Abstract: An improved system for receiving and storing a radioactive salt solution includes a tank configured to receive the radioactive salt solution while preventing criticality accidents, a solution inlet for carrying the radioactive salt solution to the tank, an overflow bottle, and a cap sealing the top end of the tank. The cap includes a lateral wye fitting having a lateral pipe configured to direct the radioactive salt solution from the solution inlet into the tank, a vertical pipe configured to direct gases from the tank to a ventilation system, and an overflow line configured to carry excess radioactive salt solution from the tank to the overflow tank. An air gap between the lateral pipe and the solution inlet prevents backflow of the radioactive salt solution into the solution inlet.

Abstract: The invention discloses a phosphorus nitride adsorbent with high-efficiency selectivity, and its application thereof. The phosphorus nitride adsorbent has a mutually cross-linked hollow tubular structure. The adsorbent can have an adsorption capacity of 435.58 mg·g?1 and 7.01 mg·g·1 for spiked seawater and natural seawater with a uranium concentration of 350 ppb, and the adsorbent has a long service life, and can still maintain 91.14% of the initial adsorption capacity after 5 cycles of adsorption and desorption. Taking into account the advantages of a short material preparation cycle, a wide range of raw material sources, a low cost, an excellent adsorption performance, and long service life, the adsorbent can be used in technical fields such as uranium-containing wastewater treatment, uranium ore resource recovery, uranium extraction from seawater and the like.

Abstract: The present invention relates to a method for reducing the HTO concentration in a tritium-containing aqueous solution. The present invention includes bringing water vapor or the like of a tritium-containing aqueous solution into contact with a porous material having pores in a pore diameter range of 500 ? or less, selectively occluding the HTO in the tritium-containing aqueous solution in the porous material, and obtaining a tritium-containing aqueous solution in which the HTO concentration thereof is reduced. The present invention relates to a device used for reducing the HTO concentration in a tritium-containing aqueous solution.

Abstract: A method of decontaminating a metal surface located on a component within a nuclear plant, in particular within the cooling system of a nuclear power plant, which is covered with a metal oxide layer containing radioactive substances, the method including a decontamination step in which a metal oxide layer pretreated in an oxidation step is contacted with an aqueous solution of an organic acid to dissolve the metal oxide layer, forming a decontamination solution containing the organic acid, metal ions and the radioactive substances, and passing the decontamination solution over an ion exchanger to immobilize metal ions and radioactive substances. An oxidant selected from oxygen, air, hydrogen peroxide and ozone is dosed into the decontamination solution to control the dissolution rate of the metal oxide layer. The method is particularly suitable for large-scale system decontamination and ensures high process stability.

Abstract: Disclosed herein are systems, methods, processes, and apparatuses for treating radioactive waste, through systems designed to bind and dry radioactive media. In some of its various embodiments, the system includes at least one helical screw designed to receive and mix liquid wastes with ion exchange media, as well as convey the resulting slurry through one or more of a binding, dewatering, and drying/off-gassing region.

Abstract: A method of decontaminating metal surfaces in a cooling system of a nuclear reactor comprises: an oxidation step, comprising at least one acidic oxidation step and at least one alkaline oxidation step wherein metal oxides and radioisotopes on the metal surfaces are contacted with aqueous permanganate oxidant solutions; followed by a decontamination step wherein an aqueous solution comprising oxalic acid, formic acid, citric acid, tartaric acid, picolinic acid, gluconic acid glyoxylic acid or mixtures thereof is used to dissolve at least part of the metal oxides and radioisotopes; and a cleaning step wherein radioisotopes are immobilized on an ion exchange resin; wherein at least one treatment cycle includes a high temperature oxidation step, wherein the permanganate oxidant solution is kept at a temperature of at least 100° C.

