23 results about "Technetium" patented technology

The invention discloses a method for extracting technetium from PUREX process waste liquid. The method comprises the working procedures of material blending, extraction, washing, reduction reverse extraction and reverse extraction. According to the technetium extraction and purification process, the technetium is separated from elements such as zirconium, niobium, ruthenium, uranium, neptunium and plutonium by using the technetium as a raw material, tributyl phosphate (TBP) as an extracting agent and hydrogenated kerosene (OK) as a diluent, meanwhile, the economy and safety of the reagents adopted by the process are verified by engineering, and the operation process of each procedure also meets the requirements of high flux, continuous treatment and remote operation.

The present invention concerns dosimetry-related uses and methods that rely on a metastable technetium-99 (99mTc) radionuclide-comprising (dosimetric) radioimmunoconjugate (RIC) to evaluate or establish therapeutic doses of yttrium-90 (90Y) radionuclide-comprising (therapeutic) radioimmunoconjugates, both RICs targeting the same molecular target. More specifically, radioimmunoconjugates that comprise a CD66-binding component are used. Another aspect of the invention relates to Single Time Point dosimetry procedures replacing the need for traditionally used Multiple Time Point dosimetry procedures for yttrium-90 (90Y) radionuclide-comprising radioimmunoconjugates.

This invention relates to a method for extracting technetium from nuclear fuel reprocessing waste using an extraction method, comprising the following steps: preparing a technetium-zirconium co-extraction feed; co-extracting technetium and zirconium using a TBP-kerosene organic solution as the extractant; reducing and back-extracting the technetium in the TBP organic phase co-extraction solution using a reducing stripping agent; and back-extracting the zirconium in the remaining TBP organic phase using a complexing stripping agent. This method improves upon existing techniques for extracting technetium from nuclear fuel reprocessing waste using extraction methods, which generate a significant amount of organic phase radioactive waste other than tributyl phosphate and require an additional solvent treatment system. It avoids introducing additional solvents and metal ion groups into the reprocessing system and is better compatible with the main process flow and facilities of reprocessing plants using the PUREX process, thus making the technetium extraction process easier to implement in reprocessing plants. The method provided by this invention prevents technetium from entering vitrification and geological disposal, thus preventing long-term hazards to the biosphere.

The invention relates to the technical field of composite carbon materials, in particular to a metal carbide modified carbon material and a preparation method and application thereof. The preparation method provided by the invention comprises the following steps: mixing a carbon material with a modification element precursor to obtain mixed powder; and calcining the mixed powder in a protective atmosphere or a vacuum environment to obtain the metal carbide modified carbon material, the modification elements in the modification element precursor comprise one or more of scandium, titanium, vanadium, chromium, manganese, iron, cobalt, nickel, yttrium, zirconium, niobium, molybdenum, technetium, ruthenium, rhodium, palladium, hafnium, tantalum, tungsten, iridium and rhenium; the carbon material is an unmodified carbon material or a three-dimensional structure graphene modified carbon material. According to the preparation method, the coupling and adhesion capacity between the metal carbide and the carbon material can be remarkably improved, so that the carbon material has excellent structural stability, and the carbon material is applied to the fields of thermal management materials, high-strength and high-conductivity metal-based composite materials, high-conductivity and high-thermal-conductivity resin-based composite materials and polishing agent abrasives.

The invention relates to a method for recovering one or more metals selected from the group consisting of technetium, ruthenium, palladium and rhenium from an aqueous solution, comprising: adding to said solution a reagent to precipitate the one or more metals in the solution, and then collecting the precipitate thus formed, characterized in that the reagent comprises or consists of 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, 1, 3-butanediol, or 1, 3-butanediol. The invention relates to a composition consisting of a mixture of 1, 10-phenanthroline and a nickel (II) salt selected from the group consisting of nickel (II) nitrate and nickel (II) sulfate. Application: recovering all or part of technetium, ruthenium and / or palladium from an aqueous solution of the fission product prior to an operation intended to vitrify the fission product; and recovering all or part of ruthenium, palladium and / or rhenium from the solution obtained by pickling the industrial and / or municipal waste with nitric acid.

A switching device including a first electrode layer, a second electrode layer arranged to face the first electrode layer, and a selection layer arranged between the first electrode layer and the second electrode layer, wherein the first electrode layer is doped with at least one of manganese (Mn), iron (Fe), niobium (Nb), molybdenum (Mo), technetium (Tc), ruthenium (Ru), rhodium (Rh), hafnium (Hf), tantalum (Ta), tungsten (W), rhenium (Re), iridium (Ir), or platinum (Pt), may be provided.

