Abstract: A system, apparatus and method of processing and/or removing radioactive materials from a body of water that utilizes the buoyancy of the water itself to minimize the load experienced by a crane and/or other lifting equipment. In one aspect, the invention is a method comprising: a) submerging a container having a top, a bottom, and a cavity in a body of water having a surface level, the cavity filling with water; b) positioning radioactive material within the cavity of the submerged container; c) raising the submerged container until the top of the containment apparatus is above the surface level of the body of water while a major portion of the container remains below the surface level of the body of water; and d) removing bulk water from the cavity while the top of the container remains above the surface level of the body of water and a portion of the container remains submerged.

Abstract: A method for restraining a chemical discharge comprising (a) deploying a binding agent into a receptacle containing a hazardous material in a liquid state upon the occurrence of at least one predetermined event that increases the risk of accidentally discharging or leaking the hazardous material from the receptacle; and (b) contacting the hazardous material with the binding agent to form a composition comprising at least a portion of the hazardous material and the binding agent and having at least one property selected from a solid or semisolid state, a viscosity greater than the viscosity of the hazardous material at ambient conditions, a vapor pressure lower than the vapor pressure of the hazardous material at ambient conditions, and a surface tension greater that the surface tension of the hazardous material.

Abstract: The invention describes a method for stripping alumina which is particularly suitable for removal of fluoride from alumina and comprises washing said alumina with an aqueous acid or alkali at elevated temperature. The method may be used for removal of unreacted radiofluoride such as [18F]fluoride from alumina following a radiofluorination reaction. Automated synthesis apparatus and cassettes therefor, which are adapted to perform the method are also claimed.

Abstract: Systems and methods for reducing the volume of radioactive waste materials through desiccation, pyrolysis and vitrification carried out by microwave heating. The final product of the advanced microwave system is a dryer, denser, compacted waste product. The invention comprises systems in which a layer of waste material is treated by microwaves within a hopper before deposited within the final waste container; systems in which a thin layer of waste material is treated by microwaves after it has been deposited within the final waste container; and systems in which waste material is treated by microwaves within a hopper before being deposited within the final waste container.

Abstract: The invention concerns a process enabling the complete combustion and oxidation of the mineral fraction of combustible waste contained in an apparatus intended to treat waste by direct incineration-vitrification, said process comprising the following steps: a step to add said waste to the apparatus for its depositing on the surface of a molten glass bath contained in the apparatus, an incineration and oxidation step of the waste on the surface of the glass bath, an incorporation step to incorporate combustion products in the glass during which the glass bath, the combustion products and any vitrification additives added to the glass bath are heated until a paste-like, liquid mass is obtained, a step during which said mass is removed from the apparatus and left to cool to obtain finally what is called a confinement matrix, said process being characterized in that the complete combustion and oxidation of the waste is achieved partly during the waste adding step and partly during the waste incineration and ox

Abstract: The present invention provides a method for extraction of metals selected from Cr, Mo, Pd, Tc, W, Re, and Pu using a new extractant of methyliminobisalkylacetamide represented by a formula (I): CH3—N—(CH2CONR2)2??(I) wherein R represents an alkyl group having 8-12 carbon atoms.

Abstract: The present invention relates to a treatment of high-level waste of radiochemical production containing radionuclides and macro-admixtures including sodium. The method of extraction of radionuclides by processing acidic aqueous waste solutions by extractants containing macrocyclic compounds selected from the group of crown ethers having aromatic fragments containing alkyl and/or hydroxyalkyl substituents of a linear and/or branched structure, and/or cyclohexane fragments containing alkyl and/or hydroxyalkyl substituents of a linear and/or branched structure, and/or fragments of —O—CHR—CH2O—, where R is the normal or branched alkyl or hydroxyalkyl in organic solvents containing polyfluorinated telomeric alcohol 1,1,7-trihydrododecafluoroheptanol-1 having the formula H(CF2CF2)nCH2OH, where n=3, and a mixture of polyoxyethylene glycol ethers of synthetic primary higher aliphatic alcohols of a fraction C12-C14 of a general formula CnH2n+1O(C2H4O)mH, where n=12-14, m=2 is proposed.