Abstract: Apparatus and methods for treating radioactive material, in particular for cleaning radioactive contaminated water, are described. One apparatus comprises a process chamber with a combustion zone for generating an oxygen rich gas and an oxidation zone, which is arranged to receive the oxygen rich gas from the combustion zone. The process chamber further comprises a feed opening for feeding the radioactive material into the oxidation zone and is configured to use the oxygen rich gas for oxidizing the radioactive material to obtain oxidized material. The apparatus further comprises a separation device operationally connected to an outlet of the process chamber and configured to at least partly separate the oxidized material into a gaseous fluid and a non-gaseous residue. This way a greatly reduced volume of the radioactive material is achieved, enabling safe and efficient handling and/or compact and space-saving disposal of the radioactive material.

Abstract: The invention relates to a technique for handling liquid radioactive waste from a nuclear fuel-energy cycle, and may be used in a process for processing liquid radioactive waste for maximally reducing the volume thereof and removing radionuclides by concentrating same in a solid phase. The aim is achieved by means of a method for processing liquid radioactive waste and for the recovery thereof, including waste oxidation, separating sludge, colloids and suspended particles from a liquid phase, and removing, from the liquid phase, radionuclides to be subsequently recovered using selective sorbents and filters; the method is characterized in that, prior to the stage for separating sludge, colloids and suspended particles from the liquid phase of the radioactive waste, selective sorbents in the form of powders are added and mixed into the liquid waste.

Abstract: The present invention relates to a method for treating tritium water-containing raw water, the method including supplying a part of raw water containing tritium water and alkali water to a circulation tank, mixing the raw water with alkali water in the circulation tank to obtain an electrolyte adjusted so as to have a desired alkali concentration, and continuously electrolyzing the electrolyte while circulating the electrolyte, thereby subjecting the raw water stored in the storage tank to alkali water electrolysis and thus gasifying the raw water. According to the invention, by gasifying tritium water-containing raw water by alkali water electrolysis, the tritium concentration in a tritium-containing hydrogen gas is diluted to 1/1,244 and the tritium water-containing raw water can be reduced in volume.

Abstract: An object to be decontaminated contaminated with radioactive material, e.g., contaminated soil or water, is introduced into eluting solvent and dissolved, and the radioactive material is separated from the object to be contaminated by elution of the radioactive material into the eluting solvent. The eluting solvent containing the radioactive materials dissolved therein and the object to be decontaminated are separated into solid and liquid. The soil after solid-liquid separation and from which the radioactive material is removed is collected, and the eluting solvent after solid-liquid separation and a separated liquid containing contaminated water are introduced into an electrolysis tank and electrolyzed. Metal ions such as those of the radioactive materials are deposited on the cathode in the electrolysis tank. Hydrogen containing tritium generated in electrolysis is collected in the electrolysis tank. The hydrogen is moved to the outside of the electrolysis tank and trapped.

Abstract: An object to be decontaminated contaminated with radioactive material, e.g., contaminated soil or water, is introduced into eluting solvent and dissolved, and the radioactive material is separated from the object to be contaminated by elution of the radioactive material into the eluting solvent. The eluting solvent containing the radioactive materials dissolved therein and the object to be decontaminated are separated into solid and liquid. The soil after solid-liquid separation and from which the radioactive material is removed is collected, and the eluting solvent after solid-liquid separation and a separated liquid containing contaminated water are introduced into an electrolysis tank and electrolyzed. Metal ions such as those of the radioactive materials are deposited on the cathode in the electrolysis tank. Hydrogen containing tritium generated in electrolysis is collected in the electrolysis tank. The hydrogen is moved to the outside of the electrolysis tank and trapped.

Abstract: A process for reducing a radioactive material of an object containing a radioactive material to a safe level in a living environment includes releasing pressure including carrying out a heating process on an aqueous liquid that includes water, water-soluble liquid, or a mixture of water and water-soluble liquid, and the object in a state where temperature is less than or equal to a critical temperature of the aqueous liquid and pressure is greater than or equal to a saturated vapor pressure of the aqueous liquid when the object is immersed in the aqueous liquid to such a degree that the object is covered or to a greater degree and then abruptly releasing the pressure, and separating, after the releasing pressure, the mixture of the object and the aqueous liquid into liquid and solid.

