Abstract: Trivalent actinides can be separated from trivalent lanthanides in aqueous solutions at H.sup.+ concentrations of 2 mol/l to 0.001 mol/liter by extraction with bis(aryl)dithiophosphinic acid and a synergist such as TBP, TOPO and TBPO in an organic solvent. A high separation efficiency results. The method is applicable to high-level liquid waste from reactors and nuclear-material processing plants.

Abstract: The present invention relates to methods for digesting diphosphonic acid substituted cation exchange resins that have become loaded with actinides, rare earth metals, or heavy metals, in a way that allows for downstream chromatographic analysis of the adsorbed species without damage to or inadequate elution from the downstream chromatographic resins. The methods of the present invention involve contacting the loaded diphosphonic acid resin with concentrated oxidizing acid in a closed vessel, and irradiating this mixture with microwave radiation. This efficiently increases the temperature of the mixture to a level suitable for digestion of the resin without the use of dehydrating acids that can damage downstream analytical resins. In order to ensure more complete digestion, the irradiated mixture can be mixed with hydrogen peroxide or other oxidant, and reirradiated with microwave radiation.

Abstract: A method of separating and recovering Pu and Np from a Pu- and Np-containing nitric acid solution. The method comprises the steps of subjecting a nitric acid solution containing Pu and Np to valence adjustment by adding a reducing agent consisting of hydroxylamine nitrate and hydrazlne to said nitric acid solution so as to reduce Pu and Np in said nitric acid solution to Pu (III) and Np (IV), respectively; adjusting a nitric acid concentration of said nitric acid solution after said valence adjustment to 6 to 8 M; bringing said nitric acid solution after said nitric acid concentration adjustment into contact with a strong basic anion exchange resin so as to cause Np to be selectively adsorbed by said resin and to separate and recover Pu as a plutonium nitrate solution; and eluting said adsorbed Np (IV) by using diluted nitric acid of 1 M or below so as to recover Np as a neptunium nitrate solution.

Abstract: A method of adsorbing and separating a heavy metal element by using a tannin adsorbent comprising:(a) adjusting the pH of a solution containing a plurality of heavy metal elements to a predetermined pH;(b) contacting the adsorbent with the solution in which the pH thereof is adjusted;(c) adjusting the pH of the solution contacted with the adsorbent to a pH different from the predetermined pH; and(d) contacting the solution in which the pH thereof is adjusted at the step (d) with the adsorbent prepared at the step (a).According to the inventive method, all of the heavy metal elements can be efficiently separated and adsorbed by using a tannin adsorbent from a solution containing a number of heavy metal elements. A method of regenerating a tannin adsorbent are also disclosed.

Abstract: A process for recovering metal and acid values from a source material containing metallic fluorides comprises digesting the source material in sulfuric acid to form a slurry, separating a fluoride containing solid phase and a metal containing first liquid phase. The solid phase is subjected to pyrohydrolysis, sulfuric and hydrofluoric acids are recovered, and the first liquid phase is processed to recover the metal values by solvent extraction or ion exchanges. The tantalum values are extracted from the first liquid aqueous phase by a water immiscible organic extractant such as methylisobutyl ketone to form a first liquid organic phase containing tantalum and a second liquid aqueous phase. The tantalum is stripped from the first organic phase using water. The process includes the additional steps of heating the separated solid phase from about ambient temperature to an elevated temperature in the presence of water vapor to evolve sulfuric acid and render the gangue chemically inert.

Abstract: A controlled multi-stage process for the stripping of uranium includes the introduction at one end of a solvent extraction device a uranium loaded organic solution. A concentrated stripping acidic aqueous solution is introduced at the other end of the solvent extraction device so that the aqueous solution and the organic solution are contacted counter-currently in the solvent extraction device at a temperature not substantially exceeding 35.degree. C.

Abstract: The isotopes of boron, .sup.10 B and .sup.11 B, are separated by means of a gas-liquid chemical exchange reaction involving the isotopic equilibrium between gaseous BF.sub.3 and a liquid BF.sub.3 . donor molecular addition complex formed between BF.sub.3 gas and a donor chosen from the group consisting of: nitromethane, acetone, methyl isobutyl ketone, or diisobutyl ketone.