Abstract: The present invention relates to a liquid radioactive waste treatment system. The liquid radioactive waste treatment system includes a plurality of evaporation plates and each of the evaporation plates has an uneven surface, in a housing comprised of a glass. A liquid radioactive waste is dispersed via a liquid waste dispersing unit to the evaporation plate, and the liquid radioactive waste is evaporated using solar heat and airflow in the housing.

Abstract: A method for treating a radioactive liquid waste containing a sodium salt, which includes: feeding a radioactive liquid waste containing at least one of sodium hydroxide, sodium hydrogencarbonate and sodium carbonate to an anode chamber in an electrolytic cell provided with an anode and a cathode on both sides of a permeable membrane, which is selectively permeable to sodium ions, and electrodialyzing the radioactive liquid waste; separating sodium ions permeated through the permeable membrane as sodium hydroxide from the radioactive liquid waste in a cathode chamber; separating a radioactive substance remaining in the anode chamber as a concentrated radioactive liquid waste; and recovering the separated sodium hydroxide and concentrated radioactive liquid waste, respectively.

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: The present invention provides a method for extraction of metals selected from Cr, Mo, Pd, Tc, W, Re, and Pu using a new extractant of methyliminobisalkylacetamide represented by a formula (I): CH3—N—(CH2CONR2)2 ??(I) wherein R represents an alkyl group having 8-12 carbon atoms.

Abstract: A method for treating nuclear sludge comprising subjecting the nuclear sludge to a plasma treatment in a plasma chamber to melt at least some of the inorganic components of the sludge, wherein the plasma chamber comprises a crucible having a cooled inner surface, this surface cooled sufficiently such that the inorganic components in contact with the inner surface are in a solid state and form a barrier between the part of surface of the crucible with which they are in contact and the molten inorganic components of the sludge.

Abstract: The present disclosure relates to a process for recycling a sodium salt by decomposition of a sodium nitride liquid waste, comprising a neutralization step in which a nitric acid liquid waste or an off-gas having nitric acid dissolved therein which is produced through a wet reprocessing process comprising a dissolution step for dissolving a spent nuclear fuel in nitric acid is neutralized by adding or contacting the nitrate liquid waste or the off-gas to or with at least one sodium salt selected from the group consisting of sodium hydroxide, sodium hydrogencarbonate and sodium carbonate, thereby yielding a sodium nitrate liquid waste; a sodium nitrate-decomposition step in which the sodium nitrate liquid waste is reductively decomposed with a reducing agent, thereby decomposing sodium nitrate into a nitrogen gas and the sodium salt; and a recycle step for recycling the sodium salt into the neutralization step or wet reprocessing process.

Abstract: This present invention is related to a method of collecting high-level radioactive contaminated uranium powder underwater. It utilizes a long rod fixed to the pipe of the sewage pump at bottom, and takes another end of the sewage pump to connect to the sediment bag to collect uranium powder which is dispersed and contaminates the pond. Inhaling the pond water contains uranium powder into the sediment bag by sewage pump that is joined to a collecting can under it and waiting for precipitation of the powder. When the uranium powder in the sediment bag is completely precipitated into the collecting can, operator can utilize submersible pump to drain the pond water from the sediment bag. Operator may hold the cable wire of the uranium powder collecting can manually to move it to the collecting shelf and repeat the process several times until the high-level radioactive uranium powder in the contaminated pond are all collected on the uranium powder collecting shelf.

Abstract: The present invention describes a process for treating phosphogypsum and in particular the related leachate that comprises the steps of feeding a continuous leachate flow with a dry content of below 5% by weight into a concentration unit heated to a predetermined temperature and equipped with dedicated mixing means, and continuously discharging a concentrated leachate flow having a dry content comprised between 20% and 95% by weight from the concentration unit.

Abstract: A system for chemically decontaminating radioactive material.