Abstract: A method of preparing a simple ceramic ingot of a spent filter having radioactive cesium trapped therein, and a ceramic ingot of a spent filter having improved properties such as leach resistance, thermal stability, and cesium content are provided. The method includes grinding and mixing a spent filter having cesium trapped therein, adding a solidifying agent, and sintering the spent filter. The method of preparing a ceramic ingot of a spent filter can be useful in preparing the ceramic ingot of the spent filter from only the spent filter by means of simple grinding and sintering, and in preparing the ceramic ingot of the spent filter by adding a small amount of a solidifying agent. The ceramic ingot of the spent filter has a high density and improved thermal stability, and shows improved leach resistance since a leach rate of a radioactive material is remarkably low. Therefore, the spent filter having radioactive cesium trapped therein can be effectively used to prepare a stable ceramic ingot.

Abstract: A method is described for the treatment of ferrous nuclear waste, produced in operations of pickling of contaminated metallic surfaces, which makes it possible to minimize the volumes of liquid phases to be treated and to recycle the by-products thereof to the process.

Abstract: Provided are a process and device for reducing radioactive material of an object containing the radioactive material to a safe level in a living environment. Included are a step of performing at least a step of carrying out a heating process on the object, into which radioactive material is absorbed and/or adsorbed from an environment or which absorbs and/or adsorbs radioactive material from an environment, in a state where temperature is less than or equal to the critical temperature of water and pressure is greater than or equal to the saturated vapor pressure of water, or a step of abruptly releasing the pressure; and a step of separating, after the above step, into liquid and solid.

Abstract: The invention relates to the treatment of carbonaceous radioactive waste, comprising the delivery of waste to one or more radioactive isotope separation stations isotopes, said isotopes being among at least carbon 14, chlorine 36, and tritium. Advantageously, the delivery to each of the stations occurs in wet form, with water being a common medium for conveying the waste to each of the separation stations.

Abstract: A waste form for and a method of rendering hazardous materials less dangerous is disclosed that includes fixing the hazardous material in nanopores of a nanoporous material, reacting the trapped hazardous material to render it less volatile/soluble, and vitrifying the nanoporous material containing the less volatile/soluble hazardous material.

Abstract: The invention is directed to a purifying device for sludge under water and a method for operating the same. The device includes a main fixing frame having an accommodating portion assess to the outside, a hollow liquid container in the accommodating portion, wherein the liquid container is provided with a liquid-flow hole and at least a backwash hole, multiple filters on the liquid container, and a pump connected to the liquid-flow hole at the liquid container through a liquid pipeline. The method includes steps: moving the liquid container having the filters to an area having a liquid to be filtered; leading the liquid to flow into the liquid container through the filters filtering out solid particles contained in the liquid; and leading a fluid to flow into the liquid container such that the filters can be backwashed accompanying with an external cleaning device if the filters are clogged.

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 and apparatus for the stabilization and safe removal of buried waste that is tested and classified as being transuranic or not transuranic waste and disposed accordingly. The buried waste (usually in vertical pipe units) is enclosed in a casing and ground and mixed with the surrounding soil. This process allows for chemical reactions to occur that stabilizes the mixture. The entire process is contained within the casing to avoid contamination. In situ or external testing is done for radio isotopes to classify the waste. If it is classified as transuranic the waste is removed in a controlled way into a retrieval enclosure and disposed off in drums. If the waste is not transuranic then grout is introduced into the mixture, allowed to set and the resulting monolith is removed and buried in trenches.

