1410 results about "Uranium" patented technology

An apparatus for selective radiation detection includes a neutron detector that facilitates detection of neutron emitters, e.g. plutonium, and the like; a gamma ray detector that facilitates detection of gamma ray sources, e.g., uranium, and the like; and/or an X-ray analyzer that facilitates detection of materials that can shield radioactive sources, e.g., lead, and the like.

The invention is related to nuclear physics, medicine and oil industry, namely to the measurement of x-ray, gamma and alpha radiation; control for trans uranium nuclides in the habitat of a man; non destructive control for the structure of heavy bodies; three dimensional positron-electron computer tomography, etc.The essence of the invention is in additional ingredients in a chemical composition of a scintillating material based on crystals of oxyorthosilicates, including cerium Ce and crystallized in a structural type Lu2SiO5.The result of the invention is the increase of the light output of the luminescence, decrease of the time of luminescence of the ions Ce3+, increase of the reproducibility of grown crystals properties, decrease of the cost of the source melting stock for growing scintillator crystals, containing a large amount of Lu2O3, the raise of the effectiveness of the introduction of SCintillating crystal luminescent radiation into a glass waveguide fibre, prevention of cracking of crystals during the production of elements, creation of waveguide properties in scintillating elements, exclusion of expensive mechanical polishing of their lateral surface.

This invention deals with multi-component composite materials and techniques for improved shielding of neutron and gamma radiation emitting from transuranic, high-level and low-level radioactive wastes. Selective naturally occurring mineral materials are utilized to formulate, in various proportions, multi-component composite materials. Such materials are enriched with atoms that provide a substantial cumulative absorptive capacity to absorb or shield neutron and gamma radiation of variable fluxes and energies. The use of naturally occurring minerals in synergistic combination with formulated modified cement grout matrix, polymer modified asphaltene and maltene grout matrix, and polymer modified polyurethane foam grout matrix provide the radiation shielding product. These grout matrices are used as carriers for the radiation shielding composite materials and offer desired engineering and thermal attributes for various radiation management applications.

This disclosure describes nuclear fuel salts usable in certain molten salt reactor designs and related systems and methods. Binary, ternary and quaternary chloride fuel salts of uranium, as well as other fissionable elements, are described. In addition, fuel salts of UCl_xF_y are disclosed as well as bromide fuel salts. This disclosure also presents methods and systems for manufacturing such fuel salts, for creating salts that reduce corrosion of the reactor components and for creating fuel salts that are not suitable for weapons applications.

The invention relates to apparatus for positioning and cooling melting lining layer when the melting lining layer flows out of the reactor shell during accident. The improvements of invention are described as following: the design of the cooling lining layer trap which has steel basket shell and arranged in sub-reactor, wherein sacrifice material is filled in the steel basket for diluting the component containing uranium and steel component of the lining layer; the shape (arrangement) of the sacrifice material in reactor core; the choice for optimum number of the sacrifice material; installation of guiding device used for the lining layer flowing into the reactor core trap. Especially, the thickness of bottom of the cooling jacket is larger than the thickness of the side wall thereof for no less than 30%, and the cooling jacket inclines to its center for 10 to 20 degree; the diluent and sacrifice material is made into blocks packed in steel shell.

The invention provides an amidoxime-based chelate polyacrylonitrile fiber and its preparation method. The preparation method of the amidoxime-based chelate polyacrylonitrile fiber comprises the following steps that 1, polyacrylonitrile fibers are subjected to irradiation cross-linking under the conditions of an inert gas atmosphere and an irradiation amount of 20 to 2000kGy; and 2, the polyacrylonitrile fibers subjected to irradiation cross-linking undergo an amidoximation reaction. The invention also provides a use of the amidoxime-based chelate polyacrylonitrile fiber as a metal ion adsorption material. Through the preparation method which is simple relatively, the amidoxime-based chelate polyacrylonitrile fiber which has good mechanical properties, high fiber breaking strength, a long reuse life and stable performances and is damaged difficultly is obtained. The amidoxime-based chelate polyacrylonitrile fiber can be utilized as a metal ion adsorption material, is suitable for the fields of purification of industrial waste water or slag, or extraction of uranium of seawater, has excellent effects of adsorbing multiple metal ions, and can greatly reduce an industrial cost.