Abstract: The present invention relates to an automatic radioactivity analyzer of mixed liquid beta emitter which comprises: a sample preparation part (1, 2, 3, 4, 5, 6, and 7) for extracting a liquid sample from liquid-phase radioactive nuclear wastes; a sample injection part (9 and 10), including a sample transportation part for transporting a bottled sample to a radioactivity detection part to perform measurement; the radioactivity detection part (11) including two photon multiplier tubes; an exterior gamma-ray source injection part (12) for compensating for measurement efficiency according to quenching effects; a signal processing part (13), including a pre-amplifier circuit (14), a high-voltage generator circuit (15), an analogue-to-digital converter circuit (21), and a digital signal processor (DSP) (24), for generating beta spectrums by the aid of a fast coincidence counter (20) and a multi-channel analyzer (22); a main control PC (25) and a graphic user interface (GUI) program (29) for remotely automatically me

Abstract: A method and apparatus for treating radioactive waste water containing contaminating ions, colloids and suspended solids having like (usually negative) charges preventing their precipitation. An electric current is passed through the waste water in an EC assembly to cause electro-coagulation of the contaminants and anodes of this assembly are made of a metal that dissolves to provide cations for neutralizing the negative charges and forming precipitates containing neutralized contaminants. Precipitates are then separated from waste water by an electro-magnetic or other filtering unit. The water pH and conductivity may be adjusted before the EC assembly and additives may be introduced into its effluent for enlargement of precipitate particles, improvement of filtration, improvement of dewaterability, and/or enhancement of magnetism.

Abstract: Methods for removing nuclear waste from a component are provided. In one embodiment, the method includes the steps of supplying oxalic acid to the nuclear waste on the component to form an oxalic acid/waste solution, feeding the oxalic acid/waste solution from the component, feeding an oxidant to the oxalic acid/waste solution to form carbon-dioxide, water and a precipitate, and separating the precipitate from the water. A system for removing nuclear waste is also provided.

Abstract: Radiation shields and techniques for radiation shielding are provided. Bitumen substances, such as asphalt or tar, are mixed with radioactive waste, leaded glass, or a radioactive waste and leaded glass composite. In embodiments where the bitumen substance is mixed with leaded glass, the resulting mixture can be coated onto containers that house radioactive waste or the resulting mixture can be coated onto the outer surface of the radioactive waste.

Abstract: A method of removing metal scale from surfaces that includes contacting the surfaces with a first aqueous solution of a chelating agent, allowing the chelating agent to dissolve the metal scale, acidifying the solution to form a precipitant of the chelating agent and a precipitant of the metal from the metal scale, isolating the precipitant of the chelating agent and the precipitant of the metal from the first solution, selectively dissolving the precipitated chelating agent in a second aqueous solution, and removing the precipitated metal from the second solution is disclosed.

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

Abstract: A mixed extractant solvent including calix[4]arene-bis-(tert-octylbenzo)-crown-6 (“BOBCalixC6”), 4?,4?,(5?)-di-(t-butyldicyclo-hexano)-18-crown-6 (“DtBu18C6”), and at least one modifier dissolved in a diluent. The mixed extractant solvent may be used to remove cesium and strontium from an acidic solution. The DtBu18C6 may be present from approximately 0.01 M to approximately 0.4M, such as from approximately 0.086 M to approximately 0.108 M. The modifier may be 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol (“Cs-7SB”) and may be present from approximately 0.01M to approximately 0.8M. In one embodiment, the mixed extractant solvent includes approximately 0.15M DtBu18C6, approximately 0.007M BOBCalixC6, and approximately 0.75M Cs-7SB modifier dissolved in an isoparaffinic hydrocarbon diluent. The mixed extractant solvent may form an organic phase in an extraction system that also includes an aqueous phase. Methods of extracting cesium and strontium as well as strontium alone are also disclosed.

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: 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: The invention concerns a ligand comprising wherein n is an integer from 1 to 5, X represents —NO2, —NH2, —NCS, —NHCOCH2-Z. NHCO—W—COCNHS, —NH-Q, —NHCS-Q, —NHCOCH2-Q, or —NHCO(CH2)m ?-Q where Q is an hapten chosen from the group consisting of steroids, enzymes, proteins, monoclonal antibodies, chimeric antibodies, or fragments thereof or any activated linker ready for coupling reaction, Y is CO2H or PO3H2 W is —(CH2)m— m is an integer from 1 to 10.