Abstract: A method of removing contaminants from slurry samples is set forth. The method includes the utilization of repeated pressurizing and depressurizing steps to disrupt solidified particles in solid-containing slurries thereby increasing decontamination efficiency. An expansion fluid is injected into the slurry sample sufficient to create microbubbles when the slurry sample is depressurized. The micro bubbles mechanically disrupt the solidified particles increasing contaminant exposure. The microbubbles also provide for increased interfacial regions where contaminants can accumulate at gas-liquid thin films that are in close proximity to and can be effectively removed using a suitable expansion fluid and optional decontamination agents.

Abstract: An apparatus for treating a radioactive nitrate waste liquid includes: a denitrification tank (12) that accommodates active sludge which adsorbs or takes in a radioactive substance in a nitrate waste liquid (11) containing nitrate and the radioactive substance and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows; a reaeration tank (14) that aerates and mixes a denitrification-treated liquid (24) treated in the denitrification tank (12) with the active sludge in which the aerobic microorganism grows; and a sludge dissolution tank (81) that dissolves redundant sludge (26A, 26B) discharged from the denitrification tank (12) and the reaeration tank (14). Acetic peracid (80) is supplied to the sludge dissolution tank (81) to dissolve redundant sludge, a sludge lysate is supplied to the denitrification tank (12) as a carbon source (22), and acetic acid is supplied to the denitrification tank (12).

Abstract: A process of incorporating technetium into an electroless deposit, forming an alloy that is extremely resistant to corrosion and reduces the mobility of technetium on a geologic time scale is disclosed and claimed. The process includes providing a liquid containing technetium, such as an aqueous waste stream generated during the used nuclear fuel reprocessing activities. The technetium is collected and concentrated, and provided into an electroless deposition bath. A substrate, such as suitably prepared zero valent iron or stainless steel, is introduced into the bath to initiate autocatalytic electroless deposition of the technetium onto the substrate due to the difference in electrochemical potential between the plating bath and ti metals in solution. This causes a layer of technetium metal to form on the substrate.

Abstract: The invention provides methods for the production of radioisotopes or for the treatment of nuclear waste. In methods of the invention, a solution of heavy water and target material including fissile material present in subcritical amounts is provided in a shielded irradiation vessel. Bremsstrahlung photons are introduced into the solution, and have an energy sufficient to generate photoneutrons by interacting with the nucleus of the deuterons present in the heavy water and the resulting photoneutrons in turn cause fission of the fissile material. The bremsstrahlung photons can be generated with an electron beam and an x-ray converter. Devices of the invention can be small and generate radioisotopes on site, such as at medical facilities and industrial facilities. Solution can be recycled for continued use after recovery of products.

Abstract: A method for processing a coolant includes filtering a coolant using a first filtration system to generate a first filtered material, and filtering the filtered coolant using a second filtration system to generate a second filtered material. The second filtration system is different from the first filtration system. The first filtered material is transferred to a first waste treatment container and converted to a first waste product for permanent disposal, and the second waste product is transferred to a second waste treatment container and converted to a second waste product for permanent disposal.

Abstract: A method for removing corrosion products from a system, the method including: adjusting the system temperature to between 115° F. to 212° F.; injecting a cleaning dissolution solvent into the system; injecting a gas into the system after the system is filled with the cleaning dissolution solvent; the gas mixing with the solvent in the system; draining the solvent from the system after a predetermined period of time of dissolution; injecting a passivation composition into the system; injecting a gas into the system, the gas mixing the passivation composition; draining the system of the composition after a predetermined period of time of passivation; rinsing the system with a low volume solution; and rinsing the system at with a full volume solution.

Abstract: A method is described for the treatment of ferrous nuclear waste, produced in operations of pickling of contaminated metallic surfaces, which makes it possible to minimize the volumes of liquid phases to be treated and to recycle the by-products thereof to the process.

Abstract: A method for precipitating uranium from an aqueous solution and/or sediment comprising uranium and/or vanadium is presented. The method includes precipitating uranium as a uranyl vanadate through mixing an aqueous solution and/or sediment comprising uranium and/or vanadium and a solution comprising a monovalent or divalent cation to form the corresponding cation uranyl vanadate precipitate. The method also provides a pathway for extraction of uranium and vanadium from an aqueous solution and/or sediment.