Abstract: A hydrogen storing member which comprises at least two kinds of hydrogen storing materials comprising a first hydrogen storing material and a second hydrogen storing material capable of releasing or storing hydrogen in a range of at least one of temperature and pressure at which the first hydrogen storing material is in the .beta. phase having a high capacity for storing hydrogen.

Abstract: Spent nuclear reactor coolant system decontamination solutions are first processed to remove the radioactive burden and to separate metallic impurities using cation exchange resins. The cation-processed waste solution is then flowed through an anion exchange resin held between anion-permeable membranes within an electrochemical cell. The application of electric current to the electrochemical cell causes hydroxide ions to replace captured anions held on the ion exchange resin, thereby regenerating the anion exchange resin for subsequent processing. The displaced anions migrate into the anode compartment where they can be oxidized to carbon dioxide for simple disposal.

Abstract: A method for removing ruthenium from a uranium-containing solution comprising adding a nitrite to the solution and passing the solution over a quaternary ammonium or tertiary amine resin to selectively fix the ruthenium.

Abstract: A process for cleaning solutions of an alkylphosphate in a hydrophobic organic solvent. The process involves the steps of washing the solution with aqueous sulphuric acid, and contacting the washed solution with an ion exchange material. The ion exchange material may be a resin of the type designated as styrene divinyl benzene copolymers.

Abstract: There is disclosed an insoluble tannin preparation process, in which a tannin is first dissolved in an aldehyde aqueous solution. Then, ammonia is added to the solution to produce a precipitate, and the resulting precipitate is subjected to aging. The insoluble tannin thus prepared is employed to treat a waste liquid. Furthermore, there is disclosed an adsorption process in which a tannin is first dissolved in a solution containing a material to be adsorbed. Thereafter, an aldehyde is added to the solution to produce a precipitate having tannin as a principal constituent. The material to be adsorbed is adsorbed to this precipitate.

Abstract: New diphosphonic derivatives are capable of extracting the uranium in phosphoric acid with a P205 content ranging from 10 to 50 per unit.The new diphosphonic derivatives have the following formula: ##STR1## wherein R is selected from the group consisting of hydrogen and alkyl radicals having 1 to 15 carbon atoms and having at most 2 branchings in the alkyl radical, and A is a mono- or di- substituted methylene radical wherein the substituents are selected from the group consisting of an alkyl group having at most two branchings, hydroxyl, an amino group, and hydrogen, wherein the total number of carbon atoms in the R and A groups is from 15 to 50 carbon atoms, with the proviso that when A is substituted by hydroxy or amino. A has 15 to 25 carbon atoms, and wherein A is linked to a water insoluble crosslinked ion exchange resin skeleton.

Abstract: A simple, low temperature process for separating uranium isotopes and producing substantially pure uranium isotope fractions from a crude uranium isotope-containing aqueous feed stock is provided. A uranium isotope-containing solution is fed to an anion exchange resin in a continuously rotating annular chromatograph as a suitable eluant is also simultaneously fed to the rotating chromatograph. A Uranium 235, a Uranium 238 and an impurity fraction are recovered from the chromatograph. The substantially pure uranium 235 and uranium 238 fractions may be subjected to further processing, preferably precipitation with ammonium hydroxide to produce ammonium diuranate, and then calcining to form uranium oxide rich in the desired isotope.

Abstract: Cations and anions contained as impurities in a spent organic solvent can be removed by an alkaline aqueous solution passing through inside of hydrophobic porous membrane such as hollow fibers for capturing anions and by an acidic aqueous solution passing through inside of hydrophobic porous membrane such as hollow fibers for capturing cations at the same time so as to regenerate the organic solvent. An apparatus suitable for removing cations and anions simultaneously from the spent organic solvent is also provided.

Abstract: The separation of uranium, iron and scandium is achieved by the following process wherein a material containing these values is dissolved in mineral acid to form an aqueous solution, thereafter, an iminodiacetic acid cationic ion exchange resin is used to retain a major portion of the scandium and uranium. A raffinate containing the iron is formed. The ion exchange resin having said scandium and uranium retained thereon is rinsed with a dilute acid to remove residual metals other than scandium and uranium. The resin is eluted with an aqueous solution of an organic chelating acid to remove a major portion of the scandium from said resin and to form an aqueous solution containing scandium. The resin is then eluted with an aqueous solution of a mineral acid to remove a major portion of the uranium from the resin to form an aqueous solution containing the uranium. The scandium and uranium are recovered as solids.

