Abstract: A polymer coating is applied to the surface of a phosphate ceramic composite to effectively immobilize soluble salt anions encapsulated within the phosphate ceramic composite. The polymer coating is made from ceramic materials, including at least one inorganic metal compound, that wet and adhere to the surface structure of the phosphate ceramic composite, thereby isolating the soluble salt anions from the environment and ensuring long-term integrity of the phosphate ceramic composite.

Abstract: A process for the consolidation and containment of solid or semisolid hazardous waste, which process comprises closing an end of a circular hollow cylinder, filling the cylinder with the hazardous waste, and then cold working the cylinder to reduce its diameter while simultaneously compacting the waste. The open end of the cylinder can be sealed prior to or after the cold working process. The preferred method of cold working is to draw the sealed cylinder containing the hazardous waste through a plurality of dies to simultaneously reduce the diameter of the tube while compacting the waste. This process provides a quick continuous process for consolidating hazardous waste, including radioactive waste.

Abstract: In a method and an apparatus for dewatering and containing radioactive, aqueous waste (44), the latter is introduced into a filtration container (12) and is ultimately disposed of in a disposable container structure (12, 48, 50), which comprises the filtration container (12) holding the dewatered waste (44), as well as an outer container (50) enclosing the filtration container (12). The filtration container is an inner sack (12) having a bottom (13) which is provided with a straining cloth and through which essentially all the dewatering is carried out. After the dewatering operation has been completed, the inner sack (12) is sealed and placed in the outer container (50) in order to be ultimately disposed of. For purposes of cleaning, the filtration water (17) may be recirculated through the waste during the dewatering operation.

Abstract: The invention relates to a process for the conditioning of radioactive waste using silicated apatites as the confinement matrix.This process consists of incorporating the waste (1) into a phosphosilicated apatite-based confinement matrix (3) e.g. complying with the formula:M.sub.t Ca.sub.x Ln.sub.y A.sub.z (PO.sub.4).sub.u (SiO.sub.4).sub.6-u X (I)in which M is an alkali metal, Ln a rare earth, A an actinide, X is S.sup.2-, 2F.sup.-, 2Cl.sup.-, 2Br.sup.-, 2I.sup.- or 2OH.sup.- and u is between 0 and 6.

Abstract: The present invention relates to a method for treating radioactive laundry waste water generated from a nuclear power plant, nuclear fuel reprocessing plant, or radioactive nuclides handling facilities, and provides a method for treating the waste water safely and for reducing the volume of generated radioactive waste to a minimum.Radioactive laundry waste water containing a detergent of which major contents are a nonionic surface active agent and inorganic builders is concentrated by an evaporating concentrator, the concentrated waste water is dried and pulverized to dry powder by a rotary centrifugal thin film dryer, and the dry powder is incinerated. By using the above detergent, foaming at the concentration can be reduced, and the concentrated waste water can be easily dried and pulverized. Further, the dried powder can be incinerated stably and safely without influencing undesirable effect on the body of the incinerator.

Abstract: A method of treating a waste material, such as a radioactive waste material, by introducing the material into a high temperature reaction vessel such as an electric arc furnace. By controlling the reducing potential of the oxide phase formed in the reaction vessel, the concentration of the waste material in the oxide phase can be controlled. In addition, the morphology of the oxide phase can be controlled so that once the concentration of the waste material in the oxide phase is adjusted as desired, the waste material can be sequestered within the crystal lattice of the cooled oxide phase for subsequent disposal.

Abstract: A safely encapsulated hazardous waste product free of unwanted leaching of contaminated metal constitutents and process therefor wherein a quantity of hazardous waste and a quantity of shale material characterized by having an aluminosilicate content containing at least 9.35% by weight of Fe.sub.2 O.sub.3, wherein the shale is reduced to a fine mesh before mixing, and defines a brick-like form as fired at about 1,150.degree. C. for a period of time to change the shale material into a semi-molten state and thereafter cooled to safely encapsulate the waste within the shale as a unit-handled product. The product of the invention satisfies and exceeds RCRA and LDR (EPA) requirements as shown by Toxicity Characteristic Leaching Procedure (TCLP) results.

Abstract: A method for the in-situ vitrification of waste materials in a disposable can that includes an inner container and an outer container is disclosed. The method includes the steps of adding frit and waste materials to the inner container, removing any excess water, heating the inner container such that the frit and waste materials melt and vitrify after cooling, while maintaining the outer container at a significantly lower temperature than the inner container. The disposable can is then cooled to ambient temperatures and stored. A device for the in-situ vitrification of waste material in a disposable can is also disclosed.

Abstract: Disclosed is a method for converting spent nuclear fuel and surplus plutonium into a vitreous ceramic final waste form wherein spent nuclear fuel is bound in a crystalline matrix which is in turn bound within glass.

Abstract: This specification presents methods of containing/absorbing alpha, beta, gamma, X-ray, and neutron radiation using Fullerene molecules employing the resonant relativistic absorption phenomena based on a frequency dependant Doppler effect. Radioactive atoms, ions, and molecules (X.sub.rad) are encapsulated in a variety of Fullerene molecules (C.sub.60, onions, nanotubes, and capsules). The resulting radioactive material-holding Fullerene complexes (X.sub.rad @C.sub.n) will either decrease the intensity of escaping internally-generated radioactive emissions or, in certain optimal cases, absorb and/or contain all such internally-generated radioactive emissions which occur as the radioactive material decays.

Abstract: A process for preparing radioactive and other hazardous liquid wastes for treatment by the method of vitrification or melting is provided for.