The invention discloses a method for solidifying radionuclide technetium in a solution and application of the method, and belongs to the technical field of spent fuel post-treatment. The curing method comprises the following steps: firstly, acidifying a technetium-containing solution by using nitric acid, then adding the acidified technetium-containing solution into a sodium metaaluminate solution, and reacting under a low-temperature condition to ensure that technetium is covalently bonded with high-activity mu < 1->-OH in a curing product in a form of pertechnetate (TcO4 <->) and is cured in pseudo-boehmite-like mixed crystals with stable crystal structures; the curing method is mild in operation condition and simple in technological process, overcomes the defects that in a traditional cement curing and glass curing method, the size of a cured product is increased, technetium is prone to volatilization and the like, has the advantages of being small in cured product size, high in curing efficiency, low in cost and environmentally friendly, the maximum curing amount can reach 100 mg / g, and the curing method is suitable for industrial production. The method is suitable for treating technetium-containing acidic nucleus waste liquid generated in the spent fuel post-treatment process.

This invention provides a technetium-99m preparation system, its usage method, and its applications. The system includes a reaction module and a purification module. The reaction module includes an electron beam channel, a shell, and a gamma photon fabrication unit inside the shell, forming a Mo-containing solution region between the shell and the gamma photon fabrication unit. The gamma photon fabrication unit is used to produce gamma photons, and the Mo-containing solution region is used to contain the Mo-containing solution and carry out the reaction. The gamma photon fabrication unit includes a metal outer layer and at least one target plate disposed inside the metal outer layer. The electron beam channel penetrates the shell and connects to the metal outer layer of the gamma photon fabrication unit. The Mo-containing solution region and the purification module also include a circulation loop. The mixture generated in the Mo-containing solution region is passed through the circulation loop into the purification module for purification to form a technetium-99m separation solution and a Mo separation solution. The Mo separation solution is returned to the Mo-containing solution region through the circulation loop. By using the technetium-99m preparation system provided by this invention, low-cost and high-efficiency production of technetium-99m can be achieved.

To extract, at hospitals or other use points, a radiopharmaceutical made from low-specific-activity radioactive molybdenum-99 and of radioactive technetium-99m as a raw material for a labeling compound for the radiopharmaceutical. Pack activated charcoal or alumina into a cylindrical metal mesh rather than a column, with which the flow rate is limited, in advance, adjust the molybdenum solution to a low-pH, acidic state, and then immerse the cylindrical metal mesh into the molybdenum solution while the molybdenum solution is flowing as a result of being stirred, to prepare a molybdenum-99-adsorbing column in which the molybdenum-99 has been adsorbed onto the activated charcoal or alumina, then elute the technetium-99m produced from the molybdenum-99 from the molybdenum-99-adsorbing column, and pass the eluate through a technetium-99m-purifying column packed with aluminum oxide to trap the technetium-99m, and then pass neutral-pH physiological saline through the technetium-99m-purifying column to separate and elute the technetium-99m, thereby purifying and recovering the technetium-99m while removing impurities.

This invention discloses a uranium-aluminum alloy for uranium-molybdenum-technetium targets and its preparation method. The uranium content in the uranium-aluminum alloy is 25.4 ± 2.0 wt%. The impurity content of the uranium-aluminum alloy meets the following requirements: B ≤ 1 μg / g, Cd ≤ 2 μg / g, Li ≤ 8 μg / g, C ≤ 600 μg / g, Cr ≤ 240 μg / g, Cu ≤ 100 μg / g, Fe ≤ 1000 μg / g, Ni ≤ 250 μg / g, and Gd ≤ 0.25 μg / g. The density of the uranium-aluminum alloy is 3.0–3.4 g / cm³. 3 The uranium-aluminum alloy comprises Al, UAl3, and UAl4 phases. The preparation method of the uranium-aluminum alloy includes the following steps: first, heating to 1190-1220℃ within 50-80 minutes; then holding at 1190-1220℃ for 57-63 minutes; then cooling naturally to 990-1030℃; finally, tilting and casting into a mold at 350-500℃, followed by annealing the prepared ingot under vacuum at 580-640℃ for 10-25 hours. This invention enables the production of uranium-aluminum alloys with controllable composition, density, phase, and microstructure through induction melting-tilt casting-annealing, meeting production requirements. 99 Requirements for uranium-aluminum alloy targets for Mo.

A generator column assembly for the elution of a radioisotope, having a generator column container having a bottom wall defining a flow outlet aperture, an open top end, and a sidewall extending from the open top end to the bottom wall that defines an interior volume having a substantially cylindrical upper volume portion and a substantially cylindrical lower volume portion, the upper volume portion having a diameter that is greater than a diameter of the lower volume portion, and a closure cap assembly including a container cap defining a flow inlet aperture.

