Abstract: The current methods of removing the surface-deposited layer contaminated by radioisotopes has the disadvantage of rigorous applicability conditions in such terms as the substance to be treated, the area and the shape, so the scope of their applicability is quite limited; in addition, 100% decontamination is not guaranteed and in spite of more than 50% that can be decontaminated, a greater part of the deposited radioisotopes remain unremoved, limiting the decontamination performance of the methods. Parts, structural components, etc. that have their surfaces contaminated by radioisotopes in the accelerators, nuclear reactors, RI product manufacturing factories, nuclear fuel factories, nuclear fuel reprocessing factories, etc.

Abstract: A chemical decontamination apparatus of the present invention chemically dissolves radioactive substance-containing oxide films formed or adhering on the surface of a decontamination object by using ozone water to conduct decontamination. The chemical decontamination apparatus includes an ozone generating unit for generating ozone gas, an ozone supplying device for supplying the generated ozone gas to an ozone supplying unit in water, and a sintered metal element 37 which is disposed in the ozone supplying unit and to which ozone gas is supplied from the ozone supplying device. The ozone gas supplied to a sintered metal element interior from the ozone supplying device is allowed to flow out of the element into water so as to generate ozone water.

Abstract: A method of decontaminating porous surfaces contaminated with water soluble radionuclides by contacting the contaminated porous surfaces with an ionic solution capable of solubilizing radionuclides present in the porous surfaces followed by contacting the solubilized radionuclides with a gel containing a radionuclide chelator to bind the radionuclides to the gel, and physically removing the gel from the porous surfaces. A dry mix is also disclosed of a cross-linked ionic polymer salt, a linear ionic polymer salt, a radionuclide chelator, and a gel formation controller present in the range of from 0% to about 40% by weight of the dry mix, wherein the ionic polymer salts are granular and the non cross-linked ionic polymer salt is present as a minor constituent.

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: Disclosed herein is provided an evaporator/calciner in which hazardous materials, such as radioactive liquids, are converted into chemically stable, solid forms by evaporating, drying and calcination within a single vessel, that can then be sealed and used for long term storage.

Abstract: A method and apparatus for reducing the radioactivity of a particle is disclosed. The method comprises the steps of: accelerating one or more first particle(s) comprising one or more neutron(s), proton(s) and electron(s) to a first velocity; colliding the accelerated particles(s) with one or more second particles in a collision zone located within a housing causing the first particle(s) and second particle(s) to form one or more collision mass(es) comprising alpha particles and electrons or/and protons and electrons, and in which substantially all neutrons of the first or second particles are converted into alpha particles or/and protons and electrons as a result of the collision; controlling the position of the collision mass(es) with electric or/and magnetic fields; and exhausting the collision mass from the housing wherein the collision mass comprises substantially only alpha particles or/and protons and electrons.

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: Briefly described, compositions, materials including the compositions, methods of using the compositions, and methods of degrading contaminants, are described herein. The composition can include a polyoxometalate/ cationic silica material. In addition, the compositions can be made of a polyoxometalate/cationic silica material, a copper (II) salt having a weakly bound anion, and a nitrate salts. Further, the compositions can be made of a polyoxometalate/cationic silica material, a copper (II) salt having a weakly bound anion, a compound selected from tetraethylammonium (TEA) nitrate, tetra-n-butylammonium (TBA) nitrate, and combinations thereof.

Abstract: A system for chemically decontaminating radioactive material.

Abstract: A system for filling a container with hazardous waste includes a primary confinement chamber that houses a lid handling mechanism and a filling head. The lid handling mechanism may be used to remove and/or recouple the lid to the container as part of the process of filling the container in such a way to ensure the exterior of the container is not contaminated by the hazardous waste. The filling head may be configured to add the hazardous waste to the container, mix the contents of the container, and/or vent air from the container. The system may include additional mechanisms in the primary confinement chamber to add dry cementitious material to the container, add premixed wet cementitious materials to the container, add cementitious material to seal off the top of the lid, and/or measure the level and test whether the contents of the container meet quality assurance requirements. The hazardous waste may include radioactive hazardous waste.

