Abstract: Waste solidification compositions and methods of using them repurpose calcium-containing industrial by-products. The compositions comprise either 1) a) auto shred residue; and b) a particulate wood-based product, or 2) a) a solid, particulate calcium-containing compound; and b) a superabsorbent material. The method of repurposing a solid, particulate calcium-containing industrial by-product comprises a) blending the by-product with a superabsorbent material to form a waste solidification composition; b) adding the waste solidification composition to a liquid industrial waste stream; and c) allowing the waste solidification composition to absorb at least 1 times its weight of the liquid industrial waste stream to form a solid waste product. The solid waste product passes Paint Filter Liquids Test Method 9095B.

Abstract: Systems and methods for long-term disposal of high-level nuclear waste in deep geologic formations may include largely intact spent nuclear fuel rod assemblies that may be placed into waste-capsules (e.g., carrier tubes); which may then be placed into various well boreholes. Example embodiments may provide waste-capsules capable of containing and disposing of nuclear waste generated from spent nuclear fuel; including methods for harvesting the nuclear waste from cooling pools and/or surface storage; and operationally processing the waste, fuel assemblies for inclusion in the waste-capsules with various engineered barriers; along with storage in relatively large substantially horizontal well boreholes; which may be drilled into closed deep geologic formations.

Abstract: Waste capsules for disposal of radioactive materials, including weapons grade plutonium, are made entirely from rock, such as, but not limited to, granite; wherein a cap of such a waste capsule is seamlessly rock welded to a lower body portion of the given waste capsule, to form the given waste capsule, with the radioactive materials inside of a cavity of that waste capsule. The weld region is homogenous with both the cap and the lower body portion. A cooling system may be used during or after rock welding. The sealed and rock welded waste capsule, with the internal radioactive materials, is then loaded into a wellbore system that extends into a deep geological rock repository, thousands of feet below the Earth's surface, such that the waste capsule comes to rest in a wellbore located within the deep geological rock repository. The waste capsule may include insulating material in the cavity.

Abstract: A volume reduction treatment method able to reduce the volume of the final disposal waste of a spent uranium catalyst. As a result, the disposal cost of the spent uranium catalyst is able to be reduced and the utilization of waste repositories are able to be improved.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: Provided is a method of treating radioactive liquid waste which reduces the amount of radioactive waste to be generated and is capable of removing a radioactive nuclide from radioactive liquid waste to the extent that the concentration thereof is less than or equal to the measurement lower limit using a simple apparatus configuration. A filtration device is connected to a colloid removal device by a connection pipe. An adsorption tower positioned at the highest stream of an adsorption device is connected to the colloid removal device by a connection pipe. The colloid removal device includes an electrostatic filter. Respective adsorption towers in the adsorption device are sequentially connected by a pipe. A discharge pipe is connected to the adsorption tower positioned at the lowest stream of the adsorption device. Radioactive liquid waste, containing particles having a particle diameter of 1 ?m or greater, negatively charged colloids, and a radioactive nuclide, is supplied to the filtration device.

Abstract: A method of decommissioning a nuclear reactor including a vessel defining an inner chamber and reactor internal components positioned within the inner chamber includes removing at least some of the internal components and repackaging at least some of the removed internal components in at least one of the first and second predefined sections of the vessel defining a cutting zone between the at least first and second predefined sections. The method further includes disposing the vessel in a container, and encapsulating the internal components in the vessel and encapsulating the vessel in the container to generate a package. A package including components from a decommissioned and dismantled nuclear reactor includes a vessel and a plurality of components encapsulated in the vessel.

Abstract: A method of immobilizing a radioisotope and vitrified chemically bonded phosphate ceramic (CBPC) articles formed by the method are described. The method comprises combining a radioisotope-containing material, MgO, a source of phosphate, and optionally, a reducing agent, in water at a temperature of less than 100° C. to form a slurry; curing the slurry to form a solid intermediate CBPC article comprising the radioisotope therefrom; comminuting the intermediate CBPC article, mixing the comminuted material with glass frits, and heating the mixture at a temperature in the range of about 900 to about 1500° C. to form a vitrified CBPC article comprising the radioisotope immobilized therein.