The invention provides a radiopharmaceutical operation cabinet. The radiopharmaceutical operation cabinet comprises a cabinet body, a fourth inner cavity formed in the upper portion of the cabinet body, and a second through hole and a third through hole which are formed in the upper portion of the cabinet body and communicated with the fourth inner cavity. The molybdenum-technetium generator can be placed in the operation cabinet by lifting the molybdenum-technetium generator to a position higher than the first sliding plate, while the molybdenum-technetium generator can be placed in the operation cabinet only by lifting the molybdenum-technetium generator to a position higher than the fixed frame in the prior art, so that compared with the prior art, the parts can achieve the effects of saving manpower and improving the working efficiency. The molybdenum technetium generator can be conveniently placed in the operation cabinet; a contact block is extruded through a pushing block, the contact block rotates to drive a rotating shaft to rotate, the rotating shaft rotates to drive a rotating frame to rotate, so that the rotating frame is unfolded, at the moment, an operator can move the molybdenum-technetium generator out of a first sliding plate or move a new molybdenum-technetium generator on the first sliding plate, and therefore the rotating frame is automatically rotated and unfolded; and the molybdenum-technetium generator can be conveniently moved out or in by people.

It would be advantageous with the close quarters in breast procedure rooms to create a Gamma Probe system incorporating multi modalities, sensitivities, and multiple ranges for utilization in breast intervention procedures for the diagnosis and verification for breast cancer. Systems for specimen radiography assist radiologists and surgeons in the detection and classification of abnormal lesions in medical images gleaned from breast biopsies / lumpectomies and a Gamma Probe system with a sensor to detect radioactive, metallic / magnet, or RFID waves to help locate a lump or abnormality which were placed by an Interventional Radiologist. Radio guided surgery (RGS) uses gamma probes to localize radioactive targets. The three common RGS radioisotopes: technetium-99m (99mTc), indium-111 (111In), and iodine-125 (125I). The aspects of the disclosed embodiments pertain to a gamma probe that is able to simultaneously detect 2-3 isotopes automatically without actuating a physical switch.

The invention relates to the technical field of analysis and detection, in particular to a method for determining the total tin content in a medicine box matched with technetium [99mTc]. The method provided by the invention comprises the following steps: determining the total tin content in a medicine box matched with technetium [99mTc] by adopting a volt-ampere polarography; the test solution is a solution obtained by mixing and reacting a to-be-detected medicine box containing stannous chloride with concentrated hydrochloric acid in an oxygen-containing atmosphere and then adding water for dilution; an electrolyte solution is used as a determination medium, and the electrolyte solution is a concentrated hydrochloric acid diluent. According to the method for detecting the total tin content in the matched medicine box for technetium [99mTc], the total tin content can be effectively detected, the analysis speed is high, and the accuracy is high.

The invention belongs to the technical field of metal element separation, and relates to a separation method of molybdenum and cadmium in an irradiation thorium target, the separation method comprises the following steps: (1) using a first hydrochloric acid solution to shower an anion resin chromatographic column to obtain a pretreated chromatographic column; (2) pretreating a hydrochloric acid solution for irradiating the thorium target, and passing through the pretreated chromatographic column obtained in the step (1); the pretreatment chromatographic column after sample loading is subjected to first washing with a second hydrochloric acid solution and second washing with a third hydrochloric acid solution or ultrapure water, and the first washing and the second washing have no sequence; and respectively collecting the molybdenum-containing effluent after the first washing and the cadmium-containing effluent after the second washing. According to the invention, rapid and efficient separation of molybdenum and cadmium can be realized, and support is provided for subsequent production of medical-grade molybdenum-99 / technetium-99m and cadmium-115 / indium-115m generators.

A zirconia compositional substance includes at least one first transition metal element selected from the group consisting of manganese, iron, cobalt, nickel, copper, molybdenum, technetium, ruthenium, rhodium, palladium, and silver; a coloring element that is either or both of a lanthanoid rare earth element and a second transition metal element different from the first transition metal element; and stabilizing element-containing zirconia, wherein a content of the first transition metal element is 100 mass ppm or more, a content of the second transition metal element is less than 100 mass ppm.

