Abstract: According to an embodiment, a nuclear fuel material recovery method of recovering a nuclear fuel material containing thorium metal by reprocessing an oxide of a nuclear fuel material containing thorium oxide in a spent fuel is provided. The method has: a first electrolytic reduction step of electrolytically reducing thorium oxide in a first molten salt of alkaline-earth metal halide; a first reduction product washing step of washing a reduction product; and a main electrolytic separation step of separating the reduction product. The first molten salt further contains alkali metal halide, and contains at least one out of a group consisting of calcium chloride, magnesium chloride, calcium fluoride and magnesium fluoride. The method may further has a second electrolytic reduction step of electrolytically reducing uranium oxide, plutonium oxide, and minor actinoid oxide in a second molten salt of alkali metal halide.

Abstract: Various embodiments provide a process roasting a metal bearing material under oxidizing conditions to produce an oxidized metal bearing material, roasting the oxidized metal bearing material under reducing conditions to produce a roasted metal bearing material, leaching the roasted metal hearing material in a basic medium to yield a pregnant leach solution, conditioning the pregnant leach solution to thrill a preprocessed metal bearing material; and leaching the preprocessed metal bearing material in acid medium.

Abstract: Methods for the extraction of metals such as rare earth metals and thorium from metal compounds and solutions. The methods may include the selective precipitation of rare earth elements from pregnant liquor solutions as rare earth oxalates. The rare earth oxalates are converted to rare earth carbonates in a metathesis reaction before being digested in an acid and treated for the extraction of thorium. A two-step extraction method may be applied to precipitate thorium as thorium hydroxide under controlled pH conditions such that pure thorium precipitate is recovered from a first step and a thorium-free rare earth solution is recovered at the subsequent step. The resulting rare earth solutions are of extremely high purity and may be processed directly in a solvent extraction circuit for the separation of rare earth elements, or may be processed for the direct production of a 99.9% bulk rare earth hydroxide/oxide concentrate.

Abstract: In a preferred embodiment, a process for extracting uranium from wet-process phosphoric acid (WPA), comprises separating uranium from WPA to produce a loaded uranium solution stream and a uranium depleted WPA stream. The loaded uranium solution stream is then contacted by with an ion exchange resin. Uranium species bound to the ion exchange resin are eluted by contacting the resin with a solution comprising anions to produce a loaded uranium eluant stream. The loaded uranium eluant stream is treated to provide a uranium containing product.

Abstract: The invention relates to a process which makes it possible to separate together all the actinide(III), (IV), (V) and (VI) entities present in a highly acidic aqueous phase from fission products, in particular lanthanides, also present in this phase by using a solvating extractant in a salting-out medium. Applications: reprocessing of irradiated nuclear fuels, in particular for recovering plutonium, neptunium, americium, curium and possibly uranium, present in the form of traces, in a pooled but selective fashion with regard to lanthanides, from a solution for the dissolution of an irradiated nuclear fuel, downstream of a cycle for the extraction of uranium.

Abstract: A method for forming nanoparticles containing uranium oxide is described. The method includes combining a uranium-containing feedstock with an ionic liquid to form a mixture and holding the mixture at an elevated temperature for a period of time to form the product nanoparticles. The method can be carried out at low temperatures, for instance less than about 300° C.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: A process for the extraction of uranium compounds from wet-process phosphoric acid includes lowering the iron concentration of the wet-process phosphoric acid and reducing the valency of any remaining ferric iron in the wet-process phosphoric acid to ferrous iron, and then extracting uranium compounds from the wet-process phosphoric acid. The process can include separating a side stream from a feed stream of wet-process phosphoric acid, wherein the side stream has a greater concentration of the uranium compounds than the feed stream by filtration. Extracting uranium compounds from the wet-process phosphoric acid can be by ion exchange process or by solvent extraction.

Abstract: The invention deals with a method for precipitating at least one solute in a reactor comprising: a) a step in which a first liquid phase comprising the solute and a second liquid phase comprising a solute precipitation reagent are brought into contact in co-current in a reactor, as a result of which an emulsion mix is obtained comprising precipitate particles in suspension, and a third liquid phase forming a dispersing phase for said emulsion mix; and b) a step in which the mix mentioned in step a) is fluidised by the third phase.

