Abstract: A method of separating lutetium from a solution containing Lu and Yb, particularly reactor-produced 177Lu and 177Yb, includes the steps of: providing a chromatographic separation apparatus containing LN resin; loading the apparatus with a solution containing Lu and Yb; and eluting the apparatus to chromatographically separate the Lu and the Yb in order to produce high-specific-activity 177Yb.

Abstract: An electrically conductive composite is provided having a mesoporous architecture that improves the accessibility of a nanoscopic catalyst, supported on conductive carbon, to a mass-transported reactant, or substrate, thereby leading to enhanced catalytic activity. In particular, the composite is useful for a new class of fuel-cell electrode architectures based on a composite aerogel that improves the accessibility of a carbon-supported Pt electrocatalyst to methanol (MeOH), leading to higher MeOH oxidation activities than observed at the native carbon supported Pt electrocatalyst. The composite comprises a nanoscopic Pt electrocatalyst, a carbon black electron-conducting support, and a silica aerogel.

Abstract: Inorganic-organic hybrid gels can be employed to extract chemical species such as lanthanides and actinides and their preparation. The gels include a network of inorganic components with the formula: where M is Si, Ti, Zr or Al, wherein organic molecules which can complex the species to be extracted are integrated, with each organic molecule being covalently linked to one or several M atoms in the network. The network is made by a sol-gel process from metallic alkoxides functionalized with complexing groups (amino, ether, hydroxy, amido, pyridino and bipyridino), capable of extracting metals such as lanthanides and actinides.

Abstract: The present invention provides a generator capable of producing therapeutic Bismuth-213 doses. Also disclosed are methods of preparing Bismuth-213-labeled compounds using such generator and applications of the labeled compounds.

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: 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: The invention relates to adsorption material for radionuclides based on a fine fibrous material containing chitin. The material is obtained from fungi and insoluble ferrocyanides of the transition metals, especially copper ferrocyanide. The adsorption material is especially well suited for cleaning liquid radioactive wastes produced in the atomic industry and by nuclear power stations.

Abstract: The present invention relates to solvents, and methods, for selectively extracting and recovering radionuclides, especially cesium and strontium, rare earths and actinides from liquid radioactive wastes. More specifically, the invention relates to extracting agent solvent compositions comprising complex organoboron compounds, substituted polyethylene glycols, and neutral organophosphorus compounds in a diluent. The preferred solvent comprises a chlorinated cobalt dicarbollide, diphenyl-dibutylmethylenecarbamoylphosphine oxide, PEG-400, and a diluent of phenylpolyfluoroalkyl sulfone. The invention also provides a method of using the invention extracting agents to recover cesium, strontium, rare earths and actinides from liquid radioactive waste.

Abstract: The present invention is a method of separating a short-lived daughter isotope from a longer lived parent isotope, with recovery of the parent isotope for further use. Using a system with a bi-directional pump and one or more valves, a solution of the parent isotope is processed to generate two separate solutions, one of which contains the daughter isotope, from which the parent has been removed with a high decontamination factor, and the other solution contains the recovered parent isotope. The process can be repeated on this solution of the parent isotope. The system with the fluid drive and one or more valves is controlled by a program on a microprocessor executing a series of steps to accomplish the operation. In one approach, the cow solution is passed through a separation medium that selectively retains the desired daughter isotope, while the parent isotope and the matrix pass through the medium. After washing this medium, the daughter is released from the separation medium using another solution.

Abstract: The present invention is a method of removing an impurity of plutonium, lead or a combination thereof from a mixture of radionuclides that contains the impurity and at least one parent radionuclide. The method has the steps of (a) insuring that the mixture is a hydrochloric acid mixture; (b) oxidizing the acidic mixture and specifically oxidizing the impurity to its highest oxidation state; and (c) passing the oxidized mixture through a chloride form anion exchange column whereupon the oxidized impurity absorbs to the chloride form anion exchange column and the 22.sup.9 Th or 2.sup.27 Ac "cow" radionuclide passes through the chloride form anion exchange column. The plutonium is removed for the purpose of obtaining other alpha emitting radionuclides in a highly purified form suitable for medical therapy.