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 invention concerns a ligand comprising (I) wherein n is an integer from 1 to 5 Y is CO2H or PO3H2T represents —X or -phenyl-X, wherein X represents NO2, NH2, NCS, NHCOCH2-Z, NHCO—W—COCNHS, —NH-Q, —NHCS-Q, —NHCOCH2-Q, or NHCO(CH2)m-Q where Q is a hapten chosen from the group consisting of steroids, enzymes, proteins, monoclonal antibodies, chimeric antibodies, or fragments thereof or any activated linker ready for coupling reaction, W is —(CH2)m- m is an integer from 1 to 10 Z is chloride, bromide or iodine

Abstract: This invention concerns a method and an installation for the treatment of radioactive wastes produced as a consequence of the operation of nuclear power plants with pressurized water reactors and boron reactivity regulation and will take place for the production of substances like borax with environmentally allowed content of radioactive isotopes that contains only the cesium isotopes with maximum total concentration of 800 Bq in one kilogram of borax; calcium, magnesium or calcium-magnesium borates with environmentally allowed content of radioactive isotopes; boron acid solution with environmentally allowed content of radioactive isotopes; sodium hydroxide solution containing only the cesium isotopes with maximum total concentration of 800 Bq in one kilogram sodium hydroxide; radioactive waste containing under 5 g/l of boron acid.

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 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 pump station for transferring radioactive particle containing waste water, includes: (a.) an enclosed sump having a vertically elongated right frusto conical wall surface and a bottom surface and (b.) a submersible volute centrifugal pump having a horizontally rotating impeller and a volute exterior surface. The sump interior surface, the bottom surface and the volute exterior surface are made of stainless steel having a 30 Ra or finer surface finish. A 15 Ra finish has been found to be most cost effective. The pump station is used for transferring waste water, without accumulation of radioactive fines.

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

Abstract: A 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 present invention relates to apparatuses for processing homogeneous/heterogeneous radioactive wastes comprising ion-exchange reisns. A cooled discharge unit comprises a discharge pipe, a cooling jacket having a U-shaped form in cross section, a collector for feeding a coolant into the jacket, a discharge gate comprising a pipe, on one end of which a cone-shaped tip is positioned, on the other end a lid with an aperture. A cooled induction melter comprises a housing, side walls and bottom of which are made of metal pipes disposed with a gap therebetween and combined by a collector for supplying and discharging the coolant, an inductor positioned adaptable for displacement along the longitudinal axis of the melter and concentrically encompassing the side walls of the housing, the gaps between the pipes of which ensure transparency of the housing for an electromagnetic field of the inductor.

Abstract: A process for reducing a level of radionuclei in a phosphogypsum material having a radioactive contaminant is provided, wherein the process includes the steps of: (a) mixing the phosphogypsum material with an acid decontamination solution to form a mixture; (b) separating the mixture into a liquid stream having a large portion of the radioactive contamination and a solid stream having a residual component of the acid decontamination solution; and (c) removing the residual component of the acid decontamination solution from the solid stream, thus providing a gypsum material that meets or exceeds federal standards for level of radioactive nuclei and providing a process that is environmentally friendly and capable of avoiding the generation of phosphogypsum stacks, as well as eliminating existing or abandone phosphogypsum stacks, further providing for recovery of the radioactive nuclei if desired.

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: A decontamination formulation is provided which is effective against a broad spectrum of chemical and biological warfare agents and radioactive dusts, comprising an active decontamination agent, a co-solvent, a buffer system to optimize the initial reaction pH above 8.5 and more preferably in the range of 10 to 11 for favoring oxidation of VX and HD and hydrolysis of G agents, and a surfactant similar to fire-fighting foaming agent. Formulations comprise, in water by weight, 1% to 15% of a hydrated chloroisocyanuric acid salt, 1% to 10% of a polypropylene glycol co-solvent, 1% to 15% surfactant and a buffer system to initially maintain said formulation at a pH from about 11 to about 8.5 for sufficient duration to effect decontamination. The formulation can be provided in kit form or concentrate form, be prepared, in part, in advance or on site, and be dispensed in foam form which aids in coating and adhering of the decontamination formulation to contaminated surfaces.