Abstract: Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: Compositions are provided that include nuclear fuel. Methods for treating nuclear fuel are provided which can include exposing the fuel to a carbonate-peroxide solution. Methods can also include exposing the fuel to an ammonium solution. Methods for acquiring molybdenum from a uranium comprising material are provided.

Abstract: Method of producing anhydrous thorium(IV) tetrahalide complexes, utilizing Th(NO3)4(H2O)x, where x is at least 4, as a reagent; method of producing thorium-containing complexes utilizing ThCl4(DME)2 as a precursor; method of producing purified ThCl4(ligand)x compounds, where x is from 2 to 9; and novel compounds having the structures:

Abstract: Embodiments of the invention provide a method of making non-spherical nanoparticles that includes (a) combining a source of a Group 12, 13, 14, or 15 metal or metalloid; a source of a Group 15 or 16 element; and a source of a quaternary ammonium compound or phosphonium compound; and (b) isolating non-spherical nanoparticles from the resulting reaction mixture. Other embodiments of the invention provide non-spherical nanoparticle compositions, that are the reaction product of a source of a Group 12, 13, 14, or 15 metal or metalloid; a source of a Group 15 or 16 element; and a source of a quaternary ammonium compound or phosphonium compound; wherein nanoparticle tetrapods comprise 75-100 number percent of the nanoparticle products.

Abstract: Cd-112 isotope is recycled from a Cd-112 chemical separated solution or a remainder of an electroplating solution having a Cd-112 target. The present invention recycles Cd-112 isotope with a low cost, a high purity and a high recycle rate. The recycled Cd-112 isotope can be easily stored. And, the Cd-112 isotope can be used as an imaging agent in nuclear medicine.

Abstract: A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450° C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80° C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Abstract: Method for producing a uranium concentrate in the form of solid particles, by precipitation from a uranium-containing solution using a precipitating agent, in a vertical reactor comprising a base, a top, a central part, an upper part, and a lower part, the solid particles of the uranium concentrate forming a fluidized bed under the action of a rising liquid current which circulates from the base towards the top of the reactor successively passing through the lower part, the central part and the upper part of the reactor, and which is created by introducing a liquid recycling current (flow) at the base of the reactor, said liquid recycling current being tapped at a first determined level (A) in the upper part of the reactor and sent back without settling to the base of the reactor, excess liquid being also evacuated via an overflow located at a second determined level (B) in the upper part of the reactor; a method in which the upper limit (C) of the fluidized bed of solid particles is controlled so that it is

Abstract: A separation and recycling method for recycling uranium and fluoride from a waste liquid sequentially and separately is disclosed. The method comprises a uranium-recycling process and a fluoride-recycling process. In the uranium-recycling process, an alkali metal compound or monovalent cation and a coagulant aid are added into the waste liquid to promote the precipitation of uranium. In the fluoride-recycling process, an alkaline earth metal compound, a strong acid and a coagulant aid are added into the uranium-removed waste liquid to precipitate fluoride. By means of the method of the present invention, the uranium and fluoride contents of the uranium-removed and fluoride-removed waste liquid are compliant with the effluent standards of the environmental laws.

Abstract: The present invention relates to a method for the volumetric decontamination of radioactive metals. The method includes the step of precipitating out radioactive gamma and beta emitting nucleotides and then recovering non-radioactive metal compounds.

Abstract: A reuse apparatus of eutectic salt waste produced in an electro refining process and a method thereof is a technology that in order to collect the eutectic salt of the eutectic salt waste, oxidizes/precipitates nuclides (rare earth and TRU) within the eutectic salt waste, an oxygen dispersing method is used to perform a layer separation into the eutectic salt layer and the precipitate layer. Then, the precipitate layer and eutectic salt layer are separated and collected, so that the eutectic salt layer is directly reused and the eutectic salt within the precipitates is reused by separating and collecting it using distillation/condensation processes. The reuse apparatus of the eutectic salt waste and a method thereof thereby increases the collecting efficiency of the eutectic salt and allows the compositions of the collected eutectic salt to have the same compositions as the eutectic salt used in the electro refining process.

Abstract: A separation and recycling method for recycling uranium and fluoride from a waste liquid sequentially and separately is disclosed. The method comprises a uranium-recycling process and a fluoride-recycling process. In the uranium-recycling process, an alkali metal compound or monovalent cation and a coagulant aid are added into the waste liquid to promote the precipitation of uranium. In the fluoride-recycling process, an alkaline earth metal compound, a strong acid and a coagulant aid are added into the uranium-removed waste liquid to precipitate fluoride. By means of the method of the present invention, the uranium and fluoride contents of the uranium-removed and fluoride-removed waste liquid are compliant with the effluent standards of the environmental laws.