Abstract: A method and device for limiting the degassing of tritiated waste issued from the nuclear industry are provided. The method reduces an amount of generated tritiated hydrogen (T2 or HT) and/or tritiated water (HTO or T2O) including at least one piece of tritiated waste from the nuclear industry. The method includes placing the package in contact with a mixture including manganese dioxide (MnO2) combined with a component that includes silver; and placing the package in contact with a molecular sieve.

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: To provide an apparatus for treating a radioactive nitrate waste liquid that includes a denitrification tank (12A) which accommodates active sludge that adsorbs or takes in the radioactive substance in a nitrate waste liquid (11) and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows, and a reaeration tank (14) in which a denitrification-treated liquid (24) treated in the denitrification tank (12A) is aerated and mixed with active sludge. A pH adjuster (21), a carbon source (22), and nitrogen gas are supplied to the denitrification tank (12A) so as to separate a denitrified liquid into a solid content and the denitrification-treated liquid (24) by using a first solid-liquid separating film (25), and the denitrification-treated liquid (24) treated with the active sludge in the reaeration tank (14) is reaerated and separated into a solid content and a reaeration-treated liquid (27) by using a second solid-liquid separating film (28).

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: Materials and methods of synthesizing mixed-layered bismuth oxy-iodine materials, which can be synthesized in the presence of aqueous radioactive iodine species found in caustic solutions (e.g. NaOH or KOH). This technology provides a one-step process for both iodine sequestration and storage from nuclear fuel cycles. It results in materials that will be durable for repository conditions much like those found in Waste Isolation Pilot Plant (WIPP) and estimated for Yucca Mountain (YMP). By controlled reactant concentrations, optimized compositions of these mixed-layered bismuth oxy-iodine inorganic materials are produced that have both a high iodine weight percentage and a low solubility in groundwater environments.

Abstract: A method for treating before calcination a nitric aqueous solution comprising at least one radionuclide and ruthenium is provided. The method comprises a step for adding to the solution a compound selected from lignins, lignocelluloses, optionally as salts and mixtures thereof.

Abstract: Described are method of treating a radioactive organic waste stream comprising: (a) mixing a radioactive organic waste stream comprising organic compounds and radionuclides with phosphoric acid to form a reaction mixture; (b) heating the reaction mixture to a desired temperature in the presence of an oxidant to oxidize organic compounds present in the waste stream, and removing oxidized organic compounds from the reaction mixture; (c) optionally, adding a reducing agent to the reaction mixture to form insoluble radioactive metal phosphate compounds comprising one or more of the radionuclides, and separating the insoluble radioactive metal phosphate compounds from the reaction mixture; (d) optionally, adding a fluorine compound to the reaction mixture to react with uranium that may be present in the reaction mixture to form uranium hexafluoride, and removing uranium hexafluoride from the reaction mixture; (e) adding ammonia to the reaction mixture to neutralize phosphoric acid and to form ammonium phosphate

Abstract: A filtration separation method for waste resin containing highly radioactive uranium powder and device thereof is mainly used for the waste resin after water treatment process in a nuclear facility. The uranium powder contained in the waste resin is highly radioactive. Thus, prior to the treatment of the waste resin, it is necessary to filter and separate the highly radioactive uranium powder to reduce the radioactivity. It is to put the uranium powder containing waste resin into an underwater holding tank and withdraw the waste resin by an underwater pump into a uranium powder filtration box. Then the uranium powder filtration box is lifted by a hoist to move to an underwater ultrasonic cleaner. High-pressure water flushing is applied and followed by underwater ultrasonic cleaning. Then the cleaning water (containing precipitated uranium powder) from the ultrasonic cleaner is drained to uranium powder collection device.