Abstract: System and Method is described that controls the release of contaminated water by rapidly freezing the ground water, including salt water, which permeates the area underneath the a contamination source, so that the resulting ice lens mitigates the extent to which radioactive water is released into the environment. An aperture in the containment area allows the dispersal and dilution of the contaminates by allowing in ground water from outside, and/or removing water from the containment area. The variable aperture may be a physical valve or preferably an opening in the ice shield which size may be controlled by freezing or thawing portions of the ice shield.

Abstract: The present invention provides an efficient and low cost method for processing radioactively-contaminated water. The method for processing radioactively-contaminated water comprising a freeze concentration step of generating ice having lowered concentration of radioactive substance from radioactive substance containing contaminated water and concentrating the radioactive substances in the residual contaminated water by the interface progressive freeze concentration process. Preferably, the method further comprises a nitrogen substitution step of reducing dissolved oxygen in the contaminated water and adding nitrogen gas to the contaminated water, as a previous step of the freeze concentration step. Preferably, the radioactive substance is radioactive cesium.

Abstract: A solidification method of radioactive waste is provided, including kneading a binder and an inorganic adsorbent to obtain a kneaded object, the in organic adsorbent included radionuclides; extruding the kneaded object to obtain an extruded material object; cutting the extruded material object to obtain at least one extruded material block; and firing the at least one extruded material block to solidify the at least one extruded material block.

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 describes solid waste forms and methods of processing waste. In one particular implementation, the invention provides a method of processing waste that may be particularly suitable for processing hazardous waste. In this method, a waste component is combined with an aluminum oxide and an acidic phosphate component in a slurry. A molar ratio of aluminum to phosphorus in the slurry is greater than one. Water in the slurry may be evaporated while mixing the slurry at a temperature of about 140-200° C. The mixed slurry may be allowed to cure into a solid waste form. This solid waste form includes an anhydrous aluminum phosphate with at least a residual portion of the waste component bound therein.

Abstract: A method of injecting mortar into a container fastened to a first tank and to a second tank, the first tank communicating with the container via a first orifice and the second tank communicating with the container by a second orifice, the method comprising the following operations: a continuous circulation of a first stream of mortar is made to flow in a circulation loop; during the continuous circulation, a second stream of mortar is drawn off from the circulation loop, the second stream being smaller than the first stream of mortar; the second stream of mortar is injected into the container, ensuring that there is dynamic confinement of the gaseous effluents; and the appearance of mortar in the second tank is monitored and, when this appearance is detected, the removal of mortar from the circulation loop is brought to an end.

Abstract: The invention provides a process for stabilizing waste by landfilling, including mixing the waste with ash after burning oil shale, whereby safely stabilizing inorganic or organic waste by embedding it in a hardened matter exhibiting a high compression strength.

Abstract: The invention relates to a coating composition for the storage or containment of waste that is toxic to health and/or the environment, comprising a composition based on an epoxy resin and a curing composition free from an aromatic amine curing agent. The invention also relates to the use of this composition for the coating of said waste.

Abstract: The present invention relates to a method for treating various waste petroleum into eco-friendly solid so that leaching of oil would not occur. The method of the present invention can treat radioactive waste petroleum as well as various waste petroleum, thereby stabilizing waste petroleum chemically and physically, wherein the method comprises mixing waste petroleum with a sulfuric acid and a nitric acid; adding sodium hydroxide, thereby carrying out a polymerization reaction to produce solid particles; colloidizing a mixture obtained by uniformly stirring the solid particles; adding a diisocyanate compound in reactor, thereby carrying out a subsequent polymerization reaction to obtain a new compound in the form of powder; discharging a generated gas into the atmosphere; and filling the powder into a resin as a filling material and compression molding and reclaiming the filled powder.

Abstract: A method for treating before calcination a nitric aqueous solution comprising at least one radionuclide and ruthenium is provided. The method comprises a step for adding to the solution a compound selected from lignins, lignocelluloses, optionally as salts and mixtures thereof.