Abstract: A process for the extraction of cations from an aqueous effluent. An effective quantity of at least one polymer or copolymer, which is physically or chemically crosslinked, is introduced into the aqueous effluent. The polymer or copolymer is based on one or more units selected from salified acrylic acid, quarternized acrylic acid, and a copolymer comprising from 40 to 60 mole % of acrylic acid and from 60 to 40 mole % of at least one dialkylaminoalkyl acrylate in which each alkyl group contains from 1 to 4 carbon atoms. This process is used to treat aqueous effluents.

Abstract: The present invention provides a process for producing a uranium oxide by dissolving a yellow cake in sulfuric acid or hydrochloric acid, bringing the obtained solution into contact with a chelating resin of diaminocarboxylic acid type and subjecting the product to neutralizing precipitation followed by heat treatment. By the contact of the solution with the chelating resin, iron, copper, molybdenum and vanadium among the metallic impurities are removed and in the subsequent neutralizing precipitation step, other metallic impurities such as aluminum, calcium, magnesium, sodium and potassium are removed. This process can easily produce uranium oxide having a high purity using a simple apparatus.

Abstract: A process for continuously separating uranium isotopes by an oxidation-reduction reaction using an anion exchanger comprising the steps of feeding an eluant containing a deactivated oxidizing agent and a deactivated reducing agent to an anion exchanger to regenerate them and to form an oxidizing agent zone in the anion exchanger, and reusing the regenerate reducing agent and the oxidizing agent zone for further uranium isotope separation.

Abstract: Chelating agents on which metals such as uranium have been adsorbed are brought into contact with an eluent of aqueous solution containing reducing agents and basic compounds, until the metals are eluted. The elution is effected with large speed without degradation of the chelating agents. Preferred chelating agents are rather stable under basic conditions but not under acidic conditions, for example, those having .dbd.NOH group in their molecules. The reducing agents and basic compounds in the eluent are, for example, sulfurous acid or salts thereof and sodium hydroxide, respectively.

Abstract: A method for eluting a metal adsorbed on a chelating agent with an eluent, which comprises using as the eluent an aqueous solution containing a water-soluble inorganic sulfide and a basic compound at concentrations of 0.005 to 3N and 0.1N or more, respectively.

Abstract: A process for separately recovering uranium and hydrofluoric acid from a waste liquor containing uranium and fluorine comprises a neutralizing precipitation step wherein a magnesium compound is added to the waste liquor containing uranium and fluorine to form a precipitate and the thus formed precipitate is then separated; a distillation step wherein an aqueous solution of sulfuric acid is added to the precipitate separated in said neutralizing precipitation step to dissolve the precipitate and the thus formed solution is then distilled to recover hydrofluoric acid as a distillate; and a uranium recovery step wherein uranium is recovered from a residue produced by said distillation step.

Abstract: A method for a quantitative separation of scandium from thorium comprises adsorption of both metals on a cation exchange resin followed by selective elution of scandium with an acidic solution of a chelating agent followed by the elution of thorium by a six normal hydrochloric acid solution.

Abstract: In a process for continuously separating uranium isotopes, .sup.235 U and .sup.238 U, by oxidation-reduction chromatography using a column of an anion exchanger, the separation of uranium isotopes and the regeneration of the oxidizing agent and reducing agent both deactivated during the separation of uranium isotopes can be simultaneously effected in a single column by returning to and passing through the column of an anion exchanger having a uranium adsorption zone an eluate containing a deactivated oxidizing agent and a deactivated reducing agent to regenerate the deactivated oxidizing agent and the deactivated reducing agent and to form an oxidizing agent zone, and returning to and passing through the column the regenerated reducing agent for the reduction of uranium.