Abstract: Tropospheric volume elements enriched with vital elements and/or protective substances as well as procedures for their production and application. The term “vital elements” applies to all matter supporting the development of life within the earth's biosphere and the term “protective substances” means all those substances which contribute directly or indirectly to the prevention of harmful effects on the earth's biosphere and in particular on man. Tropospheric volume elements in the form of clouds which contain contaminants and which can escape from industrial facilities due to damage or malfunction are enriched with protective substances which prevent the organism from taking in radioactive elements and minimize the extent of the area affected by the clouds and possess additional warning and identification properties.

Abstract: An apparatus for disposal of low-level surface radioactive pollutants has a container body which center has accommodating space with opening at its top. Around the accommodating space there is a layer of barrier space for holding water. At the periphery of the barrier space there is a non-barrier hollow channel for the accommodating space to connect outside. A rotating disk is located at the bottom of the accommodating space in the container body for holding and turning the low-level radioactive waste to be treated and also facilitating the worker to pass through the non-barrier hollow channel for cleaning task.

Abstract: Methods of preparing a proteoliposome comprise the step of contacting a liposome with an effective portion of RalBP1 to create a proteoliposome. RalBP1 is effective for the protection and treatment of mammals and the environment against the accumulation of toxic compounds and prevents accumulation of one or more toxic compounds, reduces the concentration of toxic compounds, and protects against further contamination with one or more toxic compounds.

Abstract: A molten metal reactor (10) quickly entrains a feed material in the molten reactant metal (16) and provides the necessary contact between the molten reactant metal and the feed material to effect the desired chemical reduction of the feed material. The reactor (10) includes a unique feed structure (24) adapted to quickly entrain the feed material into the molten reactant metal (16) and then transfer the molten reactant metal, feed material, and initial reaction products into a treatment chamber (12). A majority of the desired reactions occur in the treatment chamber (12). Reaction products and unspent reactant metal are directed from the treatment chamber (12) to an output chamber (14) where reaction products are removed from the reactor. Unspent reactant metal (16) is then transferred to a heating chamber (15) where it is reheated for recycling through the system.

Abstract: An improved radioactive waste (RW) processing is usable, e.g., for reducing radioactivity of wastes produced by nuclear plants. The processing comprises an entry and a base stages. The entry stage is exemplified either by a “standard” option or an “own” option encompassing the grinding RW and obtaining a pulp. Both the options provide technological mixture for the base processing, which encompasses intermingling the mixture with predetermined substances, kneading and filtering the mixture, removing wastes, processing the mixture in either a regular or in a special dissolver, or in both. The resulted product is subjected to rotation in a special centrifuge utilizing the Searl effect that allows further reducing the RW radioactivity. The so treated RW is then packed into specially constructed glass containers, or compressed and covered by special glass-mass. Composition and processing instructions for the substances and glass-mass, and the constructions of the special dissolver and centrifuge are disclosed.

Abstract: A method and apparatus for removing uranium (IV) and uranium (VI) from sands and soils. The method and device assays a volume of soil and determines a presence and position of uranium enrichment. A concentration of uranium enrichment is determined, and compared to a threshold concentration. The volume of soil is processed and transported to a coarse screen, deck screen and classified in a classifier. In the classifier, the volume of soil is separated into a washed and fines fraction, assayed and transported to a uranium recovery facility, then transported to a silicon removal process and subsequent ammonia and metals removal processes.