Abstract: An adsorbent for radioelement-containing waste composed of the following spherical layered double hydroxide (A) or spherical metal hydroxide (B) is provided. (A) is a nonstoichiometric compound represented by general formula (a) or (b): [M2+1-xM3+x(OH)2]x+[An?x/n·mH2O]x? . . . (a), [Al2Li(OH)6]x+[An?x/n·mH2O]x? . . . (b) wherein 0.1?x?0.4, 0<m, n represents a natural number of 1 to 4, M2+ represents at least one divalent metal, M3+ represents at least one trivalent metal, and An? represents at least one n-valent ion-exchangeable anion, and (B) is a spherical metal hydroxide containing a metal selected from the group consisting of the metal atoms belonging to Group II, Group IV, Group V, Group VI, Group XI, Group XII, and Group XIII of the periodic table, and the group consisting of Mn, Fe, Co, Ni, Pb, and Bi.

Abstract: Method of immobilizing nuclear waste comprising: a mixing step during which nuclear waste is mixed with a mineral composition and water, a drying step during which the mixture obtained in the mixing step is dried so as to form an immobilizing matrix, characterized in that the mineral composition is obtained by a manufacturing process that comprises the steps consisting of: preparation of a base comprising a predetermined quantity of a mineral material synthesized by at least one part of a living structure selected from the vegetable kingdom, animal kingdom and/or microorganisms; and treatment of said base so as to convert it into an inactivated material with a predefined texture.

Abstract: The present invention discloses a method for solidifying and stabilizing waste acid including steps of condensing waste acid containing phosphoric acid to reduce the volume; mixing the condensed waste acid with waste acid containing fluoroboric acid to solidify and stabilize the mixed waste acid. The pH of the mixed acid is adjusted by adding barium hydroxide as a neutralizer. The efficiency of solidifying waste acid can be improved by partially granulating and by adding solidifying agent indirectly. The method of the present invention can prevent intensely exothermic reaction caused by adding solidifying agents. Furthermore, the method of the present invention is controlled in a temperature of 30 to 45° C. to improve the polymerization of the mixed waste acid so that the efficiency of solidification thereof can be also improved.

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: A method for disposing of liquid aqueous laboratory waste is described, which involves solidifying the liquid waste with a suitable isovolumic, space-filling superabsorbent polymer within a disposable impermeable film-type container held in an open top, reusable rigid outer container such as a laboratory beaker, and closing and removing the disposable container containing the solidified waste from the rigid outer container. The waste held in the film-type container can be finally disposed of through incineration or deposit in a suitable landfill.

Abstract: The invention relates to a process for separating, in an aqueous medium, at least one actinide element from one or more lanthanide elements by using at least one molecule which sequesters the said actinide element to be separated and membrane filtration, the said process successively comprising: a) a step of bringing at least one molecule which sequesters the said actinide element in contact with the aqueous medium, the said molecule not being retained in the non-complexed state by the said membrane and being capable of forming a complex with the actinide element to be separated, comprising the said element and at least two of the said sequestering molecules, which complex is capable of being retained by the membrane; b) a step of passing the aqueous medium over the membrane in order to form a permeate on one side, comprising an aqueous effluent depleted of the said actinide element, and a retentate comprising the said complex.

Abstract: Methods and devices for removing inflammable gases produced by radiolysis in a closed chamber containing radioactive matters comprising organic compounds and possibly water, or radioactive matters in the presence of organic compounds and possibly water. Inside the chamber there may be placed a catalyst of at least one reaction for oxidizing the inflammable gases by oxygen contained in the chamber atmosphere, supported by an inert solid support; a catalyst of at least the reaction for oxidizing CO to CO2; possibly an oxygen source; and possibly a hygroscopic microporous inert solid support. Also, chambers for radioactive matters containing such devices.

Abstract: The present invention discloses a method for solidifying and stabilizing waste acid including steps of condensing waste acid containing phosphoric acid to reduce the volume; mixing the condensed waste acid with waste acid containing fluoroboric acid to solidify and stabilize the mixed waste acid. The pH of the mixed acid is adjusted by adding barium hydroxide as a neutralizer. The efficiency of solidifying waste acid can be improved by partially granulating and by adding solidifying agent indirectly. The method of the present invention can prevent intensely exothermic reaction caused by adding solidifying agents. Furthermore, the method of the present invention is controlled in a temperature of 30 to 45° C. to improve the polymerization of the mixed waste acid so that the efficiency of solidification thereof can be also improved.