Abstract: An organic extraction phase, containing fission materials uranium and plutonium, as well as a minor amount of fission- and corrosion-products, from an extraction step or from a purification step of a nuclear fuel reprocessing process is treated with an aqueous, basic, carbonate ion containing solution, whereby the fission materials and at least a part of the fission- and corrosion-products are stripped into the aqueous phase and the uranium is present in the form of uranium-carbonato-complex. The organic phase is then separated from the aqueous phase. Plutonium is then removed from the separated aqueous phase. The remaining aqueous solution containing the uranium-carbonato-complex and a small residual amount (C) of fission products is adjusted to a ratio of uranyl ion concentration to carbonate ion- or CO.sub.3.sup.-- /HCO.sub.3.sup.-- concentration of 1(UO.sub.2.sup.++) to 4.5(CO.sub.3.sup.-- or CO.sub.3.sup.-- /HNO.sub.3.sup.-), respectively, or more, at a maximum U-concentration of not more than 6- g/l.

Abstract: A material is, herein disclosed for collecting radionuclides and heavy metals that comprises an acrylic fiber element having MnO.sub.2 and/or ferrocyanide compound of the formula K.sub.2 M.sup.II [Fe(CN).sub.6 ] (wherein M.sup.II is a divalent metal such as cobalt, zirconium, zinc or nickel) supported thereon. The material of the present invention is used in collecting and concentrating radionuclides present in liquid radioactive wastes discharged from nuclear power plants and other facilities using radioisotopes such as hospitals and non-destructive testing laboratories. The material is also used in collecting and concentrating radionuclides and heavy metals present in the seawater, rivers and industrial waste-water.

Abstract: A process for separating large amounts of uranium from small amounts of radioactive fission products, which are present in basic, aqueous carbonate containing solutions, by means of a basic, organic anion exchanger. Uranium values present as uranykl-carbonato complex in a basic, aqueous, carbonate containing solution can be separated from fission products of the group ruthenium, zirconium, niobium and lanthanide, and with a relatively high degree of decontamination as well. The aqueous solution is adjusted to a ratio of uranyl ion concentration to carbonate ion- or CO.sub.3.sup.-- /HCO.sub.3.sup.- concentration of 1(UO.sub.2.sup.++) to 4.5(CO.sub.3.sup.--, or CO.sub.3.sup.-- /HCO.sub.3.sup.-), or more, at a maximum U concentration of not more than 60 g/l. The adjusted solution is led over a basic anion exchanger made from a polyalkene matrix provided with a preponderant part tertiary and a minor part quaternary amino groups to adsorb fission products ions or fission products containing ions.

Abstract: Process for the removal of plutonium polymer and ionic actinides from aqueous solutions by absorption onto a solid extractant loaded on a solid inert support such as polystyrenedivinylbenzene. The absorbed actinides can then be recovered by incineration, by stripping with organic solvents, or by acid digestion. Preferred solid extractants are trioctylphosphine oxide and octylphenyl-N,N-diisobutylcarbamoylmethylphosphine oxide and the like.

Abstract: The invention relates to a process for bituminizing radioactive waste constituted by cation and/or anion exchange resins.This process comprises a resin pretreatment stage performed so as to replace the H.sup.+ ions of the cation exchange resins by ions chosen from the group including Ca.sup.++, Sr.sup.++ and Ba.sup.++ and/or for replacing the OH.sup.- and/or Cl.sup.- ions of the anion exchange resins by an anion chosen from the group including NO.sub.3.sup.-, HCO.sub.2.sup.- and CH.sub.3 CO.sub.2.sup.-.This pretreatment makes it possible to increase the capacity of the coating installation and limit the swelling in water of the bituminized coatings obtained.

Abstract: A novel ion exchange agent comprises a magnetically susceptible core or nucleus surrounded by a coating of plastic or resin ion exchange material, preferably a microreticular coating immediately surrounding the magnetically susceptible core and a macroreticular coating surrounding the microreticular coating and a novel method of preparing the same. A novel ion exchange process is disclosed, including applying a magnetic field to the body of the magnetically susceptible ion exchange agent to at least temporarily immobilize the ion exchange agent while passing liquids therethrough, particularly in the washing or other treatment of ground ores to recover mineral values therefrom in a resin-in-pulp process.