Abstract: A system and method for storing high level waste. In one aspect, the invention is a system comprising: an inner shell forming a cavity for receiving high level waste, the cavity having a top and a bottom; an outer shell surrounding the inner shell so as to form a space between the inner shell and the outer shell; at least one opening in the inner shell at or near the bottom of the cavity, the at least one opening forming a passageway from the space into the cavity; a lid positioned atop the inner and outer shells, the lid having at least one inlet vent forming a passageway from an ambient atmosphere to the space and at least one outlet vent forming a passageway from the cavity to the ambient atmosphere. In another aspect, the invention is a method of using the system to store high level waste. In yet another aspect, the invention does not require the aforementioned lid but further comprises a floor plate wherein the inner and outer shells rest atop and are connected to the floor plate.

Abstract: The present invention relates to a method and a device for trapping ruthenium present in a gaseous effluent. The trapping method of the present invention comprises the use of an aqueous solution or slurry of an alkylene glycol polymer or an alkylene glycol copolymer, in which the alkylene(s) has (have) from 2 to 6 carbon atoms, in order to trap ruthenium present in a gaseous effluent. The ruthenium-trapping cartridge of the present invention has a surface (S) on which an alkylene glycol polymer or an alkylene glycol copolymer is placed, in which the alkylene(s) has (have) from 2 to 6 carbon atoms. The present invention makes it possible in particular both to favour the trapping and the chemical operation of reducing ruthenium oxide RuO4.

Abstract: The invention relates to a process for the preparation of a product based on a phosphate of at least one element M(IV), for example of thorium and/or of actinide(IV)(s). This process comprises the following stages: a) mixing a solution of thorium(IV) and/or of at least one actinide(IV) with a phosphoric acid solution in amounts such that the molar ratio PO 4 M ? ? ( IV ) ?is from 1.4 to 2, b) heating the mixture of the solutions in a closed container at a temperature of 50 to 250° C. in order to precipitate a product comprising a phosphate of at least one element M chosen from thorium(IV) and actinide(IV)s having a P/M molar ratio of 1.5, and c) separating the precipitated product from the solution. The precipitate can be converted to phosphate/diphosphate of thorium and of actinide(s). The process also applies to the separation of uranyl ions from other cations.

Abstract: Uses of cucurbituril derivatives are disclosed.

Abstract: An exemplary embodiment is a method and system for performing an automated independent technical review. The method includes receiving an assay result of a radioactive waste container, determining whether the assay result is within a predetermined parameter, determining whether a review is required if the assay result is not within the predetermined parameter and rejecting the assay result if the review is not required and the assay result is not within the predetermined parameter.

Abstract: A method for chemically decontaminating radioactive material. The method includes reducing-dissolving step for setting surface of radioactive material in contact with reducing decontamination liquid including mono-carboxylic acid and di-carboxylic acid as dissolvent; and oxidizing-dissolving step for setting the surface of the radioactive material in contact with oxidizing decontamination liquid including oxidizer. The method may include repeated pairs of steps, each pair including the reducing-dissolving step and the oxidizing-dissolving step. The mono-carboxylic acid may include formic acid, and the di-carboxylic acid includes oxalic acid. The oxidizer may be ozone, permanganic acid or permanganate.

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: In a chemical decontamination apparatus, a catalyst decomposition column in an upstream side of an ion exchange resin column and a hydrogen peroxide injection apparatus in a further upstream side, reduce the amount of waste products caused by a chemical decontaminating agent where a mixed decontaminating agent for a composition trapped in a cation resin column and for a composition trapped in an anion exchange resin are used for the chemical decontaminating agent, selectively decompose the composition trapped in the cation resin column in an inlet side of a cleaning apparatus when radioactive nuclides in the decontaminating agent are cleansed using the cation resin column during decontamination, and decompose both compositions after completion of the decontamination. The chemical decontamination thus selectively decomposes the chemical decontaminating agent, which is a component of the load to the cation resin column.

Abstract: A radioactive substance decontamination method and apparatus which decontaminates a metal member contaminated by radioactive substance in a short period of time. This apparatus has (1) multiple reducing decontamination tanks having different radiation control values; (2) a carrier for immersing the metal member into the multiple reducing decontamination tanks and a washing tank; (3) a tube for transferring into the second reducing decontamination tank the reducing decontamination agent in the first reducing decontamination tank; (4) a reducing agent decomposer for decomposing a component contained in the reducing decontamination agent of the reducing decontamination tank where the radiation control value is the highest out of the reducing decontamination tanks connected by the tube; and (5) a washing tank for washing the reducing decontamination agent deposited on the decontaminated metal member.