The invention provides a method for determining the content of organic carbon in hydrocarbon source rock. The method comprises the following steps: S1. experimentally measuring the content of organic carbon of a hydrocarbon source rock core; S2. acquiring the log values of uranium content, resistivity, compensated sonic time difference and compensated density in logging curves at rock core depths; S3. calculating delta lgR according to the log values of S2; S4. establishing the correspondence relation between the content value of organic carbon and the delta IgR as well as uranium log value according to the results of S1 to S3, and fitting into a function; S5. according to the fitted function, determining the content of organic carbon of hydrocarbon source rock at all positions of a log. The invention provides a device embodiment for determining the content of organic carbon in hydrocarbon source rock. According to the method and the device for determining the content of organic carbon in hydrocarbon source rock, the content of organic carbon is determined based on combination of uranium logging curves and delta lgR, the accuracy of the calculation results is high, and the method and the device have obviously practically application effects in the calculation on the content of organic carbon in the hydrocarbon source rock.

The invention discloses a method for determining uranium lead age of baddeleyite by using secondary ion mass spectroscopy, which belongs to the field of secondary ion mass spectroscopy dating method in geochronology and particularly relates to a method applied to a large secondary ion mass spectrometer having oxygen blowing conditions. In the measurement method described by the invention, the degree of vacuum in a sample cavity of the spectrometer is enabled to reach 1E-8Torr at first, and an oxygen leak valve is then opened to blow the oxygen with the purity of 99.99% to the surface of the sample so that the oxygen partial pressure of the sample cavity reaches 1E-5Torr. The method of the invention can reduce the optical axis effect, which is exhibited by secondary ion mass spectroscopy during the uranium-lead dating of baddeleyite, to be below 2%, thus the accuracy of uranium-lead age when the baddeleyite is subjected to secondary ion mass chromatographic analysis is significantly improved. After uranium-lead measurement ratio is obtained, uranium-lead fractionation correction is performed by adopting a linear relation between ln(Pb+/U+) and ln(UO2+/U+) in order to obtain accurate age.

Oxyanions of various contaminant elements, such as chromium, antimony, molybdenum, tungsten, vanadium and uranium, are removed from water and other aqueous feeds by treating the feed with (1) a sorbent comprising one or more rare earth compounds, usually mixed or supported on particulate solids having a cation exchange capacity less than 20 milliequivalents per 100 grams or (2) an aqueous solution of one or more soluble rare earth compounds.

The invention discloses a nuclear reactor core capable of improving neutron flux rate; the core is a circular structure; the center of the core is a Be thermal neutron well region, the circumference of which is a fuel component region, the circumference of which is a Be reflecting layer; the whole core is a circular structure. The maximum thermal neutron flux rate of the central Be thermal neutron well region can be over twice larger than that of the fuel region ; as the thermal neutron well region can be provided with fuel component with high uranium content, the fuel component center can form fast neutron region, therefore, the generated fission neutron has high flux rate.

The invention discloses a method for deeply purifying and recovering hyperhaline fluoric-U radioactive waste solution. The method comprises three sequent treatment phases including pretreatment, enrichment and separation of U-F and recovery of U, and can be further finely divided into 7 treatment units, including lower temperature reduction separation, chemical sedimentation, film filtering, step-grade organism enrichment-film filtering separation or step ion exchange, deposit fluorine removal solid-liquid separation, microwave tower treatment and resource recovery of U; under the condition that the salinity of waste solution to be treated is 0.1-10%, the initial U concentration is 0.1-1000mg/L, and the fluorine concentration is 0.01-15g/L, after treatment, the concentration of U is not larger than 0.05mg/L, the concentration of the fluorine is not higher than 10mg/L, and the recovery rate of U is larger than 90%. The method for deeply purifying and recovering hyperhaline fluoric-U radioactive waste solution is suitable for treating hyperhaline fluoric-U radioactive waste solution and the resource recovery of radioactive U.