Abstract: Method for coprecipitation (or simultaneous precipitation) of at least one actinide in oxidation state (IV) with at least one actinide in oxidation state (III), wherein: a solution i.e. mixture of actinide(s) in oxidation state (IV) and actinide(s) in oxidation state (III) is prepared by adding to it a singly charged cation whose presence makes it possible to stabilize the aforementioned oxidation states in the mixture, or a singly charged cation which does not act to stabilize the aforementioned oxidation states in the mixture; a solution containing oxalate ions is mixed with the said mixture of actinides in order to carry out coprecipitation, i.e. simultaneous precipitation, of the said actinides in oxidation states (IV) and (III) and a fraction of the singly charged cation. According to another embodiment, a solution i.e.

Abstract: An apparatus for the removal of uranium from a body of material is provided. The apparatus has at least one ultrasonic extractor, having a bottom and a top. The at least one ultrasonic extractor is configured to accept solids at the bottom and acid at the top, and has a mixing screw and at least one source of ultrasonic energy. The mixing screw is configured to transport the solids in a direction countercurrent to the acid in the at least one ultrasonic extractor; and the source of ultrasonic energy is configured to impart ultrasonic energy into the solids and the acid, as the solids and the acid traverse the at least one ultrasonic extractor countercurrently.

Abstract: The present invention relates to a process for the recovery of uranium in high silica environments comprising the use of a strong base macroreticular ion exchange resin.

Abstract: The invention relates to a method for separating uranium(VI) from one or more actinides selected from actinides(IV) and actinides(VI) other than uranium(VI), characterized in that it comprises the following steps: a) bringing an organic phase, which is immiscible with water and contains the said uranium and the said actinide or actinides, in contact with an aqueous acidic solution containing at least one lacunary heteropolyanion and, if the said actinide or at least one of the said actinides is an actinide(VI), a reducing agent capable of selectively reducing this actinide(VI); and b) separating the said organic phase from the said aqueous solution. Applications: reprocessing irradiated nuclear fuels, processing rare-earth, thorium and/or uranium ores.

Abstract: A method of separating isotopes from a mixture containing at least two isotopes in a solution is disclosed. A first isotope is precipitated and is collected from the solution. A daughter isotope is generated and collected from the first isotope. The invention includes a method of producing an actinium-225/bismuth-213 product from a material containing thorium-229 and thorium-232. A solution is formed containing nitric acid and the material containing thorium-229 and thorium-232, and iodate is added to form a thorium iodate precipitate. A supernatant is separated from the thorium iodate precipitate and a second volume of nitric acid is added to the thorium iodate precipitate. The thorium iodate precipitate is stored and a decay product comprising actinium-225 and bismuth-213 is generated in the second volume of nitric acid, which is then separated from the thorium iodate precipitate, filtered, and treated using at least one chromatographic procedure.

Abstract: A method is provided for removing uranium from water. The method includes the mixing of a boron reagent with water contaminated with uranyl dication ions, leading to removal of the uranium from that water.

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: A two-cycle countercurrent extraction process for recovery of highly pure uranium from fertilizer grade weak phosphoric acid. The proposed process uses selective extraction using di-(2-ethyl hexyl) phosphoric acid (D2EHPA) and tri-n-butyl phosphate (TBP) with refined kerosene as synergistic extractant system on hydrogen peroxide treated phosphoric acid, and stripping the loaded extract with strong phosphoric acid containing metallic iron to lower redox potential. The loaded-stripped acid is diluted with water back to weak phosphoric acid state and its redox potential raised by adding hydrogen peroxide and re-extracted with same extractant system. This extract is first scrubbed with sulfuric acid and then stripped with alkali carbonate separating iron as a precipitate, treated with sodium hydroxide precipitating sodium uranate, which is re-dissolved in sulfuric acid and converted with hydrogen peroxide to highly pure yellow cake of uranium peroxide.

Abstract: The present invention generally relates to the preparation of mixed actinide oxides, such as mixed oxides of uranium and plutonium (U, Pu) O2, by simultaneously coprecipitation and then calcinations.