Abstract: A method and system for handling hazardous materials contained in a vial includes an isolation enclosure having an opening selectively sealable about the vial, a bag body portion, and a cap portion. A latching extraction element is attached to the cap portion and has a preceding engaging member to secure the vial to the isolation enclosure, an extraction member to be inserted into the vial and remove material therefrom, and a primary engaging member to secure the vial to the extraction member. A valve is mounted outside the isolation enclosure and controls the flow of fluid from the vial. An adaptor having a reseal member permits flow when coupled to the valve and restricts flow when uncoupled from the valve. Once uncoupled, the adaptor is removably associated with a second valve located remotely from the isolation enclosure, allowing fluid to pass into the second valve.

Abstract: A package and process of using the package for disposal of radioactive cesium and strontium waste capsules. The package comprises a standard Hanford vitrified high-level waste canister as an outer container, which is approximately filled with three components: the first is a means for containing waste capsules, having a composite density less than about 3.5 grams per cubic centimeter and a melting temperature above that expected within the disposal package; the second is a means for limiting relative movement of the capsules; and, the third is a means for conducting heat away from the capsules. The package includes lids for closing the disposal package. In the method of the invention, the capsules are loaded into position within the means for containing waste capsules, encased in thermally conducting material, and then lids are added to close the package.

Abstract: Radioactive material can be processed by an apparatus that includes at least a cylindrical outer shell electrode, an inner electrode, and a plurality of prism-shaped ferromagnetic elements positioned between the outer and inner electrodes. The prism-shaped ferromagnetic elements are positioned around the inner circumference of the metal cylinder. The inner electrode component is located within the metal cylinder and is configured to cover the inwardly-pointing portions of the prism-shaped ferromagnetic elements. Radioactive material in a container is placed into the apparatus, and an AC voltage excitation signal is applied to the electrodes of the apparatus during treatment of the material. The frequency of the excitation signal is selected according to the frequency of structurization or the frequency of destructurization of the ferromagnetic material. The process can be monitored and controlled with the use of alpha, beta, and gamma radiation intensity measuring instruments.

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: A lifting lug for a nuclear-waste container has a tubular body having an inner end formed with a flange adapted to be bolted to the container and a transverse wall set in and tightly fitted to a smaller-diameter outer end of the tubular body. The wall is a separate piece from the body and is press-fitted to the tubular body. The wall is mounted at a load point of the lifting lug. In addition the wall is formed of a disk and/or a ring, typically as one unitary piece with the disk and ring spaced from each other.

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 method and apparatus to hermetically seal openings in an explosion suppression chamber. The present invention utilizes an air lock device to ensure that, in the event toxins are released from the primary explosion chamber opening closing means, the toxins are properly handled and are not inadvertently released into the atmosphere. The preferred embodiment of the present invention uses negative pressure to vacuum the entrained air within the airlock cavity subsequent to an explosion. An alternate embodiment of the present invention uses positive pressure within the air lock cavity to force air through integral filter membranes.

Abstract: Ozone gas having a high ozone concentration is generated by a solid electrolyte electrolytic process. An ozone solution is prepared by injecting the ozone gas into an acidic solution of pH 6 or below. The ozone solution heated at a temperature in the range of 50° to 90° C. is supplied to a contaminated object to oxidize and dissolve a chromium oxide film by an oxidizing dissolving process. The ozone solution used in the oxidizing dissolving process is irradiated with ultraviolet rays to decompose ozone contained in the ozone solution, the ozone solution is passed through an ion-exchange resin to remove ions contained in the ozone solution. An oxalic acid solution is supplied to the contaminated object to dissolve an iron oxide film by a reductive dissolving process.