Abstract: A process for preconcentrating and separating radium from a contaminated solution containing at least water and radium includes the steps of adding a quantity of a water-soluble macrocyclic polyether to the contaminated solution to form a combined solution. An acid is added to the combined solution to form an acidic combined solution having an ?H.sup.+ ! concentration of about 0.5M. The acidic combined solution is contacted with a sulfonic acid-based strong acid cation exchange medium or a organophilic sulfonic acid medium having a plurality of binding sites thereon to bind the radium thereto and to form a radium-depleted solution. The radium-depleted solution is separated from the strong acid cation exchange medium or organophilic sulfonic acid medium. The radium remaining bound to the exchange medium or organophilic reagent is then stripped from the exchange medium or organophilic medium and the activity of the radium is measured.

Abstract: A process for producing substantially impurity-free Bi-213 cations is disclosed. An aqueous acid feed solution containing Ac-225 cations is contacted with an ion exchange medium to bind the Ac-225 cations and form an Ac-225-laden ion exchange medium. The bound Ac-225 incubates on the ion exchange medium to form Bi-213 cations by radioactive decay. The Bi-213 cations are then recovered from the Ac-225-laden ion exchange medium to form a substantially impurity-free aqueous Bi-213 cation acid solution. An apparatus for carrying out this process is also disclosed.

Abstract: A process for the decontamination of radioactive materials which process comprises the steps of: i) contacting the material to be decontaminated with a dilute carbonate containing solution in the presence of ion exchange particles which either contains or have a chelating function bond to them; and ii) separating the ion exchange particles from the dilute carbonate containing solution. The radioactive materials which are treated may be natural materials, such as soil, or man-made materials such as concrete or steel, which have been subjected to contamination.

Abstract: A cation exchanger is loaded only with radioactive cations by repeatedly converting the non-bivalent, inactive cations into anionic complexes and flushing the complexes out of the cation exchanger. The cation exchanger, which may be integrated in a primary system of a nuclear power plant, is connected to a feed line for a complexing agent and to a discharge line through which the anionic complexes are flushed from the exchanger.

Abstract: The present invention is a method of obtaining a radionuclide product selected from the group consisting of .sup.223 Ra and .sup.225 Ac, from a radionuclide "cow" of .sup.227 Ac or .sup.229 Th respectively. The method comprises the steps of a) permitting ingrowth of at least one radionuclide daughter from said radionuclide "cow" forming an ingrown mixture; b) insuring that the ingrown mixture is a nitric acid ingrown mixture; c) passing the nitric acid ingrown mixture through a first nitrate form ion exchange column which permits separating the "cow" from at least one radionuclide daughter; d) insuring that the at least one radionuclide daughter contains the radionuclide product; e) passing the at least one radionuclide daughter through a second ion exchange column and separating the at least one radionuclide daughter from the radionuclide product and f) recycling the at least one radionuclide daughter by adding it to the "cow". In one embodiment the radionuclide "cow" is the .sup.

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: The invention concerns the recovery of Pu (IV) by using crown compounds.According to the invention, the aqueous solution containing the plutonium (IV) is placed in contact with at least one crown compound, for example DCH 18C6, dissolved in an organic diluting agent, such as benzonitrile. In the case where the solution only contains traces of Pu, it is possible to use a crown compound secured to a solid phase, such as silica.The aqueous solution is a concentrated solution of fission products, an effluent originating from an irradiated nuclear fuel reprocessing installation or a concentrated solution of plutonium with americium.

Abstract: A method for removing uranium from uranium contaminated water involves adding uranium reducing microorganisms and an electron donor to uranium contaminated water to reduce U(VI) in the uranium contaminated water to U(IV) which in turn forms a UO.sub.2 precipitate The UO.sub.2 precipitate is then separated from the contaminated water.

Abstract: A process for decontaminating radioactive material comprises the step of contacting the material with a dissolving composition to dissolve the contaminants in the material, said composition comprising a dilute solution of about 0.05 molar ethylene diamine tetraacetic acid, about 0.1 molar carbonate, about 10 grams per liter hydrogen peroxide and an effective amount of sodium hydroxide to adjust the pH of the composition to a pH from about 9 to about 11. Also included are the steps of separating the dissolving composition containing the dissolved contaminants from the contacted material and recovering dissolved contaminants from the dissolving composition that has been separated from the material. A composition for dissolving radioactive contaminants in a material, comprising a dilute solution having a basic pH and effective amounts of a chelating agent and a carbonate sufficient to dissolve radioactive contaminants is also provided.