Abstract: A multicolumn selectivity inversion generator separation method has been developed in which actinium ions, a desired daughter radionuclide, are selectively extracted from a solution of the thorium parent and daughter radionuclides by a primary separation column, stripped, and passed through a second guard column that retains any parent or other daughter interferents, while the desired daughter actinium ions and radium ions elute. This separation method minimizes the effects of radiation damage to the separation material and permits the reliable production of radionuclides of high chemical and radionuclidic purity for use in diagnostic or therapeutic nuclear medicine.

Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: A treatment for waste pickling solutions containing iron and method of ferric oxide formation, consisting of spent acid with iron content, the addition therein of sodium hydroxide to adjust the pH value, the execution of a chemical reaction in another tank into which sodium hydroxide and air are added while the admixture is exposed to an ultraviolet beam circuit in a photo-oxidation process, and finally magnetic culling to separate ferric oxide in the solution.

Abstract: The invention includes a method of separating isotopes from a mixture containing at least two isotopes in a solution. A first isotope is precipitated and is collected from the solution. A daughter isotope is generated and collected from the first isotope. The invention includes a method of producing an actinium-225/bismuth-213 product from a material containing thorium-229 and thorium-232. A solution is formed containing nitric acid and the material and iodate is added to form a thorium iodate precipitate. A supernatant is separated from the thorium iodate precipitate and a second volume of nitric acid is added to the precipitate. The precipitate is stored and a decay product comprising actinium-225 and bismuth-213 is generated in the second volume of nitric acid which is then separated from the thorium iodate precipitate, filtered, and treated using at least one chromatographic procedure. The invention also includes a system for producing an actinium-225/bismuth-213 product.

Abstract: Metal ions are removed from solid surfaces which may be contaminated with one or more radionuclides by contacting the solid surfaces with the supercritical fluid, as, for instance, carbon dioxide containing both an acidic ligand and organic amine. The metal ions are extracted from the solid surface and the extract is separated from the solid surface.

Abstract: An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (“PYRUC”) shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Abstract: A gas-free system for separating a solution of substantially impurity-free daughter products from an associated parent load solution includes a pump, a plurality of multi-port valves, separation medium and a processor. An uncoiled conduit extends between a third port of a second multi-port valve and a first multi-port valve. A processor is operably coupled to a pump, and the plurality of multi-port valves. A method for separating a solution of substantially impurity-free daughter product from an associated parent load solution is also disclosed.

Abstract: The present invention relates to a method for the continuous separation of caesium, strontium and transuranium elements contained in sodium waste which comprises the use of NaTPB, and to a device for the implementation of this method. The method of the invention comprises, in line, the following steps: (a) filling at least one of at least two feed tanks with the waste; (b) analysis of the content of Cs+, Sr++, Na+ and transuranium elements in the waste; (c) pre-treatment, adapted in relation to analysis results, of the solution in the feed tank intended to insolubilise the strontium and transuranium elements; (e1) a first caesium separation treatment; and e1a) a second caesium separation treatment, the method being conducted in continuous manner by means of the alternate use of the feed tanks.

Abstract: A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Abstract: The present invention relates to a method for the continuous separation of caesium, strontium and transuranium elements contained in sodium waste which comprises the use of NaTPB, and to a device for the implementation of this method.

Abstract: A method and apparatus for removing metal contaminants from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents may be added to precipitate the metal oxide and/or the metal as either metal oxide, metal hydroxide, or as a salt. The precipitated materials are filtered, dried and packaged for disposal as waste or can be immobilized as ceramic pellets. More than about 90% of the metals and mineral residues (ashes) present are removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be spray-dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 1.0 ppm of contaminants.

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 removing dissolved uranium from water is provided. The process basically comprises (a) mixing phosphoric acid or particulate bone ash with the water, (b) mixing calcium hydroxide with the mixture produced in step (a) to thereby form calcium hydroxy phosphate or calcium hydroxy apatite which reacts with and complexes at least a portion of the uranium in the water to form a precipitate thereof, and (c) separating the precipitate from the water.

Abstract: A process for converting UF6 to a solid uranium compound such as UO2 and CaF. The UF6 vapor form is contacted with an aqueous solution of NH4OH at a pH greater than 7 to precipitate at least some solid uranium values as a solid leaving an aqueous solution containing NH4OH and NH4F and remaining uranium values. The solid uranium values are separated from the aqueous solution of NH4OH and NH4F and remaining uranium values which is then diluted with additional water precipitating more uranium values as a solid leaving trace quantities of uranium in a dilute aqueous solution. The dilute aqueous solution is contacted with an ion-exchange resin to remove substantially all the uranium values from the dilute aqueous solution. The dilute solution being contacted with Ca(OH)2 to precipitate CaF2 leaving dilute NH4OH.

