a process for packaging radioactive wastes, in which the following successive steps are carried out: a/ radioactive wastes, the solids content of which comprises at least 90% of compounds selected from CaCO2, Fe2O3, SiO2, Al2O3 and B2O3, are supplemented, so as to achieve a target composition of said supplemented wastes after calcination, and b/ said supplemented radioactive wastes are melted and c/ said melt is poured into a container, so as to obtain, after cooling, a product comprising a vitreous or vitro-crystalline synthetic rock, having said target composition, characterized in that said target composition corresponds to the following definition, in a CaO, SiO2 and X2O3 ternary system, in which X2O3 is a trivalent oxide or a mixture of trivalent oxides selected from Al2O3, Fe2O3 or B2O3, PC and PS being the mass percentages of CaO and SiO2: PC is from 35 to 60%, and PS is from 10 to 45%.

Abstract: Provided is a method that can remove fission products in a spent electrolyte produced in a dry reprocessing process by an easy operation and can vitrify the fission products easily, the fission products including not only the fission products that generate precipitate but also the fission products that generate no precipitate. A spent electrolyte produced in a dry reprocessing process is subjected to a phosphate conversion processing to obtain a processing target substance; the processing target substance is passed through a separating material 10 including an iron phosphate glass at a temperature of not more than a softening point of the iron phosphate glass in order to remove insoluble fission products included in the processing target substance by filtration with the separating material and to sorb fission products in solution to the separating material for separation; and the iron phosphate glass holding the fission products is used as a waste vitrification material.

Abstract: Apparatuses, processes and methods for the separation, isolation, or removal of specific radioactive isotopes from liquid radioactive waste, these processes and methods employing isotope-specific media (ISM). In some embodiments, the processes and methods further include the vitrification of the separated isotopes, generally with the ISM; this isotope-specific vitrification (ISV) is often a step in a larger scheme of preparing the radioactive isotopes for long-term storage or other disposition. A variety of ISM are disclosed.

Abstract: The invention concerns a process enabling the complete combustion and oxidation of the mineral fraction of combustible waste contained in an apparatus intended to treat waste by direct incineration-vitrification, said process comprising the following steps: a step to add said waste to the apparatus for its depositing on the surface of a molten glass bath contained in the apparatus, an incineration and oxidation step of the waste on the surface of the glass bath, an incorporation step to incorporate combustion products in the glass during which the glass bath, the combustion products and any vitrification additives added to the glass bath are heated until a paste-like, liquid mass is obtained, a step during which said mass is removed from the apparatus and left to cool to obtain finally what is called a confinement matrix, said process being characterized in that the complete combustion and oxidation of the waste is achieved partly during the waste adding step and partly during the waste incineration and ox

Abstract: A method for vitrification of high level waste to reduce the formation of persistent secondary phases comprising the steps of providing a high level waste for vitrification; providing a glass frit additive for mixing with said high level waste; redistributing selected constituents of said glass frit for mixing separately as raw chemicals with said high level waste; and, feeding said high level waste, said glass frit additive, and the redistributed glass frit constituents to a melter for vitrification of said high level waste so that formation of secondary phases is suppressed.

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: The present disclosure relates to a process for recycling a sodium salt by decomposition of a sodium nitride liquid waste, comprising a neutralization step in which a nitric acid liquid waste or an off-gas having nitric acid dissolved therein which is produced through a wet reprocessing process comprising a dissolution step for dissolving a spent nuclear fuel in nitric acid is neutralized by adding or contacting the nitrate liquid waste or the off-gas to or with at least one sodium salt selected from the group consisting of sodium hydroxide, sodium hydrogencarbonate and sodium carbonate, thereby yielding a sodium nitrate liquid waste; a sodium nitrate-decomposition step in which the sodium nitrate liquid waste is reductively decomposed with a reducing agent, thereby decomposing sodium nitrate into a nitrogen gas and the sodium salt; and a recycle step for recycling the sodium salt into the neutralization step or wet reprocessing process.

Abstract: A substantially alkaline resistant calcium-iron-phosphate (CFP) glass and methods of making and using thereof. In one application, the CFP glass is drawn into a fiber and dispersed in cement to produce glass fiber reinforced concrete (GFRC) articles having the high compressive strength of concrete with the high impact, flexural and tensile strength associated with glass fibers.

