256 results about "Plutonium" 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.

A process to safely convert about 95% of the nuclear waste into a usable fuel source is disclosed. The process, involving a sub-critical power reactor and a proliferation-resistant fuel cycle, consumes depleted uranium or thorium fuel with fissionable fuel, including reactor or weapons-grade plutonium. The reactor is comprised of coaxial neutron and energy-amplifying regions separated by moderating and thermal neutron absorbing layers. Control of the water or gas-cooled reactor is provided by plutonium-helium loops with a variable volume flow rate and an external source of neutrons that quickly reacts to any fluctuations of the reactor parameters. A second embodiment of the invention is a compact sub-critical propulsion reactor utilizing fission electric cell and thermo-acoustic technology for electrical power generation.

A light water reactor to safely convert depleted uranium into a fuel source that could be used as a sustainable source of energy for centuries. The reactor is a type of breed-burn reactor uniquely combined with a proliferation-resistant fuel cycle with no uranium enrichment and no plutonium isolation. It is comprised of a compact factory-produced fast region and a thermal region that produces about 95% of the core power and contains the passageways for transports of delayed-neutron emitters to the fast region, where they can provide additional neutrons (source-based mode) or all the necessary excitation without an external neutron source (self-regulating mode). A second embodiment of the invention is a small unit driven by a neutron source with beam recycling for propulsion, electrical power or radioisotope production. It could also serve as a demonstration facility for the transmutation reactor with fission-fusion fuel.

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 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.

The invention discloses a cement-based solidified material for processing middle and low radioactive incineration ash and a method for processing middle and low radioactive incineration ash. The solidified material, using the slag, fly ash, zeolite, metakaolin, cement clinker as the material to be excited, is obtained through the even mixture with activator after grinding. The liquid activator or combined solid activator can be used as the activator. Mix the material to be excited and activator together, according to the package content of 30-40%, mix with the incineration ash, stir, and maintain to solidification in light of the solidification operation. The cement-based solidified material of the invention can prepare the nuclear waste solidified body with the incineration ash package content greater than 30%, stable mechanical performance, and low leaching rate of radionuclide ion. Especially when solidifying the plutonium-containing waste, the 42-day Pu leaching rate is 10 cm/d. The invention will play an important role in the middle and low radioactive incineration ash processing with good application prospect.

The prevent invention discloses a method for recovering a trace of uranium and/or plutonium from radioactive organic liquid waste. The method comprises: firstly, performing stripping on radioactive organic liquid waste with a carbonate solution; then performing oscillation adsorption, column chromatography or membrane filtration on a stripped aqueous phase by using a solid functional adsorption material; recycling a treated liquid again as a stripping solution; using acid to perform elution on the solid functional adsorption material that adsorbs the uranium and/or plutonium, to obtain a rich aqueous solution containing the uranium and/or plutonium; recycling the adsorption material after the elution by using an aqueous bicarbonate solution for soaking or washing. The method combines stripping of carbonate and recycling of a low concentration of uranium and/or plutonium from a (weak) alkaline aqueous solution, and achieves recycling of the functional adsorption material and the stripping solution, thereby reducing the amount of waste as much as possible.

A new device is provided for the electrorefining of uranium in spent metallic nuclear fuels by the separation of unreacted zirconium, noble metal fission products, transuranic elements, and uranium from spent fuel rods. The process comprises an electrorefiner cell. The cell includes a drum-shaped cathode horizontally immersed about half-way into an electrolyte salt bath. A conveyor belt comprising segmented perforated metal plates transports spent fuel into the salt bath. The anode comprises the conveyor belt, the containment vessel, and the spent fuel. Uranium and transuranic elements such as plutonium (Pu) are oxidized at the anode, and, subsequently, the uranium is reduced to uranium metal at the cathode. A mechanical cutter above the surface of the salt bath removes the deposited uranium metal from the cathode.

By sequentially aligning various filtration media and delivery systems, enhanced synergistic reduction of water contaminants is obtained compared to the prior art or separate use of the individual media/filters. Specific filtration media are formulated with proper proportioning and sequencing to enhance the ability to reduce metals that cause staining, odors and bad taste such as iron, copper and manganese. Also disclosed is the reduction of potentially hazardous metal radionuclides metals such as uranium, iodine, cesium, plutonium and radium. Also disclosed is improved removal of heavy metals such as arsenic, lead, chromium, and mercury as well as organic compounds such as halogenated carcinogenic compounds. The present devices and methods remove specific bacteria and their toxins from water to reduce the risk of dermatitis. Thus, the present invention enhances our ability to achieving cleaner and safer water for drinking, swimming, washing, bathing and cooking.