Abstract: A method of improving the recovery of U.sub.3 O.sub.8 from circulating aqueous leach fluids used to extract uranium from its ores in underground uranium fields, which fields are made up of a series of injection and producing wells, which aqueous fluids are contaminated by suspended and dissolved inorganic iron compounds caused by the presence of anerobic and aerobic bacteria, which comprises treating such aqueous leach fluids with at least 0.1 ppm of a mixture of 75% of 5-chloro-2-methyl-4-isothiazolin-3-one and 25% 2-methyl-4-isothiazolin-3-one.

Abstract: Apparatus for maintaining the level of an interface between separate phases in a solvent extraction column by regulating the flow from the base of the column. A steady controllable flow is achieved by passing the flow into a pressure vessel and the level in the vessel is controlled by having a plurality of vertically spaced apart apertures in the flow path within the vessel.

Abstract: A process and apparatus for recovering uranium from a carbonate solution containing uranium ions whereby the carbonate solution containing uranium ions is brought in contact with a cation exchanger so that a uranium cation is removed from solution and adsorbed by the cation exchanger, and the uranium cation is then removed from the cation exchanger. The treated carbonate solution from which uranium ions hve been removed by cation exchange is then further processed by removing carbon dioxide from the treated carbonate solution to produce a decarbonated solution, and passing the decarbonated solution through a membrane process to remove some remaining impurities.

Abstract: The present invention relates to a method for increasing the amount of uranium values loaded on an ion-exchange resin by adding an oxidant, such as hydrogen peroxide, to a uranium containing lixiviant prior to contacting the lixiviant with an ion exchange resin to adsorb the uranium values therein.

Abstract: A process for the recovery of a heavy metal which comprises:bringing a polymer having pendant groups of the formula:--CO--NH--NH--CO--COOMin which M is hydrogen, sodium or potassium, into contact with an aqueous solution containing a heavy metal to have the heavy metal adsorbed by said polymer;andhaving said heavy metal desorbed from the said polymer.

Abstract: Recovery of uranium from crude wet process phosphoric acid (WPA) by cation exchange using as cation exchanger a resin comprising active amino phosphonic or hydroxy phosphonic groups linked to a matrix. The operational steps comprise reduction of uranium in the WPA to the tetravalent state, contacting the WPA with the cation exchange resin to load the latter with tetravalent uranium, displacing Fe.sup.+2 from the resin, washing the resin with aqueous ammonia, eluting the resin with a carbonate-bicarbonate solution and precipitating uranium from the eluate.

Abstract: A method for the in situ leaching of a subterranean mineral (e.g. uranium) deposit which permits the use of downflow ion exchange columns without requiring expensive filtration equipment to remove particulate material and/or sediment from the leach solution. The injection and production wells are gravel packed so that all of the heavier particulate material will be filtered from the leach solution as it flows into the production well. The bed of ion exchange resin in the downflow column is then used to filter the finer sediment from the leach solution as the resin adsorbs the uranium from the leach solution. The barren leach solution from the ion exchange column can be used to makeup fresh leach solution without any further filtration. The loaded resin is removed from the ion exchange columns and is screened to remove the fine sediment. The resin is then eluted and screened again to remove any sediment that may accumulate during elution.

Abstract: A new class of ion exchange agents comprises water soluble polymers called polyelectrolytes sorbed onto ion exchange functional substrates.

Abstract: A process is described for the recovery of uranium from ore wherein the ore is leached, the resulting leachate is passed over an anion exchange resin, and the resin is then eluted with an elution solution containing an anion capable of replacing the uranyl anion captured on the resin, followed by treatment with a sulfite-containing solution so that polythionates captured on the ion exchange resin are reacted to form essentially thiosulfate and trithionate. Preferably, the sulfite solution also contains other anions.

Abstract: Uranium and other base metals are leached from their ores with aqueous solutions containing bicarbonate ions that have been generated or reconstituted by converting other non-bicarbonate anions into bicarbonate ions. The conversion is most conveniently effected by contacting solutions containing SO.sub.4.sup.= and Cl.sup.- ions with a basic anion exchange resin so that the SO.sub.4.sup.= and Cl.sup.- ions are converted into or exchanged for HCO.sub.3.sup.- ions. CO.sub.2 may be dissolved in the solution so it is present during the exchange. The resin is preferably in bicarbonate form prior to contact and CO.sub.2 partial pressure is adjusted so that the resin is not fouled by depositing metal precipitates. In-situ uranium mining is conducted by circulating such solutions through the ore deposit. Oxidizing agents are included in the injected lixiviant.