Abstract: The present invention relates to a process for the manufacture of a glass frit for the containment by vitrification of a material comprising at least one oxidizable or reducible chemical species, and also to a process for the containment of said material by vitrification. The process for the manufacture of the glass frit comprises a step of incorporating into a raw glass frit at least one redox couple, the nature and the amount of which make it possible to maintain said at least one chemical species in its oxidized or reduced state. The containment process comprises mixing and hot melting the resulting glass frit and the material to be contained. The present invention makes it possible to optimize the containment of pollutants such as radionucleides, metals and metalloids. The material may be nuclear waste or a material derived from the incineration of household refuse.

Abstract: Materials and methods of making low-sintering-temperature glass waste forms that sequester radioactive iodine in a strong and durable structure. First, the iodine is captured by an adsorbant, which forms an iodine-loaded material, e.g., AgI, AgI-zeolite, AgI-mordenite, Ag-silica aerogel, ZnI2, CuI, or Bi5O7I. Next, particles of the iodine-loaded material are mixed with powdered frits of low-sintering-temperature glasses (comprising various oxides of Si, B, Bi, Pb, and Zn), and then sintered at a relatively low temperature, ranging from 425° C. to 550° C. The sintering converts the mixed powders into a solid block of a glassy waste form, having low iodine leaching rates. The vitrified glassy waste form can contain as much as 60 wt % AgI. A preferred glass, having a sintering temperature of 500° C. (below the silver iodide sublimation temperature of 500° C.) was identified that contains oxides of boron, bismuth, and zinc, while containing essentially no lead or silicon.

Abstract: A mobile waste processing system and method is disclosed for processing materials in a tank, typically an underground storage tank. A mobile fluid processing module is disposed adjacent to or over the tank, and sealingly connected to the tank through a tank riser. Support equipment, which may include an electric power module, a chiller module, a heating module, and a control and monitor module, are disposed a distance from the processing module, and connected thereto. A pump and optionally a dispersal nozzle are deployed from the processing module into the tank, and effluent is extracted for processing. The processing trailer includes effluent processing equipment such as an evaporator and demister, condenser, vacuum pump, and interim storage tank. The effluent is processed, and at least a portion is removed. Some of the processed effluent may be returned to the tank.

Abstract: The invention provides a method for the separation of solids from a semi-solid viscous mass, the method comprising treating a solids-containing semi-solid viscous mass in an apparatus comprising a separating member incorporating a substantially flat solid surface and a supporting member, wherein the separating member is adapted to selectively impart directional momentum to the solids, thereby facilitating separation of the solids from the semi-solid viscous mass. Preferably the separating member comprises a substantially flat solid surface adapted for vibrational motion, the vibrational motion causing the solids material to be displaced from the remainder of the semi-solid viscous mass.

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: The apparatus is proposed for termination of radioactive and other wastes particularly for physical processing of radioactive waste with simultaneous production of hydrogen, oxygen, and electric energy. It includes an ion divider, a plasma chamber communicated with the ion divider, a controllable hydrogen-oxygen dispenser introducing hydrogen and oxygen into the plasma chamber, transformer pipe coils, a transformer chamber surrounded by the transformer coils, the transformer chamber communicated with the plasma chamber, equipment for supplying water and steam into the transformer coils, cooling equipment for receiving ionized steam from the transformer coils and cooling the steam, a sprayer receiving the ionized steam from the cooling equipment and introducing the steam into the ion divider; and dispenser equipment for introducing radio-active wastes and/or worked-out rocket fuels into the transformer chamber, wherein the radio-active wastes and/or worked-out rocket fuels are terminated.

