Abstract: A method for processing a radioactive liquid waste containing boron of the present invention includes: a molar ratio control step of adding an alkali metal or an alkali metal compound to a radioactive liquid waste containing boron to control an alkali metal/boron molar ratio in the radioactive liquid waste to be 0.8 or more; a drying step of drying the radioactive liquid waste having the controlled molar-ratio using a dryer to form a powdered waste; a dissolving step of mixing the powdered waste with kneading water to prepare a solution; and a kneading step of adding a hydraulic inorganic solidifying material to the solution, and kneading the hydraulic inorganic solidifying material and the solution for solidification.

Abstract: Methods for converting toxic waste, including nuclear waste, to quasi-natural or artificial feldspar minerals are disclosed. The disclosed methods may include converting, chemically binding, sequestering and incorporating the toxic waste into quasi-natural or artificial Feldspar minerals. The quasi-natural or artificial feldspar minerals may be configured to match naturally occurring materials at a selected disposal site. Methods for the immediate, long term, quasi-permanent disposal or storage of quasi natural or artificial feldspar materials are also disclosed.

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 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: A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

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 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 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: 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: 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 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 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 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: A method for remediating non-homogeneous radioactive waste to significantly reduce the waste mass/volume and to convert such waste to products that meet federal regulatory compliance standards is disclosed. High level waste (HLW) stored in underground tanks is typically a multi component mixture. After removal of the waste from the tanks or other storage areas the waste is isolated in a thermal desorption-type reaction vessel where the waste is pyrolized at pre-determined and carefully controlled temperatures, pressures, and atmospheres. This process eliminates organics, volatile metals, moisture and other low boiling temperature/high vapor pressure components and converts non-volatile waste to more stable metal oxides. Off-gas treatment systems scrub, treat and dispose of all off-gas components. A thermal desorption-type apparatus especially well suited for performing the method of the present invention is also disclosed.

Abstract: The present invention provides a method and an apparatus for treating boron-containing waste, which prevent clogging of a waste liquid line, achieve an increase in feeding of a waste liquid to a dryer, make in-drum mixing feasible, and increase the strength of a cement solidification product. A method for treating a boron-containing waste liquid comprises the step of adding an alkali metal element compound and an alkali earth metal element compound to the boron-containing waste liquid, and the step of then drying and powdering the boron-containing waste liquid, wherein the temperature of the boron-containing waste liquid during a period from addition of either the alkali metal element compound or the alkali earth metal element compound until the drying and powdering is maintained at not lower than the precipitation temperature of a compound containing boron and an alkali metal and a compound containing boron and an alkali earth metal.

Abstract: The aim of the present invention is a process for preparing a mineral matrix by melting, which is implemented according to the method of direct cold crucible induction melting. The method employs a step for initiating the melting. During the initiation step, a conductive mineral load is generated, by introduction of constituent elements of the matrix, into a bath of a conductive liquid, brought by induction to a suitable temperature, in the cold crucible. The intervening conductive liquid has the property of being a liquid and an electric conductor at a temperature between &thgr;1 of less than 600° C., advantageously of between 100 and 500° C., and a temperature &thgr;2 at least equal to the temperature at which the constituent elements of the matrix melt to generate the matrix. The conductive liquid is advantageously a molten sodium hydroxide bath. The preparation of such a glass matrix is advantageously implemented within the context of a method of vitrifying radioactive waste materials.

Abstract: A radiation protector treating method including the steps of pulverizing a used radiation protector to obtain a pulverized radiation protector material and mixing the pulverized radiation protector material with boron powder and bismuth powder, and casting the pulverized radiation protector material mixed with the powders into an electric melting furnace and also casting silicon powder, lead oxide powder and carbon powder into the electric melting furnace. Radioactive contamination is prevented by using a radiation attenuating and absorbing material for each of various radioactive substances attached to the radiation protector.

Abstract: The present invention relates to vitrification of ion exchange resins that have become loaded with hazardous or radioactive wastes, in a way that produces a homogenous and durable waste form and reduces the disposal volume of the resin. The methods of the present invention involve directly adding borosilicate glass formers and an oxidizer to the ion exchange resin and heating the mixture at sufficient temperature to produce homogeneous glass.