Abstract: Processes for the treatment of solutions used for the decontamination of radioactively contaminated surfaces wherein the solution contains radioactive metal ions and organic complexing agents are described herein. The processes include treating the solutions with a reagent suitable for the destruction of the complexing agent and contains a metal capable of existing in more than one oxidation state, and raising the pH of the resultant solution to a level at which the metal of the reagent precipitates or flocs out of the solution. Processes in which the contaminated solutions are treated with electromagnetic radiation, treated with UV or visible radiation, and treated at an ambient temperature are also described herein.

Abstract: An extractant medium for extracting alkaline earth cations from an aqueous acidic sample solution is described as are a method and apparatus for using the same. The separation medium is free of diluent, free-flowing and particulate, and comprises a Crown ether that is a 4,4′(5′)[C4-C8-alkylcyclohexano]18-Crown-6 dispersed on an inert substrate material.

Abstract: This invention relates to a method for conditioning a waste constituted of an aqueous solution of sodium hydroxide NaOH of 3 to 10 M, possibly radioactive.

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

Abstract: Radioactive laundry liquid wastes are supplied in a liquid waste heating vessel. Hydrogen peroxide and an alkali solution are supplied to the liquid waste heating vessel. pH of radioactive laundry liquid wastes is adjusted to 7 or higher by the alkali solution. The radioactive laundry liquid wastes are heated to 50° C. or higher by a heating device. The heated radioactive laundry liquid wastes are introduced to first and second aeration vessels. Ozone is supplied from an ozone generator by way of an ozone gas discharge port to the first aeration vessel. Ozone discharged from the first aeration vessel is introduced from the ozone gas discharge port to the second aeration vessel. Therefore, the amount of ozone dissolved into the radioactive laundry liquid wastes is increased so that the amount of hydroxy radicals formed for decomposing organic substances increases, since the laundry liquid wastes are heated to 50° C. or higher under the presence of hydrogen peroxide.

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

Abstract: A groundwater decontamination system includes a decontamination unit adapted to be lowered down through a non-pumping well into an aquifer containing contaminated water. The unit can also be raised up and out of the non-pumping well for cleaning, servicing or replacement. The decontamination unit includes a porous outer tube with a plurality of holes through which contaminated water flows, and a porous inner tube with a plurality of holes through which flows in-flowing contaminated water flowing through outer tube. A contaminant removing reactive barrier material is disposed within the inner tube for removing on contact contaminants from the in flowing water. A flow directing arrangement, preferably in the form of flow directing fins, directs in-flowing water from the holes in the outer tube to the holes in the inner tube (i.e., provider of the channeling of the ground water into the reactive barrier material). The system can monitor the contaminant removal, as well as other conditions of the groundwater.

Abstract: A permeable barrier for decontaminating groundwater having two distinct components for increasing contaminant removal. Preferably, the barrier material is a mixture or distinct layers of bone-char phosphate material and iron oxide pellets. The barrier material can be incorporated as part of a shallow trenching decontamination system, or incorporated in a non-pumping well, or array of non-pumping wells, as part of a deep underground decontamination system. The system can be used for removing, among other things, a radionuclide, such as uranium, from water.

Abstract: A process for improving waste vitrification in a disposable canister. In the process, chunks of frit are combined with waste and then added to the disposable canister. The waste and frit are then heated to melt the frit and waste. The melted mixture is then cooled to form a vitrified product in the module.

Abstract: The present invention provides a method, and an apparatus, for forming a prescribed concentration of a substance in a mixture with a fluid, from a comparatively dilute mixture. The mixture is most preferably a solution of a solute in a solvent. The following summary and description generally discuss the invention in terms of a particularly preferred solution and a preferred use for that solution; however, in its broadest embodiments, the invention encompasses both the use for other solutions, and the use for a suspension of a solid substance in a fluid to form a mixture (which can also be referred to as a slurry). The invention provides for significant economic, safety and qualify benefits over conventional evaporative systems.