Abstract: In order to vent packages of transuranic waste, such as plutonium 238, the waste is contained in at least one, preferably several plastic bags that form liners in a steel drum 60. The package is irradiated with gamma radiation which passes through the drum and accelerates cross-linking and embrittlement of the bags, rupturing the bags and releasing explosive gas, such as hydrogen or volatile organic compounds, from the bags into the drums. Explosive gas is then vented from the drums, providing packages containing transuranic waste which are transportable under Department of Transportation regulations.

Abstract: A method for removing a wide variety of radioactive contaminants from a contaminated surface sufficient for the surface to be classified as a low-level waste or as free release. Contaminated surfaces may be classified as Class C, Class B, Class A, or high-level radioactive wastes prior to treatment. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The aqueous solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, an acid may be added to the aqueous solution to aid in the contaminant removal process. However, typical metals surfaces may often be treated effectively without the use of concentrated acids or with dilute concentrations of such acids. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a variety of previously contaminated surfaces.

Abstract: A process wherein the use of a degradation-enhancing reactant or a precursor thereof, e.g., an oxidizing agent, can effectively increase the rate of reaction for the degradation of polymers in an aqueous environment and thereby facilitate the disposal of same. For example, in one aspect, the process comprises contacting a solubilized polymer and a degradation-enhancing reactant, within an aqueous environment, and subsequently reacting the polymer under conditions that are effective to provide at least one environmentally benign degradation product from the polymer.

Abstract: A process for the treatment of radioactive graphite which includes the following steps: (i) reacting the radioactive graphite at a temperature in the range of from 250° C. to 900° C. with superheated steam or gases containing water vapor to form hydrogen and carbon monoxide; (ii) reacting the hydrogen and carbon monoxide from step (i) to form water and carbon dioxide; and (iii) reacting the carbon dioxide of step (ii) with metal oxides to for carbonate salts. The process enables radioactive graphite, such as graphite moderator, to be treated either in-situ or externally of a decommissioned nuclear reactor.

Abstract: The radioactivity of a metal part is reduced in the process. In the method, firstly a layer of oxide is removed from the metal part using a decontamination solution. Then, an agent which has an oxidizing action and is still present is removed from the decontamination solution. As a result, a layer of the metal is removed. Since radionuclides are to be found only in the layer of the metal part which is close to the surface, the remaining metal can be scrapped in the conventional way.

Abstract: A method for removing radioactive contaminants from a given surface. An aqueous solution having a wetting agent and a complex substituted keto-amine is provided. The solution is left on the surface for a time sufficient to remove the radioactive contaminants into the aqueous solution which is then removed. Depending on the type and condition of the surface, a concentrated acid may be added to the aqueous solution to aid in the contaminant removal process. In such a case, a pH of less than 3.0, and preferably less than 1.5 is maintained. If a concentrated acid is used, the acidic solution containing radioactive contaminants is preferably neutralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: Disclosed is an integrated and streamlined system and method for collecting, containing and processing different types of wastes which provides an economically efficient solution of treating wastes and utilizing the collected data in an operationally useful manner either or both of the waste generator and processor. The present system and method include waste containment in compliance of the laws and governmental regulations, and processing of the wastes. In one embodiment, the system and method also includes compilation and utilization of data relating to the collected and processed wastes.

Abstract: An apparatus and method for stabilizing a spent nuclear fuel containment cask so as to prevent tipping under design basis earthquake conditions. In one aspect, the apparatus comprises a lamina and a slid plate in surface contact with the lamina, forming a slidable interface; wherein the slidable interface has a coefficient of friction within a range resulting in an acceptable level of net horizontal displacement of a cask resting on the apparatus under design basis earthquake conditions.

Abstract: The invention comprises a method for the dissolution of spent nuclear fuel pins or contaminated cladding, the method comprising contacting together the pins and both nitric and fluoroboric acids which, preferably, are present in admixture. The combination of acids dissolves both cladding and fuel at an acceptable rate and provides a solution suitable for use in subsequent purification processes such as the Purex process.