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: An apparatus for providing an inventory of an isotopically enriched acidic solution of an element that is capable of forming a water soluble weakly acidic solution at temperatures amenable to ion exchange and having a lower molecular weight isotope fraction and a higher molecular weight isotope fraction is disclosed. This apparatus includes at least two ion exchangers, each being temperature dependent with respect to their storage capacity of said lower and higher molecular weight isotopes, a hot tank containing a hot tank solution and a cold tank containing a cold tank solution both flow coupled to the ion exchangers. A first series of valves is selectively operable to direct the hot tank solution to one of the ion exchanger and to direct the cold tank solution to the other ion exchangers. A second series of valves is selectively operable to direct the flow from the ion exchangers to either the hot tank or the cold tank. An analyzer for sampling the solution is located upstream from the hot tank.

Abstract: A process for the treatment of a material which is or is suspected to contain or carry one or more actinides or compounds thereof to dissolve such actinides or compounds comprises contacting the material with an aqueous solution having a pH in the range 5.5 to 10.5 which is free of heavy metal ions and comprises ingredients which are naturally degradable to non-toxic products with or without mild physical assistance such as heat or ultra-violet radiation, said solution comprising:(a) carbonated water;(b) a conditioning agent;and (c) a complexing agent which comprises the anion of a carboxylic acid having from 2 to 6 carbon atoms.The process may be employed to separate spent nuclear fuel from its metal containment or it may be employed to decontaminate surface, e.g. concrete or soil or pipes carrying traces of actinides, or bulk materials such as soil or rubble.

Abstract: A process and an apparatus are provided which separate the lower molecular weight isotope fraction from the higher molecular weight isotope fraction of an element which forms a water soluble acid at temperatures amenable to ion exchange. The separation is accomplished by shifting the equilibrium constants for absorption and desorption by temperature variations of the acidic solution. The process and apparatus are preferably employed to enrich a boric acid solution in its boron-10 isotope fraction by employing the use of ion exchange resins which are capable of thermally storing and releasing boron isotopes.

Abstract: A method of isotope enrichment including providing a reactant. The reactant is a mixture of at least two isotopic compounds. One of the isotopic compounds has a greater percentage of a selected isotope than the other. An electron transfer agent is introduced into the reactant isotopic mixture to form a reaction mixture. The electron transfer agent facilitates transfer of electrons interactive with one or more selected nuclei of the reactant. The electron transfer agent is in an amount sufficient upon reaction of the reaction mixture to establish an equilibrium between an electron rich reaction product and an electron poor reaction product. The reaction mixture is reacted until the equilibrium is established. The electron poor reaction product is then separated from the electron rich reaction product.

Abstract: A method and a generator for separating daughter radioisotope from a stock solution containing the daughter radioisotope and its parent are provided. The generator is provided with a glass adsorbent which preferentially adsorbs daughter radioisotope relative to the parent radioisotope and circulation means for withdrawing the stock solution from a reservoir, contacting the solution with the glass adsorbent to adsorb the daughter radioisotope and returning the parent radioisotope-enriched solution to the reservoir. In the process stock solution containing daughter radioisotope having a predetermined pH is contacted with the glass adsorbent to selectively adsorb the daughter radioisotope thereon and the daughter radioisotope is eluted with an eluant at a second predetermined pH. The process also includes a step of washing the glass adsorbent with a wash solution at the first predetermined pH to remove traces of the parent radioisotope from the adsorbent prior to eluting the daughter radioiotope.

Abstract: In accordance with the present invention, there is described a method and apparatus for concentrating dissolved and solid radioactive materials carried in a waste water solution containing a hazardous chelating agent used for cleaning nuclear equipment. An oxidizing stage for receiving the waste water containing the radioactive materials and the hazardous chelating agent in the presence of an oxidizing agent oxidizes the chelating agent into non-hazardous constituents including gas and water. A separator coupled to the oxidizing chamber receives the waste water containing the radioactive material and separates the radioactive solids from the waste water containing dissolved radioactive materials. An ion exchange chamber containing an ion exchange resin receives the waste water containing the dissolved radioactive materials and removes the same from the waste water by ion exchange with the resin.