Abstract: A mixed extractant solvent that includes at least one dialkyloxycalix[4]arenebenzocrown-6 compound, 4?,4?,(5?)-di-(t-butyldicyclohexano)-18-crown-6, at least one modifier, and, optionally, a diluent. The dialkyloxycalix[4]arenebenzocrown-6 compound is 1,3-alternate-25,27-di(octyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(decyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(dodecyloxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(2-ethylhexyl-1-oxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(3,7-dimethyloctyl-1-oxy)calix[4]arenebenzocrown-6, 1,3-alternate-25,27-di(4-butyloctyl-1-oxy)calix[4]arenebenzocrown-6, or combinations thereof. The modifier is a primary alcohol. A method of separating cesium and strontium from an aqueous feed is also disclosed, as are dialkyloxycalix[4]arenebenzocrown-6 compounds and an alcohol modifier.

Abstract: A method of processing ion exchange resin radioactive waste, wherein the radioactive waste contains a plurality of fractions of radioactive waste based on density, which may include cation resin waste and anion resin waste, wherein at least one of the plurality of fractions comprises a total concentration of at least one radionuclide, including performing at least one of the following operations (i), (ii) or (iii): (i) separating the at least one of the plurality of fractions from the radioactive waste utilizing a moving freeboard; (ii) separating the radioactive waste into cation resin waste and anion resin waste; or (iii) removing at least a portion of the total concentration of the at least one radionuclide from: the radioactive waste; the at least one of the plurality of fractions separated by the moving freeboard; or at least one of the radioactive waste, the cation resin waste or the anion resin waste.

Abstract: A radioactive containment composition may be created for containing radionuclides from a radioactive material by mixing a clay mineral with water. This mixture may form an aqueous clay suspension, which in turn can be refined by filtering to remove coarse material. The aqueous clay suspension may be applied to a radioactive material, allowing the radionuclides to be exchanged with cations in the aqueous clay suspension. The resulting aqueous slurry may be collected, heated and analyzed.

Abstract: A collection process for precipitated powder substance is mainly for storing toxic or highly radioactive powder or dust in air and involves collection operation in water or other liquid to collect powder insoluble to the liquid and heavier than the liquid. To be suitable for underwater operation, a powder precipitation bag is provided with an innovative powder collection can at its bottom. New powder collection can has a mesh filter in its bottom and two-stage structure. After completion of powder collection, the collection can is pulled up from the water and dried to reduce powder volume and form powder lumps. Then, the top half is removed, so the collection can is fully filled with powder lumps. The newly designed powder can increases storage efficiency and reduces waste of space by accommodating more powder lumps than traditional powder can. With the same total amount of powder, the new powder can enables fewer number of storage cans.

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: A method for treating a radioactive liquid waste containing a sodium salt is described, including: a liquid waste reducing step of partially reducing sodium nitrate contained in a radioactive liquid waste to convert the radioactive liquid waste into a reduced solution containing at least one of sodium hydroxide, sodium hydrogencarbonate, and sodium carbonate; and an electrodialysis step of performing electrodialysis of the reduced solution by supplying the reduced solution to an anode chamber of an electrolytic cell, the electrolytic cell including an anode and a cathode disposed on both sides of a permeable membrane which a sodium ion selectively permeates.

Abstract: 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. are decontaminated by non-thermal laser peeling without suffering re-melting, re-diffusing and re-contaminating such that upon non-thermal laser irradiation, areas near the irradiated surface are evaporated and removed faster than the heat generated in the irradiated surface is transmitted to nearby areas.

Abstract: Provided is a method that can remove fission products in a spent electrolyte produced in a dry reprocessing process by an easy operation and can vitrify the fission products easily, the fission products including not only the fission products that generate precipitate but also the fission products that generate no precipitate. A spent electrolyte produced in a dry reprocessing process is subjected to a phosphate conversion processing to obtain a processing target substance; the processing target substance is passed through a separating material 10 including an iron phosphate glass at a temperature of not more than a softening point of the iron phosphate glass in order to remove insoluble fission products included in the processing target substance by filtration with the separating material and to sorb fission products in solution to the separating material for separation; and the iron phosphate glass holding the fission products is used as a waste vitrification material.