The invention relates to a high-temperature supercritical nuclear reactor. A spherical solid is used as a coolant of the nuclear reactor, which is different from common nuclear reactors that adopt light water, heavy water, gas and liquid sodium. The spherical solid coolant can be graphite spheres coated with silicon carbide, stainless steel spheres, graphite spheres coated with stainless steel and the like. The spherical solid has the characteristic of good rolling ability, so that the spherical solid can roll to a steam generator under the action of gravity after heating in the reactor, and the purpose of transferring heat from the reactor to the steam generator is achieved. The reactor is applicable to both thermal neutron reactor and fast neutron reactor, and natural uranium, low-enriched uranium, plutonium and thorium can be used as nuclear fuel. The temperature of high-temperature steam/water outputted by the steam generator can reach 900 DEG C, the same parameters to a thermal power station can be reached, the requirement of supercritical even ultra-supercritical can be satisfied, and the high-temperature supercritical nuclear reactor has high safety and reliability.

An amidoxime-modified PAN porous body obtained by reacting with hydroxylamine a polyacrylonitrile porous body that is monolithic, has a thickness of 1 mm or more and contains polyacrylonitrile (PAN) as the main component to convert a nitrile group of the polyacrylonitrile porous body into an amidoxime group. This porous body is a porous body for adsorbing a metal ion, for example, an ion of metal such as copper, iron, nickel, vanadium, indium, gallium, silver, mercury, lead, uranium, plutonium, cesium, barium, lanthanum, thallium and strontium.

Fluorine or a fluorine compound is subjected to a reaction with a spent oxide fuel to produce fluorides of uranium and plutonium, and recovering the fluorides using a difference in volatility behavior. The method includes steps of: subjecting a mixture of UO₂ and PuO₂ with hydrogen fluoride mixed with hydrogen to HF-fluorinate uranium and plutonium into UF₄ and PuF₃; subjecting UF₄ and PuF₃ with a fluorine gas to F₂-fluorinate uranium and plutonium into UF₆ and PuF₆; and fractionating UF₆ and PuF₆ using a difference in phase change of obtained UF₆and PuF₆, removing a part of UF₆, and volatilizing the remaining UF₆ and PuF₆ at the same time. By such a reprocessing method, PuF₄ hard to undergo a reaction is prevented from being formed as an intermediate fluoride, the material of a reactor is hard to be corroded, and a consumption of expensive fluorine gas is reduced.

The invention discloses a device for oxidized adjusting plutonium valence by N2O4, comprising a Pu (III) oxidized column and a nitrous acid eliminating column, wherein the Pu (III) oxidized column and the nitrous acid eliminating column are connected via a material fluid transporting pipe. The device provided by the invention increases the utilization rate of the N2O4, reduces the escaped nitric oxide and realizes the continuous operation of the plutonium valence adjusting process and the remaining nitrous acid eliminating process.

The embodiment of the invention provides a sealed high-temperature electrochemical measuring device which comprises a sealing device, a heating device, an electrode body and a gas inlet and outlet, wherein the sealing device comprises an upper flange and a lower flange; both the upper flange and the lower flange are made of high-temperature alloys; the lower flange is tank-shaped; the upper flange is cover-shaped and is positioned at the opening of the lower flange; the upper flange and the lower flange are sealed through a metal graphite sealing ring and a bolt; the heating device is positioned on the outer side of the lower part of the lower flange and wraps the lower part of the lower flange; the electrode body and the upper flange are integrally designed; the electrode body extends into molten salt contained in a crucible in the tank body of the lower flange, and is used for outputting electrochemical signals from the sealed high-temperature electrochemical measuring device; the gas inlet and outlets are integrated on the upper flange in a sealed manner and are used for feeding and discharging inert gases. According to the embodiment of the invention, the device has the functions of high-temperature heating, high-temperature sealing, inert atmosphere protection, electrochemical signal output under sealing condition, and the like, and the accuracy in fused salt electrochemical measurement on uranium, plutonium and bothridium elements is improved.