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: Treatment of a radioactive waste stream is provided by adding sodium hydroxide (NaOH) and/or potassium hydroxide (KOH) together with a rapidly dissolving form of silica, e.g., fumed silica or fly ash. Alternatively, the fumed silica can be first dissolved in a NaOH/KOH solution, which is then combined with the waste solution. Adding a binder that can be a mixture of metakaolin (Al2O3.2SiO2), ground blast furnace slag, fly ash, or other additives. Adding an “enhancer” that can be composed of a group of additives that are used to further enhance the immobilization of heavy metals and key radionuclides such as 99Tc and 129I. An additional step can involve simple mixing of the binder with the activator and enhancer, which can occur in the final waste form container, or in a mixing vessel prior to pumping into the final waste form container, depending on the particular application.

Abstract: A method for conditioning a contaminated ion exchange resin includes mixing the contaminated ion exchange resin with water and at least partly breaking up the contaminated ion exchange resin into water-soluble components or fragments through the use of an oxidizing agent added to the water. A resulting aqueous solution is consolidated with a binder, optionally after concentration by evaporation of water.

Abstract: An apparatus for encapsulating waste material (e.g radioactive sludge from nuclear processing plant) in a container (e.g Nirex box) for long term storage, comprising: a first storage vessel, for holding sludge; a second storage vessel, for holding encapsulation medium (e.g. cement based grout); an inline mixer (e.g. a static inline mixer), coupled for receiving sludge, and coupled to the second storage vessel, and producing, in use, a mixture of the sludge and grout; wherein the inline mixer is arranged for filling the container. Preferably, a dewatering unit (e.g. HydroTrans based), coupled for receiving sludge and outputting dewatered sludge to be mixed by the inline mixer. An encapsulation system comprising the encapsulation apparatus, and corresponding encapsulation methods, are also disclosed.

Abstract: Methods of treating radioactive materials are disclosed. In one aspect, a method may include mixing a radioactive isotope diluted in a filler material with a radioactive material treatment composition to form a resulting material. The radioactive material treatment composition may include mostly salt, and from 0.5 to 15 wt % sorbent. The method may further include mixing the resulting material with one or more inorganic binding agents. Other methods of treating radioactive materials are also disclosed, as well as compositions for treating radioactive materials.

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: The present invention relates to the use of a mixed carbonate of formula AB(CO3)2, in which A and B are different and chosen from alkali metals, alkaline-earth metals and rare earths, for the containment of radioactive carbon. This use may for example involve a process comprising: mixing CO2 having a radioactive carbon to be contained, or a simple carbonate of an alkali, alkaline-earth or rare-earth metal having a radioactive carbon to be contained, with an aqueous solution of a mixture of ACln and BClm or with an aqueous solution of a mixture of A(OH)n, and B(OH)m in order to obtain a precipitate of AB(CO3)2, where n and m are integers sufficient to compensate for the charge of A and B respectively; recovery of the AB(CO3)2 precipitate in powder form; and then pressing and sintering of the powder at a 20 temperature below the decarbonation temperature of the mixed carbonate manufactured in order to obtain sintered pellets of mixed carbonates for the containment of the radioactive carbon.

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: The invention relates to a method and system for the treatment of radioactive wastes produced as a consequence of the operation of nuclear power plants with pressurized water reactors and boron reactivity regulation accompanied by the simultaneous production of environmentally acceptable substances of, for example, borax, calcium-magnesium borates, boron acid and sodium hydroxide solutions.

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: An injection and solidification operation as well as a kneading and solidification operation can be performed by a single facility. A decreased amount of radioactive secondary waste is generated. A solidifying agent paste is prepared by kneading a solidifying agent and additive water. The solidifying agent paste is injected into a solidifying container. The radioactive waste is charged into the solidifying container and kneaded.

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: An injection and solidification operation as well as a kneading and solidification operation can be performed by a single facility. A decreased amount of radioactive secondary waste is generated. A solidifying agent paste is prepared by kneading a solidifying agent and additive water. The solidifying agent paste is injected into a solidifying container. The radioactive waste is charged into the solidifying container and kneaded.