Abstract: There is disclosed a waste treatment process for alkaline waste liquid in which substances to be removed exist in ionic and colloidal states. In the process, powdered tannin is added to the waste liquid to produce solid substance of the tannin and to capture the substances to be removed on the solid substance. Then, the waste liquid is subjected to filtration to thereby separate the solid substance including the substances to be removed.

Abstract: A novel bioadsorption composition suitable for removing heavy metal from waster water effluent, the composition comprising a biomass encapsulated sol-gel matrix. A process for preparing the biomass encapsulated sol-gel matrix is also provided. The bioadsorption composition may be suitably used to remove a substantial amount heavy metal (such as uranium) from a waste water effluent, particularly a dilute aqueous stream comprising a waste water effluent (such as mine water). Heavy metal may then be recovered from the bioadsorption composition, thereby rendering the latter as reusable.

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 improved method and apparatus are capable of efficient removal of radioactive europium from solutions of radioactive gadolinium in a simple way. A mixture of a zinc and a graphite powder is packed into a column and both a conditioning solution corresponding to a liquid electrolyte and a sample solution containing radioactive gadolinium and europium are allowed to pass through the column.

Abstract: An ion exchange material comprises an oxide of a first element selected from elements of Groups IVa, IVb, Va, Vb, VIa, VIb, VIIa, or lanthanide or actinide series of the Periodic Table, in combination with an oxide of at least one different element selected from elements of the afore-mentioned groups of the Periodic Table so as to form a composite material. For example, a composite ion exchange material may comprise silicon dioxide and manganese dioxide. By selecting the appropriate oxides a composite ion exchange material can be prepared which has greater mechanical and/or chemical stability than an individual oxide of the composite material.

Abstract: Ion exchange performed by contacting a solution containing ions to be absorbed with a hydrous oxide of uranium in a suitable form. The hydrous oxide may be uranium dioxide or uranium hydroxide, and may be prepared by reducing a uranium species, precipitating the reduced species, and drying the precipitate to give the required hydrous oxide.

Abstract: Process for the purification of reprocessing uranium from which have previously been separated the fission products generated in a nuclear reactor and consisting of eliminating the U232 daughter products appearing during storage, by passing said uranium in its hexafluoride form through a chemically inert porous material.

Abstract: Ruthenium in aqueous solution in a first, oxidizable oxidation state (e.g. as RuNO(NO.sub.3).sub.3) is converted to an insoluble form in a second, different oxidation state (e.g. as RuO.sub.2.nH.sub.2 O) by establishing an electrochemical cell wherein the solution is the electrolyte and electrochemically oxidizing and reducing the ruthenium in the cell. The insoluble form may be filtered from the liquid. The ruthenium treatment may be a stage in the removal of radioactive species from liquids such is in the treatment of medium and low level activity liquid waste streams, wherein actinides are precipitated and filtered off either before or after ruthenium treatment. Subsequently, residual activity may be removed from the stream by either or both of (a) absorption, followed by filtration and electro-osmotic dewatering and (b) electrochemical ion exchange. Filtration fluxes may be maintained by direct electrochemical membrane cleaning.

Abstract: This invention can treat a uranium-containing solution of high or low concentration and/or the waste generated from uranium conversion processes, etc. It is characterized by the use of acorns, nuts of oak tree, which can be easily obtained in our botanical system. By coprecipitating the uranium or other heavy metal elements with the extract of acorns, this invention can not only recover the uranium of other heavy metal elements but also can reduce them before being discharged to the environmental radioactivity level.

Abstract: A method 20 for decontaminating plastic products and materials which have become radioactively contaminated. The treatment method 20 involves dissolving such plastics in a dissolution tank 28 in an organic solvent and treating the resulting solution by a solvent extraction technique in column 36 to remove particulate and dissolved radioactive contaminants from the plastic. The contaminants can be buried in a low level radioactive waste site and the separated plastic material can be disposed of in a sanitary landfill or recycled into other plastic products.

Abstract: A method 20 for decontaminating specially selected plastic materials comprised of polymers and copolymers of unsaturated organic acids and other specially selected plastic materials, which have become radioactively contaminated, includes dissolving such plastic materials in an aqueous solvent 34 and treating the resulting solution selectively via filtration 36, ion-exchange absorption 40, and precipitation 44 processes to remove particulate and dissolved radioactive contaminants. The treated aqueous stream may be discharged to a sewage stream and the separated plastic materials can be disposed of in a sanitary landfill or recycled into other plastic products.

