Abstract: A method for securely and safely storing radiocontaminated waste matter and a container therefor are provided. Radiocontaminated waste matter and PSC are mixed and then retained and stored in a tetragonal cylindrical container tank 1 made of steel sheet, concrete, or PSC-containing concrete, so that the spatial gamma radiation dosage of the environment around the tank 1 becomes about the same as that of an environment or place which receives no fall-out radioactive substances. When a mixture of radiocontaminated waste matter and PSC is ashed, and the ash thus obtained is again mixed with PSC, and then loaded and stored in said container tank, the spatial gamma radiation dosage around said container tank is to be similar to that of an environment or place which receives no fall-out radioactive substances, and simultaneously both 134Cs and 137Cs are decreased, and as a result radiocontaminated waste matter can be securely and safely loaded and stored for a long-period of time.

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

Abstract: 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 determining the distribution of radioactive material within a region is described. A plurality of radiological measurements and associated geometrical measurements taken at a plurality of positions in said region are input to a computer system. A 3D model of structures within said region is defined and the position within the model of each radiation measurement is obtained by using the geometrical measurements. The radiological measurements are then ascribed to a distribution of sources restricted to defined locations in the 3D model. The source distribution is parameterized over the defined source locations, and each parameter is related to the calculated observable radiation field, calculated using a physical model, at each measurement position. The parameters are adjusted to optimize the correspondence between the actual radiological measurements and the calculated observable radiation field, to yield the distribution of radioactive material as defined by the adjusted parameters.

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: 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: The Decommissioning Phase SAFSTOR for Nuclear Power Plants, lasting for 50 to 60 years before dismantling begins, is to allow for natural decay of 60Co, a constituent of reactor steel to roughly 1/1000th of its original content. This can be sped up partly by minimizing its build-up in the Reactor Pressure Vessel and in the Biological Shield, partly by reducing its higher contents in the former by milling/cutting from its inner side. The outer rise of the activity (65, in FIG. 10) in the Reactor Pressure Vessel is caused by the back flow of thermalized neutrons (66) from the Biological Shield (67). An absorber (68), added to the inner liner (69), causes by blocking off this flow of neutrons a steady decline of the Co-distribution with its minimum at the outer side of the wall (70). Then, milling/cutting on the inside removes most of the 60Co exclusively from the inside of the Reactor Pressure Vessel.

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

Abstract: The invention provides a composite from which concrete featuring a sufficiently high heat resistance can be produced, as well as a high-safety sealed concrete cask having no opening (shielding defect) to offer high shielding performance that can prevent corrosion of an internal canister and release of radioactive material to the exterior. A concrete cask of the invention includes a cask body having a bottom but no lid in itself, and a lid which can open and close off a top opening of the cask body. Both the cask body and the lid are made of concrete manufactured by using a composite including Portland cement or blended cement containing Portland cement, which is mixed with water in such a manner that the content of calcium hydroxide falls in a range of 15% to 60% by mass after hardening through hydration reaction. Metallic heat-transfer fins are embedded in the cask body.

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 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: 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: 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: A new method for combining magnesium oxide, MgO, and magnesium dihydrogen phosphate to form an inexpensive compactible ceramic to stabilize very low solubility metal oxides, ashes, swarfs, and other iron or metal-based additives, to create products and waste forms which can be poured or dye cast, and to reinforce and strengthen the ceramics formed by the addition of fibers to the initial ceramic mixture.

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: 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: This invention discloses a method for the co-solidification of low-level radioactive wet wastes of BWR nuclear power plants, including concentrate waste, spent ion exchange resins and sludge wastes etc., with very high volume efficiency. In this invention, for promoting the stability of the solidified waste, sodium sulfate in the concentrate solution is converted to sodium hydroxide and barium sulfate by reacting with barium hydroxide. The conversion product barium sulfate possessing high density and stability is insoluble and used as a fine aggregate material in the solidified waste. Sodium hydroxide is used to stabilize ion exchange resins and to form a highly water-durable solidified waste form with silicates and phosphates in the solidification agent mixture. The solidification agent used in this invention is a formulated powder mixture completely made from inorganic materials. Therefore, there is no aging problem of the solidified waste.

Abstract: A method of encapsulating hazardous waste materials including heavy metals such as arsenic, mercury, nickel and chromium residues, as well as radioactive materials. The method involves adding the hazardous waste material to a settable composition, forming a slurry, and allowing the slurry to set to encapsulate the waste material. The settable composition is a powdered flowable cement composition containing calcium carbonate and a caustic magnesium oxide. Tests conducted on the encapsulated material indicate that virtually none of the hazardous waste material is leached out of the set composition which has a concrete-like appearance.

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: Known phosphate ceramic formulations are improved and the ability to produce iron-based phosphate ceramic systems is enabled by the addition of an oxidizing or reducing step during the acid-base reactions that form the phosphate ceramic products. The additives allow control of the rate of the acid-base reactions and concomitant heat generation. In an alternate embodiment, waste containing metal anions are stabilized in phosphate ceramic products by the addition of a reducing agent to the phosphate ceramic mixture. The reduced metal ions are more stable and/or reactive with the phosphate ions, resulting in the formation of insoluble metal species within the phosphate ceramic matrix, such that the resulting chemically bonded phosphate ceramic product has greater leach resistance.