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: This invention relates to a method of strong/tight packaging an object in order to prepare it for shipping. Initially, a substrate is applied to an object to create a smooth surface adapted to receive a sprayable plastic thereon. Then, the sprayable plastic is sprayed onto the substrate and the object to encapsulate the object.

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: 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 process for chemical fixation of radionuclides and radioactive compounds present in soils, solid materials, sludges and liquids. Radionuclides and other radioactive compounds are converted to low-temperature Apatite-Group structural isomorphs (general composition: (AB)5(XO4)3Z), usually phosphatic, that are insoluble, non-leachable, non-zeolitic, and pH stable by contacting with a suspension containing a sulfate, hydroxide, chloride, fluoride and/or silicate source and a phosphate anion. The Apatitic-structure end product is chemically altered from the initial material and reduced in volume and mass. The end product can be void of free liquids and exhibits sufficiently high levels of thermal stability to be effective in the presence of heat generating nuclear reactions. The process occurs at ambient temperature and pressure.

Abstract: A phosphosilicate apatite useful as a confinement matrix for radioactive waste, and having the formula: MtCaxLnyHfwPuz−w(PO4)6−u(SiO4)uF2  (I) wherein: M represents an alkaline metal, Ln represents at least one cation selected from lanthanides, and t, x, y, z, w and u are such that: 0≦t≦1, 8≦x≦10, 0≦y≦1, 0<z≦0.5, 0≦w≦z, and 0<u≦y+2z, and the total number of positive charges provided by the alkaline metal, Ca, Ln, Hf and Pu cations are equal to 20+u is prepared by a sintering-reaction of a mixture of reagent powders, in a neutral or reducing atmosphere, with application of pressure before or during sintering.

Abstract: Method and apparatus for containing, transporting, and storing or disposing of radioactive machinery, including decommissioned nuclear reactor pressure vessels. An improved, economically-produced container is provided which allows easier handling and packaging of machinery within plants where the machinery has been installed, and which provides improved shock absorption and attenuation characteristics, especially when packaging is complete. A reactor pressure vessel or similar item is disconnected from the remainder of the plant and external fittings are trimmed as close to flush with item's exterior as practicable. A storage and containment canister, optionally cut into at least two sections to ease handling and packaging, is placed nearby. The pressure vessel head or any other low-radioactive items are removed, and insulation and other items removed from the outside of the item are placed inside the item's body.

Abstract: The invention relates to the containment of caesium and/or rubidium, in particular 135Cs and 137Cs in an apatitic matrix of formula: MtCaxLny(PO4)6−u(SiO4)uX in which: M represents Cs and/or Rb, Ln represents at least one trivalent cation, X represents at least one anion chosen from among 2F−, S2−, 2Cl−, 2Br−, 2I−, 2OH− and O2−, and t, x, y, and u are such that: 0<t·2.5 2·x·8 1·y·7 0·u·6 x+y+t=10 and the total number of positive charges provided by the cations M, Ca and Ln are equal to (20+u).

Abstract: A device for improving waste vitrification in a disposable canister, the process for using the device, and the process for making the device. The disposable canister, also known as a module, has outer and inner containers with thermal insulation therebetween. The device includes an inner container of graphite having a layer of pyrographite on its external wall. The outer container is typically made of stainless steel. The inner container is heated, typically by induction, to melt the frit and waste. The melted mixture is then cooled to form a vitrified product in the module. The fabrication of the pyrographite coating on the inner container involves heating the container to about 1500 degrees centigrade in a methane atmosphere, then cooling the container to ambient temperature.

Abstract: Solidification of liquid radioactive waste, and other hazardous wastes, is accomplished by the method of the invention by incorporating the waste into a porous glass crystalline molded block. The porous block is first loaded with the liquid waste and then dehydrated and exposed to thermal treatment at 50-1,000° C. The porous glass crystalline molded block consists of glass crystalline hollow microspheres separated from fly ash (cenospheres), resulting from incineration of fossil plant coals. In a preferred embodiment, the porous glass crystalline blocks are formed from perforated cenospheres of grain size −400+50, wherein the selected cenospheres are consolidated into the porous molded block with a binder, such as liquid silicate glass. The porous blocks are then subjected to repeated cycles of saturating with liquid waste, and drying, and after the last cycle the blocks are subjected to calcination to transform the dried salts to more stable oxides.