229results about "Uranium compounds" patented technology

A perovskite compound of the formula, (Na_1/2Bi_1/2)_1-xM_x(Ti_1-yM′_y)O_3±z, where M is one or more of Ca, Sr, Ba, Pb, Y, La, Pr, Nd, Sm, Eu, Gd, Th, Dy, Ho, Er, Tm, Yb and Lu; and M′ is one or more of Zr, Hf, Sn, Ge, Mg, Zn, Al, Sc, Ga, Nb, Mo, Sb, Ta, W, Cr, Mn, Fe, Co and Ni, and 0.01<x<0.3, and 0.01<y<0.3, and z<0.1 functions as an electromechanically active material. The material may possess electrostrictive or piezoelectric characteristics.

Sodium nonatitanate compositions, a method using the composition for recovery of 82Sr from irradiated targets, and a method using the composition for generating 82Rb. The sodium nonatitanate materials of the invention are highly selective at separating strontium from solutions derived from the dissolution of irradiated target materials, thus reducing target processing times. The compositions also have a very low affinity for rubidium, making it an ideal material for use as a 82Rb generator. Sodium nonatitanate materials of this type both improve the recovery of 82Sr and provide a safer, more effective 82Rb generator system.

Systems and methods are disclosed for producing customized, predictable and reproducible supplies of radioisotopes using, for example, a reactor housing that is fabricated from a radioactive shielding material and has both an internal volume and a surface that comprises an entry port and an exit port, a chromatographic column that is positioned within said internal volume such that a first end of said column is in fluid communication with said entry port and a second end of said column is in fluid communication with said exit port, and a changeable filter module that is disposed external to said reactor housing and in fluid communication with said exit port.

The invention discloses a preparation method of amidoxime group uranium extraction sorbent, comprising the following steps: firstly, adopting polyacrylonitrile powder as a raw material, producing uranium extraction sorbent containing amidoxime group by amidoximation; adding titanium dioxide sol which is prepared by the sol-gel method for blending to ensure that the weight percent of titanium dioxide and polyacrylonitrile is 0.4-2.0%; producing hybridized amidoxime group uranium extraction sorbent by water washing, alcohol washing and drying. The invention has the advantages that the amidoxime group uranium extraction sorbent material which is hybridized with titanium dioxide has higher mechanical strength, higher adsorption quantity of uranium, and higher adsorption selectivity to uranium in the presence of calcium and magnesium ions.

A system for the treatment and recycling of graphite containing radionuclides including a two stage method that employes a thermal roaster that is operatively connected to a steam reformer. In the first stage, radioactive graphite is roasted or heated to volatize a first amount of radionuclides contained in the graphite. In the second stage, the roasted graphite is reacted with steam or gases containing water vapor so that a second amount of radionuclides is removed. Optionally, the present system also processes the radionuclides to enable their disposal.

A method for recovering and recycling catalyst coated fuel cell membranes includes dissolving the used membranes in water and solvent, heating the dissolved membranes under pressure and separating the components. Active membranes are produced from the recycled materials.

The invention relates to a method of synthesizing high-temperature melting materials. More specifically the invention relates to a containerless method of synthesizing very high temperature melting materials such as borides, carbides and transition-metal, lanthanide and actinide oxides, using an Aerodynamic Levitator and a laser. The object of the invention is to provide a method for synthesizing extremely high-temperature melting materials that are otherwise difficult to produce, without the use of containers, allowing the manipulation of the phase (amorphous/crystalline/metastable) and permitting changes of the environment such as different gaseous compositions.

The invention relates to an amidoxime group uranium extraction sorbent and a preparation method thereof. The preparation method comprises the following steps: on the basis of utilizing an organic estersil containing nitrile groups and Si(OR)4 as a mixed silicone source and a nonionic type surface active agent as a structural guide agent, performing hydrolyzation and polymerization under an acidic condition to obtain a nitrile group functional organic-inorganic hybridized porous material; and reacting the nitrile group functional organic-inorganic hybridized porous material with hydroxylamine to obtain the amidoxime group uranium extraction sorbent. In the invention, cyanogroup is introduced into an inorganic material by virtue of hydrolytic condensation reaction, and the amidoxime group functional organic-inorganic hybridized porous material is prepared by virtue of amidoximation. Compared with the polymer material, the sorbent related by the invention has the advantages of obviously increased mechanical strength, strengthened hydrophilism and high uranium adsorption quantity; the amidoxime group functional organic-inorganic hybridized porous material has quite high selection to uranium almost without the interference of other coexisting ions in the solution containing Na<+>, K<+>, Ca<2+> and Mg<2+>.