Abstract: A method of isotope enrichment comprises providing a mixture of at least two isotopic compounds of a reactant. One of the components having a greater percentage of a selected isotope than the other. An electron transfer agent is introduced into the mixture to form a reaction mixture. The electron transfer agent facilitates the transfer of electrons interactive with nuclei of the selected element in the reactant. The transfer agent is in an amount sufficient upon reaction of the reaction mixture to establish an equilibrium between an electron rich reaction product and an electron poor reaction product. The reaction mixture is reacted until the equilibrium is established. Molecules of the electron poor reaction product are then separated from molecules of the electron rich reaction product.

Abstract: The invention relates to a process for improving the effectiveness of the removal of zirconium from an aqueous, zirconium-containing nuclear fuel and/or fertile material solution in nitric acid in a liquid-liquid extraction process.Accordingly, the object of the invention is to improve the removal of zirconium from the reprocessing solutions and, at the same time, to simplify the course of the process. The invention seeks to improve decontamination of the uranium product and the plutonium product while, at the same time, reducing the outlay involved.According to the invention, this object is achieved in that, in a process step carried out before the first extraction of the nuclear fuels and/or fertile materials, the zirconium is converted from the dissolved state into a filterable or centrifugable solid phase by the use of an adsorbent from the group of inorganic ion exchangers and is removed from the aqueous solution together with the adsorbent.

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: This invention relates to the preparation of new, naturally produced chelating agents as well as to the method and resulting chelates of desorbing cultures in a bioavailable form involving Pseudomonas species or other microorganisms. A preferred microorganism is Pseudomonas aeruginosa which forms multiple chelates with thorium in the range of molecular weight 100-1,000 and also forms chelates with uranium of molecular weight in the area of 100-1,000 and 1,000-2,000.

Abstract: In operation of a boiling water-type, atomic power plant, rated operation is conducted while adjusting the pH of core water in the nuclear reactor to 7.0 to 8.5 by adding an alkali thereto, thereby controlling dissolution of radioactive corrosion products attached to fuel rods into the core water. An increase in the radioactive corrosion products in the core water and in the dose rate can be prevented.

Abstract: Synthetic Crandallites of the general composition MeAl.sub.3 (PO.sub.4) (OH).sub.5.H.sub.2 O (Me=alkaline earth metal) are used for the separation of cations, especially alkaline earth metals, lanthanides and actinides, at a pH greater than 6 and temperature above 40.degree. C.

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: A liquid containing radioactive ions is purified (decontaminated) by contacting the same with an inorganic ion exchange composition having ion exchange sites which can be occupied by the radioactive ions from the liquid. The ion exchange composition is a mixture of an ion exchange medium and an additive which is relatively inert to the ion exchange process and which is a sintering aid for the ion exchange medium designed to lower the sintering temperature of the ion exchange composition. The ion exchange composition may be disposed within a suitable container (e.g., cannister), e.g., made of 304L stainless steel or Inconel 601 and the ion exchange process may be carried out in such container. Alternatively, the ion exchange medium can be employed without being previously admixed with the additive. The additive, if desired, can be admixed at a later stage with the contaminated medium.

Abstract: Carbon dioxide is selectively removed from an acid gas stream containing carbon dioxide and sulfur-containing components including hydrogen sulfide by counter-currently contacting the acid gas stream with an aqueous solution of a treating agent in order to selectively absorb carbon dioxide from the gas stream, the treating agent being an aqueous solution of a polyoxypropylene triamine having the formula: ##STR1## wherein R' represents a methylene group and R" represents hydrogen or methyl or ethyl and wherein the sum of x+y+z is a positive integer having a value of from about 4 to about 6.

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: Hydrazine is added to low-level radioactive waste, and the waste is contacted with an iron hydroxide-cation exchange resin so that its radioactive concentration may be lowered. The resin is a strongly acid cation exchange resin treated with ferric chloride and aqueous ammonia and containing a product of hydrolysis of ferric ions.

Abstract: Hydroxylapatite, a compound having the formulaM.sub.10 (PO.sub.4).sub.6 (OH).sub.2,wherein M is calcium, strontium, barium, lead, iron, sodium, potassium, zinc, cadmium, magnesium, aluminum or a rare earth metal, as a support medium for strontium-82 in a strontium-82/rubidium-82 parent-daughter radionuclide generator.