A safe, reliable and rapid system for the detection of nuclear materials within containers includes the use of pulsed high-intensity gamma rays that can penetrate a container and its contents and can be detected outside the container to provide a display in which high-Z material, including lead, uranium, plutonium and other nuclear substances that absorb gamma rays are detected as black regions on the display. In one embodiment, orthogonal pulsed gamma ray beams illuminate the container from two different directions to provide three-dimensional slices from which the existence and location of nuclear threat materials can be ascertained in as little as four seconds for a 40-foot container.

The present invention provides a method of simultaneously removing uranium and transuranics from metallic nuclear fuel in an electrorefiner. In the method, a potential difference is established between the anode basket and solid cathode of the electrorefiner, thereby creating a diffusion layer of uranium and transuranic ions at the solid cathode, a first current density at the anode basket, and a second current density at the solid cathode. The ratio of anode basket area to solid cathode area that is selected based on the total concentration of uranium and transuranic metals in a molten halide electrolyte in the refiner and the effective thickness of the diffusion layer at the solid cathode, such that the established first and second current densities result in both codeposition of uranium and transuranic metals on the solid cathode and oxidation of the metallic nuclear fuel in the anode basket. Deposited material on the solid cathode is removed, and the first current density at the anode basket is maintained to prevent substantial oxidation of the anode basket during operation of the electrorefiner.

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 method for treating uranium-containing waste water comprises uranium removal and phosphorus removal two techniques, firstly adding uranium removing agent in a uranium removal tank of uranium-containing waste water according to that the mass concentration ratio of uranium to uranium removing agent is 2-20 to 1, stirring and settling the uranium-containing waste water in the uranium removal tank, processing solid-liquid separation to the flocculated waste water through plate sedimentation tank to remove the uranium in the uranium-containing waste water, so that the uranium removal reaches up to 99%, and then, adding lime milk clarifying solution in the flocculation basin, stirring, settling and solid-liquid separation to remove the phosphorus in the waste water, in which the phosphorus removal reaches up to 99.99%. At last, supernatant is added with acid to regulate the pH value to 6 to 9 in the regulating tank.

The invention relates a device for measuring the relative proportions of uranium and plutonium in a radioactive package (16), having:a source of photons (10) for irradiating the package,at least one delayed neutron detector (18) able to deliver counting signals for neutrons emitted by the package,means (22, 30) of acquiring counting signals, able to establish a decrease over time in the neutrons emitted, characteristic of the radioactive package,calculation means (32) for comparing the decay characteristic of the radioactive package with the respective characteristic decays of uranium and plutonium and for establishing relative proportions of uranium and plutonium in the package.

A method for preparing ceramsite light aggregate by utilizing sludge belongs to solid waste treatment, proposal and resource technical field, which produces dry sludge through thermal drying dehydrated mud cake whose moisture percentage is 75-85%, wherein the dry sludge is broken and sieved through 45 mm screen, 18-25% fly ash by mass of the dry sludge is charged to stir, the stirred hybrid material is injected in the powder tablet compressing machine to squeeze and mould under the pressure of 8 Mpa to obtain ceramsite light aggregate blank, the blank is heated to 350 DEG C and is stayed for 20 min in order to lead to volatilize and exhaust completely and is calcined under the temperature of 1080-1100 DEG C for 30 min and is cooled to indoor temperature naturally, and then the ceramsite light aggregate is obtained. The invention for processing living sewage and sludge improves strength of the calcined product, and saves the calcined energy consumption greatly. The obtained ceramsite light aggregate is compact in uranium face, light in weight, high in strength, and satisfies national standard GB/T17431.2-1998, which can be used in the projects such as buildings, roads, bridges and the like.

Refractory superalloys consist essentially of a primary constituent selected from the group consisting of iridium, rhodium, and a mixture thereof, and one or more additive elements selected from the group consisting of niobium, tantalum, hafnium, zirconium, uranium, vanadium, titanium and aluminum, and the superalloys having a microstructure containing an FCC-type crystalline structure phase and an L12-type crystalline structure phase are precipitated. Preferably the amount of additive element(s) is 2 to 22 atom %.