Abstract: A process for treating a halogen-containing waste material. The process provides a bath of molten glass containing a sacrificial metal oxide capable of reacting with a halogen in the waste material. The sacrificial metal oxide is present in the molten glass in at least a stoichiometric amount with respect to the halogen in the waste material. The waste material is introduced into the bath of molten glass to cause a reaction between the halogen in the waste material and the sacrificial metal oxide to yield a metal halide. The metal halide is a gas at the temperature of the molten glass. The gaseous metal halide is separated from the molten glass and contacted with an aqueous scrubber solution of an alkali metal hydroxide to yield a metal hydroxide or metal oxide-containing precipitate and a soluble alkali metal halide. The precipitate is then separated from the aqueous scrubber solution. The molten glass containing the treated waste material is removed from the bath as a waste glass.

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: A processing system for radioactive waste is composed of an adjusting tank having a sampling port, a solidification processing system, and a package inspection apparatus, and a package, of which inventory per a package has been exactly grasped, is prepared by solidification of the waste with the processing system after determining radioactivity of the waste by measurement before the solidifying process.In accordance with the present invention, data on radioactivity before and after preparation of package of waste become clear, and management of each package at transportation and intermediate storage of the packages is facilitated.

Abstract: The present invention relates to a method of solidifying radioactive waste with cement, comprising forming a mixture comprising water, a hydrophilic material and cement substantially non-shrinkable or expansible with respect to volume change upon hardening, mixing said mixture with the radioactive waste, followed by hardening to form a solid body.The present invention enables the formation of a compact solid body having voids, such as capillary voids, of reduced volume, which makes it possible to reduce the leaching rate of hazardous materials. Further, since no shrinkage accompanies hardening, no tensile stress occurs in the cement surrounding minute waste particles within the hardened material, thereby enabling a decrease in the strength of the solid body to be minimized. This in turn enables an increase in the amount of packing of waste. Prior addition of a hydrophilic material enables the cement fluidity before hardening to be maintained even after complete absorption of water by a water absorptive waste.

Abstract: The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide.

Abstract: A process for treating radioactive waste in the form of contaminated powdery ionic exchange resin to make it suitable for final storage by reducing its volume as much as possible, includes mechanically dewatering the ion exchange resin. The dewatered ion exchange resin is mixed with a calcium compound. The mixture is dried at temperatures of up to 120.degree. C. and preferably about 50.degree. C. to 60.degree. C., and at a pressure of from 120 to 200 hPa, until a residual moisture content of less than 10% of the mass of the mixture is reached. The dry mixture is thermally treated at a pressure below atmospheric pressure by heating up to a temperature of from at least 120.degree. C. to at most 190.degree. C. The ion exchange resin thereby loses its water absorption and swelling capability. Ion exchange resins treated in this manner are processed with cement or bitumen to form blocks that are suitable for final storage.

Abstract: Processes and devices are disclosed for converting an aqueous waste solution, such as a radioactive waste stream, into a highly concentrated sludge having substantially little or no water. In turn, the sludge may be converted into a solid polymeric form suitable for storage. The processes utilize a carrier liquid which has a boiling point higher than that of water, and which facilitates the transfer of heat to the water to vaporize the water into steam. A subsequent separation of the dewatered solution yields the highly concentrated sludge.

Abstract: A method for treating materials such as wastes for solidification to form a solid, substantially nonleachable product. Addition of reactive silica rather than ordinary silica to the material when bringing the initial molar ratio of its silica constituent to a desired ratio within a preselected range increases the solubility and retention of the materials in the solidified matrix. Materials include hazardous, radioactive, mixed, and heavy metal species. Amounts of other constituents of the material, in addition to its silica content are also added so that the molar ratio of each of these constituents is within the preselected ranges for the final solidified product. The mixture is then solidified by cement solidification or vitrification.

Abstract: A method of containing hazardous and toxic wastes includes the steps of irporating the dried waste, in a salt form, in melted polymer, such as asphalt, and forming the waste salt and asphalt blend into aggregate pellets. The pellets are coated with a powdered coating material that is compatible with a portland cement-based mortar or other cementitious material which is used. The coated particles are mixed with mortar to form a polymer-aggregate concrete and cast into wasteforms for storage or burial. If it is desirable to produce a waste form with a continuous layer of mortar on the exterior of the concrete monolith the mold can be placed on a turntable and spun, or otherwise exposed to a centrifugal force to force the mortar to the outside of the mold. Centrifugal separation is possible because the polymer-waste mixture typically has a specific gravity near 1.5 while that of the cementitious mixture is typically greater than 2.0.

Abstract: A method and composition for stabilizing and isolating hazardous, radioactive or mixed waste materials of particulate and solid types, which comprises providing a non-toxic one component aqueous mixture of an acrylic polymer containing a thixotropic agent, a vinyl acetate-ethylene copolymer containing a thixotropic agent, or a vinyl chloride copolymer latex containing a thixotropic agent, and applying the mixture over surfaces of hazardous material in an amount sufficient to form a flexible impermeable coating or foam. The mixture may be applied by spraying to form a coating having a thickness of about 0.5 to about 5.0 centimeters when dry. The coated waste material may then be disposed of in conventional manner.

Abstract: A leach-resistant vitrified composition, particularly suitable for long term encapsulation and disposal of nuclear wastes, comprises a high iron, high potassium aluminosilicate, fired at a temperature of at least 1850.degree. F. for at least 12 h.

Abstract: Method and apparatus for the incineration and vitrification of radioactive waste materials. Waste materials are fed into a melter containing molten glass wherein the waste is incinerated and vitrified into the glass matrix. Gases produced are combusted in an extended plenum area preferably designed to provide a residence time of at least 3 seconds. A portion of the sulfur compounds in the combustion gases are incorporated into a solid matrix of gypsum by contacting the sulfur compounds with an aqueous solution and calcium hydroxide.