Abstract: The mass to be vitrified undergoes a reduction operation in order to have the ruthenium pass from an oxidized state to a metal state in order to reduce the viscosity, the electric conductivity and to obtain good chemical kinetics.

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 method of disposing of waste material in a waste stream, including positioning a porous foamed glass member characterized by an open-cell interconnected pore network in contact with a volume of liquid to be purified and removing an amount of an undesired material from the volume of liquid.

Abstract: The present invention relates to a process for the manufacture of a glass frit for the containment by vitrification of a material comprising at least one oxidizable or reducible chemical species, and also to a process for the containment of said material by vitrification. The process for the manufacture of the glass frit comprises a step of incorporating into a raw glass frit at least one redox couple, the nature and the amount of which make it possible to maintain said at least one chemical species in its oxidized or reduced state. The containment process comprises mixing and hot melting the resulting glass frit and the material to be contained. The present invention makes it possible to optimize the containment of pollutants such as radionucleides, metals and metalloids. The material may be nuclear waste or a material derived from the incineration of household refuse.

Abstract: A process for melting material to be treated includes placing material to be treated in a container that may include an insulating lining, heating the material to be treated and melting the material to be treated, preferably allowing the melted material to cool to form a vitrified and/or crystalline mass, and disposing of the mass. The mass is either disposed while contained in container or removed from container after cooling and disposed. Insulating lining may comprise one or more layers of a thermal insulating material, one or more layers of refractory material, or a combination thereof. The material to be treated may be heated by placing at least two electrodes in the material to be treated and passing a current between the electrodes, or alternatively, by placing at least one heating element in the material to be treated and passing heat into the material to be treated.

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: The invention relates to a confinement matrix for the storage or incineration of at least one long-life radioactive element, comprising at least one crystalline boron compound of a rhombohedral structure including the long-life radioactive element(s). The boron compound may be of the B3Si, B6O or B4C type.

Abstract: The present invention relates to an immobilising medium for the encapsulation of radioactive waste. The waste immobilising 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 method of treating aqueous salt solutions to provide a solution suitable for vitrification to a stable glass matrix for long term storage is described. In particular, salt solutions composed of aqueous nuclear waste materials are suitable for treatment by the described method. Specifically, salt solutions which have a sulfate to sodium mole ratio that does not permit easy vitrification into stable glasses may be treated by the present invention. The present method decreases the volume of vitrified glass.

Abstract: Hollow glass microspheres obtained from fly ash (cenospheres) are impregnated with extractants/ion-exchangers and used to remove hazardous material from fluid waste. In a preferred embodiment the microsphere material is loaded with ammonium molybdophosphonate (AMP) and used to remove radioactive ions, such as cesium-137, from acidic liquid wastes. In another preferred embodiment, the microsphere material is loaded with octyl(phenyl)-N—N-diisobutyl-carbamoylmethylphosphine oxide (CMPO) and used to remove americium and plutonium from acidic liquid wastes.

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: A porous sintered body in which the leaching amounts of heavy metals are small and having high availability contains 0.5-15 weight % of B2O3, preferably contains 20-60 weight % of Al2O3, 18-60 weight % of SiO2, 1-12 weight % of the sum of Na2O, K2O, Li2O, and P2O5, 1-30 weight % of the sum of CaO, SrO, BaO, and MgO, and 0.5-15 weight % of B2O3.

Abstract: The subject of the present invention is a process for immobilizing metallic sodium in glass form, comprising the steps consisting in:

Abstract: A method for vitrifying a plurality of nuclear waste kernels includes coating the kernels with a glass layer, and mixing the glass-coated kernels in a glass melt. Subsequent cooling solidifies the glass melt and vitrifies the nuclear waste kernels in bulk vitrification glass. Importantly, the glass layer has a softening temperature that is higher than the softening temperature of the glass melt. The glass layer also has a variable thermal expansion coefficient across the layer. Additionally, the glass melt has substantially the same specific gravity as the glass-coated kernels in order to effect a uniform distribution of the glass-coated kernels throughout the bulk vitrification glass.