Abstract: The steps for recovering uranium and transuranic elements are simplified, and the generation of waste solvent and waste materials is suppressed. Spent nuclear fuel is dissolved in nitric acid (S100) and the resulting solution is subjected to electrolytic oxidation so that U, Np, Pu, Am is oxidized to VI using Ce as oxidation catalyst. The solution is cooled, and nitrates of valence VI thereby deposit as crystals and are separated from the mother liquor (S104). The mother liquor is heated and concentrated (S114). The mixed crystalline deposit is dissolved in nitric acid (S106), uranyl nitrate is deposited alone by cooling (S108), and the crystals are separated from the U, Np, Pu, Am mixed solution (S110). The uranyl nitrate is dissolved in nitric acid (S112), and the heated and concentrated mother liquor is added to it to prepare another mixed solution.

Abstract: Thermodynamically-unstable complexing agents which are diphosphonic acids and diphosphonic acid derivatives (or sulphur containing analogs), like carboxyhydroxymethanediphosphonic acid and vinylidene-1,1-diphosphonic acid, are capable of complexing with metal ions, and especially metal ions in the II, III, IV, V and VI oxidation states, to form stable, water-soluble metal ion complexes in moderately alkaline to highly-acidic media. However, the complexing agents can be decomposed, under mild conditions, into non-organic compounds which, for many purposes are environmentally-nondamaging compounds thereby degrading the complex and releasing the metal ion for disposal or recovery. Uses for such complexing agents as well as methods for their manufacture are also described.

Abstract: The present invention describes a process to reduce the volume and/or weight of magnesium slag when the magnesium slag contains radioactive thorium. The process contacts the magnesium slag as an aqueous slurry with an acid in a pH range from about 4.0 to about 8.0, preferably from about 5.0 to about 5.5, followed by separating insoluble solids from the aqueous solution. Optionally, the acid digested solids are heated, either before or after the acid digestion, at a temperature from about 350.degree. to about 500.degree. C. The solid waste can then be further compacted, if desired, prior to disposal.

Abstract: A method of separating exothermic elements of Cs and Sr from a high-level radioactive liquid waste. The method comprises a denitration step wherein formic acid is added to the high-level liquid waste so as to adjust its pH to about 5, thereby causing most of the elements other than Cs and Sr to precipitate in the high-level liquid waste to obtain a denitrated liquid waste containing Cs and Sr in high concentrations. The denitrated liquid waste is then subjected to a pH adjustment step wherein ammonia is added to the denitrated liquid waste so as to adjust its pH to about 7.5 to 9, thereby removing by precipitation the elements other than Cs and Sr remaining in the denitrated liquid waste. When the thus resulting precipitate freed of Cs and Sr is vitrified, the waste content in the vitrified waste can be increased.

Abstract: The process for separating U or Th values from values of at least one other metal such as Ta, Nb, Sc, Zr, Ti or Mg, wherein all of the values are contained in a difficult soluble matrix of highly fluorinated feed materials such as U or Th bearing ores or the processing products, processing intermediates, or processing residues thereof, the method comprising the steps of:(a) digesting the feed materials with an aqueous, mineral acid, digest medium containing one or more complexing materials such as H.sub.3 BO.sub.3, or ions of group 2A, 3A, 4A, 5A, 6A, or 7A metals, Fe, B, Al or Si;(b) maintaining the temperature of the medium between about 55.degree. C. and about 85.degree. C.

Abstract: In this invention, a novel process for saline waters and saline solutions conversion has been provided that requires only a fair amount of a miscible organic solvent and heat transfer. Such requirements are ordinary in the nature of precipitation and vaporization. The invented process consists of adding a miscible (strongly associated) organic solvent to saline water so that salt precipitates of the saline water is formed. The resultant salt precipitates (pure solids) is then separated from the organic-water mixture. After separating the salt precipitates, the miscible organic solvent is removed and recovered from the organic-water mixture by applying vacuum with or without heating, or by using distillation methods. The separated miscible organic solvent can then be condensed and returned to the process and water is stripped of trace of miscible organic solvent, and removed from the system as product water.