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: Disclosed is a method of disposal of metallic aluminum-containing solid waste comprising the steps of reacting radioactive solid waste containing metallic aluminum with an alkali solution to generate a hydrogen gas and then mixing the resulting reaction liquid with a solidifying material containing a latent hydraulic material as a main component to solidify the liquid. According to the method of the invention, solidification of aluminum-containing radioactive solid waste in a highly filled state (high concentration) is feasible and a solid having excellent mechanical strength and properties of preventing oozing of radioactive nuclide can be obtained.

Abstract: There are provided a solidifying material used for radioactive waste which comprises a latent hydraulic material, an ultra-fine powder substance and a dispersing agent, and a process for the solidification of radioactive wastes which comprises kneading the above solidifying material with a cure stimulating agent, radioactive waste and, as occasion demands, water, and then curing the mixture by aging. Since the solidifying material of the present invention has excellent properties such as large treating capacity, excellent water resistance and large compressive strength, not only temporary storage but also permanent storage of radioactive wastes can be made.

Abstract: The present invention provides a method for encapsulating and stabilizing radioactive, hazardous and mixed wastes in a modified sulfur cement composition. The waste may be incinerator fly ash or bottom ash including radioactive contaminants, toxic metal salts and other wastes commonly found in refuse. The process may use glass fibers mixed into the composition to improve the tensile strength and a low concentration of anhydrous sodium sulfide to reduce toxic metal solubility. The present invention preferably includes a method for encapsulating radioactive, hazardous and mixed wastes by combining substantially anhydrous wastes, molten modified sulfur cement, preferably glass fibers, as well as anhydrous sodium sulfide or calcium hydroxide or sodium hydroxide in a heated double-planetary orbital mixer. The modified sulfur cement is preheated to about 135.degree..+-.5.degree. C., then the remaining substantially dry components are added and mixed to homogeneity.

Abstract: The present invention provides methods and processes for immobilizing and solidifying harmful heavy metal and radioactive species within a waste material. The processes of the present invention are also particularly advantageous for immobilizing and solidifying nitrate compounds with a waste material. One embodiment of the present invention is a method that can be carried out by admixing the waste material with cement and a complexant compound to form a grout admixture. Preferably, the complexant compound is an iron compound that can form a hydrated iron oxide in the presence of an aqueous solution. This grout admixture is then allowed to cure and solidify. The grout admixture is placed within a suitable containment vessel for final storage and disposal.

Abstract: A method of immobilizing toxic waste material slurries is disclosed wherein aggregates are formed containing the toxic waste and the aggregates admixed with and embedded in a sulfur cement to form a concrete shape.

Abstract: An article, such as a containment system (10), having sides (12) with walls (2) or (24) is made; in one method by using cast, cooled, melted, radioactive metal components where the melted metal has a specific activity over 130 Bq/g; or by providing a contaminated material in the form of a solid, liquid or mixture, and then mixing the contaminated material, to which no more than about 15 weight % of uncontaminated material has been reacted, with a binder, followed by forming the composition into a containment system and then curing it into a mass which contains both contaminated material, and uncontaminated binder acting as a matrix for the contaminated material. This article need not be a containment system but can be a wide variety of objects which are made out of radioactive waste, hazardous waste, and their mixtures.

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: Objective of the present invention is to provide a preferable solidified body ensuring performance of artificial barrier in a solidifying process for incombustible miscellaneous solid waste and other wastes such as calcined ashes by preventing generation of hydrogen gas by a reaction of amphoteric metal included in the waste with solidifying materials.In order to suppress the reaction of the amphoteric metal with the solidifying materials, any one or arbitrary plurality of the following three means are provided in the solidifying materials;(1) A means for forming protection film on surface of the amphoteric metal,(2) A means for enhancing hydration reaction of cement,(3) A means for reducing basicity of the solidifying materials.In accordance with the present invention, a preferable solidified body can be obtained by a simple and easy method which does not necessitate separation and pre-treatment of the waste.

Abstract: The present invention discloses a method for preparing the inorganic hardenable slurry and the use of same in the solidification of wastes. In addition to water, the essential parts of the slurry are inorganic components including borates, cement-base powder and other additives such as magnesium oxide, gypsum and silica. The slurry is low in viscosity before solidification and flows freely,and within approximately 30 minutes after preparation the slurry becomes hardened into a highly solidified substance. The solidification of the slurry is resulted mainly from reaction of borates and the cement-base powder and to obtain a best result the weight of borates must be the same as that of the cement-base powder or even higher than the weight of the latter. The solidification mechanism differs entirely from the simple hydration reaction that brings about solidification of cement.The present invention also teaches a method for solidifying wastes with this hardenable slurry, i.e.

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: A method for decontaminating a radioactively contaminated metallic object. A radioactively contaminated metallic object is placed into a first bath and thus contacted with a non-radioactive, aqueous solution containing formic acid until the formic acid is completely stoichiometrically depleted thereby forming an aqueous, stoichiometrically depleted solution. The metallic object is then placed into a second bath of the same chemical composition. The non-radioactive, aqueous solution of the second bath is also preferably completely stoichiometrically depleted. The concentration of the aqueous solution containing formic acid is preferably about 0.3 Mol/l. These steps are repeated until the residual radioactivity level of the metallic object is beneath a permissible threshold level, such as 0.37 Bq/cm.sup.2. The radioactive metallic oxides and metallic hydroxides are sedimented out, and the sludge is solidified with cement and subsequently decontaminated.