Abstract: A container body of a concrete cask includes a storage portion that stores a canister. A refilling system includes a reduction cylinder having an outside diameter smaller than the diameter of the storage portion and an inside diameter larger than the diameter of the canister, a first lift mechanism configured to raise and lower the reduction cylinder between a down position in the storage portion and an up position wherein it is drawn out of the storage portion, and a second lift mechanism having a holding portion for holding one end portion of the canister and configured to raise and lower the canister with respect to the container body. In refilling operations, the canister is loaded into and unloaded from the storage portion of the container body by the second lift mechanism with the reduction cylinder set in the storage portion.

Abstract: A boiling water reactor is chemically decontaminated by circulating a decontamination solution through reactor recirculation loops and the annulus region of a reactor pressure vessel that surrounds the central core region while bypassing the central core region. The decontamination solution may also be circulated between the annulus region and a lower internals region while bypassing the central core region. The solution dissolves or breaks down metal oxide layers on the surfaces of the boiling water reactor. The metal oxide layers in the central core region and the activated metal ions contained in these layers, which do not substantially contribute to personnel exposure, are not released and, therefore, do not need to be removed from the solution.

Abstract: The present invention is a delivery system for safety injecting solid waste particles, including mixed wastes, into a molten salt bath for destruction by the process of molten salt oxidation. The delivery system includes a feeder system and an injector that allow the solid waste stream to be accurately metered, evenly dispersed in the oxidant gas, and maintained at a temperature below incineration temperature while entering the molten salt reactor.

Abstract: A stable treatment process for a radioactive material (100) includes combining a sufficient minority ratio of said material (100) and a sufficient quantity ratio of a suitable moderator material (102) within a sufficient majority ratio mass of suitable benign radioactive poison metals (104). The benign radioactive poison metals (104) shall include a minimum of one benign radioactive poison metal Lead. Said radioactive material (100), moderator (102) and benign radioactive poison metal (104) being sufficiently combined results in unstable combined mass (105). Said mass (105) consisting of constituents (100,102,104) is impacted by suitable means producing sufficient force (63) from impact upon mass (105) to render (64) sufficient high-pressure four dimensional complete inelastic collisions amongst the constituents (100,102,104) within combined mass (105).

Abstract: A method for removing radioactive contaminants from a given surface. An acidic solution having a pH of less than 3.0, and preferably less than 1.5, of a complex substituted keto-amine, and a mixture of a saturated and unsaturated lower alcohol (e.g. isopropyl alcohol and propargyl alcohol), and concentrated acid is applied to a contaminated surface. The solution is left on said surface for a time sufficient to remove the radioactive contaminants into the acidic solution and then removed. The acidic solution containing the radioactive contaminants is preferably neuralized by an alkaline material to a pH of between 5.5 and 9.0. Removal of thorium contamination from railcars is one useful application of the invention. The method of the present invention has the effect of removing substantially all of the radioactive contaminants from a previously contaminated surface.

Abstract: Disclosed is a radioactive waste treatment process for transmuting long-lived radioisotopes into short-lived radioisotopes through applied nuclear physics. Nuclear reactions, specifically of the (&ggr;, n) type, also known as photodisintegration, are utilized to accomplish this transmutation from troublesome, long-lived radioactive waste isotope(s) of given atomic mass to shorter-lived or stable materials of lower atomic mass, by exposing the troublesome isotopes to a high energy photon flux for a sustained time. Generally speaking, the target nucleus of the radioisotope(s) to be treated is irradiated by gamma photons of an energy greater than the binding energy of the neutron in the target nucleus. This causes the irradiated nucleus to absorb the gamma rays, thereby placing the nucleus in an excited state. Upon relaxation, the nucleus ejects a neutron through the (&ggr;, n) reaction, thereby transmuting the element to an isotope of lower atomic mass and shorter half-life.

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.