The present invention relates generally to a process for recovering copper and other metal values from metal-containing materials using controlled, super-fine grinding and medium temperature pressure leaching. Processes embodying aspects of the present invention may be beneficial for recovering a variety of metals such as copper, gold, silver, nickel, cobalt, molybdenum, rhenium, zinc, uranium, and platinum group metals, from metal-bearing materials, and find particular utility in connection with the extraction of copper from copper sulfide ores and concentrates.

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.

The present invention relates to an MRI contrast agent that includes zinc-containing water-soluble metal oxide nanoparticles and has an improved contrast effect. The zinc-containing water-soluble metal oxide nanoparticles are characterized by addition of zinc to a matrix comprising the metal oxide nanoparticles or by substitution of metal in the matrix, resulting in the improved contrast effect of MRI. In addition, the zinc-containing metal oxide nanoparticles of the present invention include the MRI contrast agent t having hybrid structures of “zinc-containing metal oxide nanoparticle—biologically/chemically active material” in which the nanoparticle is conjugated with a bioactive material such as proteins, antibodies, and chemical materials.

Microporous glutaraldehyde-crosslinked chitosan sorbents, methods of making and using them, and a generator for the radioisotope ⁹⁹Mo containing the sorbents.

A method of recovering daughter isotopes from a radioisotope mixture. The method comprises providing a radioisotope mixture solution comprising at least one parent isotope. The at least one parent isotope is extracted into an organic phase, which comprises an extractant and a solvent. The organic phase is substantially continuously contacted with an aqueous phase to extract at least one daughter isotope into the aqueous phase. The aqueous phase is separated from the organic phase, such as by using an annular centrifugal contactor. The at least one daughter isotope is purified from the aqueous phase, such as by ion exchange chromatography or extraction chromatography. The at least one daughter isotope may include actinium-225, radium-225, bismuth-213, or mixtures thereof. A liquid-liquid extraction system for recovering at least one daughter isotope from a source material is also disclosed.

A method of manufacturing a component. In a preferred embodiment, the method includes enriching an element with an isotope and using the enriched element as a material of the component. A property of the first isotope being the same as a property of a second isotope and is preferably a mechanical, chemical, or electrical property. A second element can also be used as a material of the component, for instance, where the material is an alloy or a composite material. Further, the first isotope can be a lighter isotope of the element. Lightweight components may be manufactured using the method such that mobile platforms (e.g. spacecraft) can be assembled from the component(s). In other exemplary embodiments, the element can be hydrogen, lithium, boron, magnesium, titanium, or iron. Additionally, the component may carry a load. Components including isotopically enriched elements are also provided.

The invention includes a radionuclide generator having an ion exchange sorbent that comprises oxygen-containing functional groups grafted by organic linking groups to an inorganic oxygen-linked network and a parent isotope. For ²¹²Bi or ²¹³Bi generators, the parent isotope may be ²²⁴Ra, ²²⁵Ra or ²²⁵Ac. The surface area of the sorbent is preferably less than about 10 m²/g and more preferably less than about 1 m²/g. The exchange sorbent may be formed of any covalently bonded inorganic oxide that is capable of forming oxygen-linked networks. The oxidized functional groups may include sulfonato groups, may include moieties selected from —SO₃H, —SO₃Na, —SO₃K, —SO₃Li, —SO₃NH₄ or may include moieties selected from —PO(OX)₂ or —COOX, wherein X is selected from H, Na, K or NH₄ or combinations thereof. A ²¹³Bi or ²¹²Bi generator process includes eluting ²¹³Bi or ²¹²Bi with an aqueous solvent that includes ²²⁵Ac or ²²⁵Ra or ²²⁴Ra on the above support medium.

The present invention relates to a method for separating metals by micellar ultrafiltration, that can be used to process radioactive waste. This process consists of adding to the aqueous solution a complexing anionic surfactant having the formula: M+=cation, for example NA+m=1 or 2 R=aliphatic group of at least eight carbon atoms, and optionally a nonionic surfactant such as Triton X-100 or Brij 35 to produce micelles (10) in the solution, and subsequently of filtering the solution through an ultrafiltration membrane (3) to retain the previously formed micelles.

A system and method for tritium separation systems using a mixed bed catalytic exchange process in a Liquid Phase Catalytic Exchange/Closed Loop Continuous Process (LPCE/CLCP) system, that operates as a low temperature and low pressure continuous balanced process, designed to rapidly, economically and safely extract and isolate isotope specific products without generating unwanted products in the form of new waste streams.

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