The invention discloses a radioactive wastewater treatment technology and the synthesis of a treatment agent of radioactive wastewater. A modified product Y3-8103 is used for chemically reacting with radioactive wastewater, all heavy metal elements and all radioactive elements, namely cesium137, uranium 235, thorium, plutonium, thallium, protium, deuterium, tritium, radon, Sr-90 and radium, listed in the periodic table of elements and wastewater generated by all artificial nuclides. Alpha radioactive rays, beta radioactive rays, gamma radioactive rays and neutron radioactive rays are attenuated in a nationally specified mode. The radioactive wastewater treatment technology and the synthesis of the treatment agent of the radioactive wastewater can also be used for treating radioactive seawater or waste residues generated by wastewater treatment and can be used for treating heavy metal wastewater discharged by a lead-acid battery enterprise and an electronic plant and mine nucleus waste residues. The treatment cost of the radioactive wastewater is different along with different radioactive element concentration contents or different waste residue radioactive element concentration contents and different atomic weights in the wastewater, and the treatment cost of the wastewater and the waste residues ranges from RMB 90 yuan/m<3> to RMB 680 yuan/m<3>, radioactive energy with the alpha, beta and gamma radioactive ray radiant quantity smaller than or equal to 0.01Bq/1-1.10Bq/1 can be discharged to the ambient environment by the wastewater and the waste residues, the emission limit standard allowed by the international is reached, and secondary pollution is avoided. The radioactive wastewater treatment technology and the synthesis of the treatment agent of the radioactive wastewater are suitable for treating wastewater and waste residues of a nuclear power plant and treating wastewater and waste residues in mining with the radioactive elements, and are also suitable for treating wastewater containing heavy metal, namely lead, cadmium, mercury, arsenic, chromium, nickel, copper and manganese.

The invention belongs to the technical field of fuel-short after-treatment and discloses a plutonium purification and concentration method. The method comprises the following steps: extracting tetravalent plutonium in a nitric acid solution into a 30% TBP (tributyl phosphate)-kerosene solution with relatively small volume for purification and concentration in a centrifugal extractor, reducing tetravalent plutonium into trivalent plutonium by using a dimethylhydroxylamine-containing nitric acid solution with relatively small volume, performing back extraction on trivalent plutonium, and further purifying and concentrating plutonium in a water phase. In the whole concentration process, a plutonium replenishment and extraction link is removed. The method is simple in step, short in time and small in solvent radiation effect; the plutonium back-extraction yield reaches up to 99.98% while the plutonium concentration is increased by more than 10 times.

The invention relates to a radiation attenuation elastomeric material of the type comprising an elastomer in which a powder of metal oxides is dispersed, and which is characterized in that the powder of metal oxides comprises from 70 to 90% by mass of bismuth trioxide, from 5 to 15% by mass of tungsten trioxide and from 5 to 15% of lanthanum trioxide.

It also relates to the use of this elastomeric material for making individual protective articles against ionizing radiations.

It further relates to a multilayer protective glove against ionizing radiations, at least one layer of which is formed by said elastomeric material, as well as to the use of this glove for protection against ionizing radiations emitted by powders of nuclear fuels, notably with plutonium.

Fields of application: the nuclear industry for handling powders of nuclear fuels but also medical imaging, interventional radiology, nuclear medicine, processing of plastic materials, inspection and control of manufactured parts, etc.

The invention relates to a chemical separation procedure for burnup analysis of a spent fuel element. The chemical separation procedure comprises the following steps of: (I) dissolving a spent fuel element block, namely putting the spent fuel element block into concentrated nitric acid and concentrated hydrochloric acid mixed liquid with the volume ratio of 3:1 to make the spent fuel element block be completely dissolved, and cooling the dissolving liquid to room temperature; (II) diluting the dissolving liquid, namely mixing the dissolving liquid cooled in the step (I) and nitric acid, uniformly stirring the dissolving liquid and the nitric acid to obtain diluted dissolving liquid; and (III) separating uranium, plutonium, molybdenum and neodymium by a column chromatography method through the diluted dissolving liquid obtained in the step (II). The invention establishes a novel chemical separation procedure for measuring the burnup of the spent fuel element; according to the chemical separation procedure, some novel levextrel, ion exchange material and high performance liquid chromatography technologies are adopted, so that chemical separation of the uranium, the plutonium, the molybdenum and the neodymium is realized; the operation procedure is greatly simplified; the operation difficulty is reduced; and the radiation to researchers is reduced.