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: 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: A pumpable ceramic composition is provided comprising an inorganic oxide, potassium phosphate, and an oxide coating material. Also provided is a method for preparing pumpable ceramic-based waste forms comprising selecting inorganic oxides based on solubility, surface area and morphology criteria; mixing the selected oxides with phosphate solution and waste to form a first mixture; combining an additive to the first mixture to create a second mixture; adding water to the second mixture to create a reactive mixture; homogenizing the reactive mixture; and allowing the reactive mixture to cure.

Abstract: A unique method for solidification of solutions containing boric acid and/or borates is disclosed in this invention. The boron species in the solutions are polymerized to form polyborates, and the solutions are then solidified by mixing with solidification agents which are prepared completely from inorganic materials. Therefore, the solid form produced by this method has no aging problem. The boron species in the solution are not merely wastes to be encapsulated or embedded, they take part in the solidification reaction and share a major portion of total reactants. Thus, the total volume of solid forms produced in this invention is less than 1/10 of that produced in conventional cementation.

Abstract: A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes.

Abstract: A process for determining one or more leachate concentrations of one or more components of a glass composition in an aqueous solution of the glass composition by identifying the components of the glass composition, including associated oxides, determining a preliminary glass dissolution estimator, .DELTA.G.sub.p, based upon the free energies of hydration for the component reactant species, determining an accelerated glass dissolution function, .DELTA.G.sub.a, based upon the free energy associated with weak acid dissociation, .DELTA.G.sub.a.sup.WA, and accelerated matrix dissolution at high pH, .DELTA.G.sub.a.sup.SB associated with solution strong base formation, and determining a final hydration free energy, .DELTA.G.sub.f. This final hydration free energy is then used to determine leachate concentrations for elements of interest using a regression analysis and the formula log.sub.10 (N C.sub.i (g/L))=a.sub.i +b.sub.i .DELTA.G.sub.f.

Abstract: This invention provides a process wherein hazardous or radioactive wastes in the form of liquids, slurries, or finely divided solids are mixed with finely divided glassformers (silica, alumina, soda, etc.) and injected directly into the plume of a non-transferred arc plasma torch. The extremely high temperatures and heat transfer rates makes it possible to convert the waste-glassformer mixture into a fully vitrified molten glass product in a matter of milliseconds. The molten product may then be collected in a crucible for casting into final wasteform geometry, quenching in water, or further holding time to improve homogeneity and eliminate bubbles.

Abstract: A method for immobilizing waste chloride salts containing radionuclides and hazardous nuclear material for permanent disposal starting with a substantially dry zeolite and sufficient glass to form leach resistant sodalite with occluded radionuclides and hazardous nuclear material. The zeolite and glass are heated to a temperature up to about 1000.degree. K. to convert the zeolite to sodalite and thereafter maintained at a pressure and temperature sufficient to form a sodalite product near theoretical density. Pressure is used on the formed sodalite to produce the required density.

Abstract: A method of vitrifying a high-level radioactive liquid waste comprising removing a precipitate composed mainly of Mo and Zr from the high-level liquid waste, mixing the resulting high-level liquid waste with a raw glass material having a chemical composition wherein the B.sub.2 O.sub.3 /SiO.sub.2, ZnO/Li.sub.2 O and Al.sub.2 O.sub.3 /Li.sub.2 O ratios are at least 0.41, at least 1.00 and at least 2.58, respectively, and melt-solidifying the mixture to thereby form a vitrified waste. By using such a raw glass material, there can be obtained a vitrifled waste having the waste content of about 45% by oxide weight in which the same leaching rate as that of a conventional vitrified waste having the waste content of 25% by oxide weight is ensured without suffering from yellow phase separation.

Abstract: The present invention relates to a process for nuclear waste disposal. In it, a glass forming mixture including an aqueous solution of one or more metal alkoxides, alcohol, and solubilized, low level radioactive waste having a pH effective to hydrolyze the one or more metal alkoxides is formed. The one or more metal alkoxides in the glass forming mixture are converted to a network of corresponding one or more metal oxides. A gel is then formed from the glass forming mixture containing the network of one or more metal oxides. The gel is dried and sintered under conditions effective to form a densified glass.