Abstract: The present invention relates to a process for recovery of uranium from a carbonate lixiviant additionally containing other contaminants such as molybdenum and silica, by adjusting the pH of the lixiviant to a value of at least 4 and treating the lixiviant to obtain a carbonate concentration of at least 50 ppm by weight. Subsequently the lixiviant is treated with an aqueous solution containing a sufficient amount of ferric iron to precipitate the contaminants without substantial precipitation with the uranium values. The precipitate is separated from the lixiviant. The treated lixiviant is passed through an ion exchange to retain uranium values.

Abstract: A process for the separation of actinide ions from aqueous, basic, carbonate containing solutions with the use of basic ion exchangers. The actinide ions are sorbed at a weakly basic anion exchanger comprised of a polyalkene matrix provided in the majority with tertiary and in the minority with quaternary amino groups. The charged ion exchanger is separated from the solution and then the actinide ions are eluted from the exchanger with nitric acid.

Abstract: A method of chemically separating plutonium from thorium. Plutonium and thorium to be separated are dissolved in an aqueous feed solution, preferably as the nitrate salts. The feed solution is acidified and sodium nitrite is added to the solution to adjust the valence of the plutonium to the +4 state. A chloride salt, preferably sodium chloride, is then added to the solution to induce formation of an anionic plutonium chloride complex. The anionic plutonium chloride complex and the thorium in solution are then separated by ion exchange on a strong base anion exchange column.

Abstract: The present invention relates to a process for the recovery of uranium and other minerals from uranium ore where at least part of the uranium is present as refractory uranium-mineral complexes, comprising subjecting the uranium ore to mild oxidative carbonate leach fluid to dissolve and remove readily soluble uranium minerals, subsequently subjecting the uranium ore to an oxidative chemically severe leaching system to dissolve and remove the refractory uranium-mineral complexes, and separating and recovering the uranium and other mineral species in the leachate fluids. The process may be applied to in-situ uranium leaching operations or to surface leaching operations.

Abstract: The density and capacity of macroporous ion exchange resins are increased by incorporating therein at least one of titania hydrate, zirconia hydrate, hydrated titanium phosphorous, and hydrated zirconium phosphate hydrates. Such resin composites are useful, e.g., in removing uranium values from aqueous solutions.

Abstract: A process for recovering soluble uranium with an anion exchange system from carbonate-containing water. The process includes means for overcoming the deleterious effects upon the ion exchange system of carbon dioxide gas resulting from the carbonate-contents of the water in the presence of an acidic elutriate.

Abstract: Purifying a nitric acid solution containing U/Pu ions and contaminated by metal impurities by(a) oxidizing the U/Pu ions of the solution to the hexavalent form,(b) conducting the solution after oxidation through a cation exchanger column in which the impurities, especially americium, are retained,(c) flushing the cation exchanger column to remove impurities therefrom, which may be taken to waste processing or utilization plant.

Abstract: The present invention relates to an improved method for the recovery of uranium values from a uranium-containing lixiviant by passing the lixiviant through a heated ion exchange resin to cause the resin to retain uranium values. The uranium values are subsequently recovered from the ion-exchange resin.

Abstract: In an ion exchange process for recovering valuable components from a weak feed solution containing impurities, an improved loading of the desired component on the resin and a purer and more concentrated eluate can be obtained by subjecting the resin to a conditioning step after the step of exhaustion of the feed solution onto the resin and before the resin elution step. In the conditioning step, a predetermined isolated quantity of the loaded resin is contracted with a conditioning solution capable of preferentially removing the unwanted impurity component.

Abstract: A process as disclosed for recovering uranium values from a carbonate leach solution which comprises directly eluting a column of resin onto which uranium has been sorbed by flowing a concentrated acidic eluant through the column without preconditioning and/or post-conditioning the resin. The concentrated acidic eluant may be flowed upward or, preferably, downward through the column.