The invention discloses a preparation method and application of a photocatalytic material for selective technetium separation. Amorphous titanium dioxide prepared through a hydrolysis method, ZnCl2 and KCl are mixed and ground into powder, then the powder is calcined in a muffle furnace for 2-4 hours at the temperature of 250-350 DEG C, solids obtained after cooling are washed with water for multiple times and then dried, the photocatalytic material for selective technetium separation is obtained, and the molar ratio of the dosage of ZnCl2 to the dosage of KCl is 55: 45. The preparation method comprises the following steps: adding the photocatalytic material into an aqueous solution containing technetium and containing high-concentration nitrate radicals, simulating sunlight irradiation for 4-5 hours to extract pertechnetate radicals, then physically filtering and recycling a mixture of the photocatalytic material and a technetium product, finally extracting technetium through hydrogen peroxide oxidation, and secondarily filtering and regenerating the photocatalytic material for reuse. The photocatalyst material prepared by the invention is especially suitable for separating technetium from an aqueous solution containing high-concentration nitrate radicals, and overcomes the key problems of low reduction separation efficiency and low selectivity of high-technetium acid radicals caused by interference of the nitrate radicals on the separation process.

The invention belongs to the technical field of radionuclide separation, and particularly provides a method for extracting technetium from a molybdenum-technetium mixed solution. The invention aims to solve the problems of poor adsorption selectivity, infirm adsorption, need of specific pH or ion concentration and high cost when an existing adsorption material is used for adsorbing 99mTc. Therefore, the method for extracting technetium from the molybdenum-technetium mixed solution comprises the following steps: S1, preparing a plant-based biomass carbon adsorption column; s2, adsorbing technetium in the molybdenum-technetium mixed solution by using a plant-based biomass carbon adsorption column; s3, leaching the plant-based biomass carbon adsorption column by using a strong alkali solution to obtain technetium-containing leacheate; wherein the filler in the plant-based biomass carbon adsorption column is an oxidized plant-based biomass carbon material. According to the method, the plant-based biomass carbon adsorption column is used for adsorbing the < 99m > Tc in the molybdenum-technetium mixed solution, the affinity of the oxidized plant-based biomass carbon material and the < 99m > Tc is higher, the adsorption effect of the < 99m > Tc is good, and the cost is low.

The invention provides a preparation method and application of a magnetic bentonite composite material, and belongs to the technical field of environmental protection engineering. The invention provides a magnetic bentonite composite material preparation method, which comprises: (1) fully mixing bentonite, CaCl2, Fe (NO3) 3.9 H2O and citric acid monohydrate, dissolving in an aqueous solution, and carrying out a heating reaction to obtain a Bt (at) CF composite material; and (2) fully mixing and dissolving the Bt-CF composite material in an aqueous solution, adding FeSO4. 7H2O, adjusting the pH value of the mixed solution to 9.0-11.0 by using stronger ammonia water, adding H2O2, magnetically stirring at 60-80 DEG C for 2-3 hours, aging for 2 hours, washing the precipitate until the pH value is 7.0, and drying to obtain the Bt-CF-F composite material. The magnetic bentonite composite material prepared by the method disclosed by the invention can be used as a uranium and / technetium adsorbent, has the characteristics of high selectivity and high efficiency in adsorbing uranium and / or technetium, can be recycled, and has a great application value in the aspect of radioactive element wastewater treatment.

A generator column assembly for the elution of a radioisotope, having a generator column container having a bottom wall defining a flow outlet aperture, an open top end, and a sidewall extending from the open top end to the bottom wall that defines an interior volume having a substantially cylindrical upper volume portion and a substantially cylindrical lower volume portion, the upper volume portion having a diameter that is greater than a diameter of the lower volume portion, and a closure cap assembly including a substantially cylindrical container cap defining a flow inlet aperture, the container cap being configured to be slidably received in the open top end of the generator column container.

This invention relates to an adsorption chromatographic material for technetium in radioactive waste and its preparation method, comprising the following steps: S1, purging air from the pores of a silica matrix; S2, mixing acetophenone, diethyl phthalate, vinylpyridine, and divinylbenzene evenly, then adding initiator azobisisobutyronitrile and initiator V-40 and mixing evenly to obtain a mixture; S3, mixing the mixture with the silica matrix under vacuum conditions, then restoring to atmospheric pressure, and then cycling under vacuum-atmospheric pressure conditions multiple times, finally restoring to atmospheric pressure; S4, subjecting the reaction system after restoring to atmospheric pressure to multiple continuous heatings, finally restoring to room temperature, and rinsing multiple times to obtain the adsorption chromatographic material. Compared with the prior art, this invention directly obtains a pyridine-type weakly basic anion exchange resin through copolymerization, eliminating the chloromethylation and amination processes in the traditional process, greatly simplifying the resin synthesis process, and improving the adsorption chromatographic material for technetium in radioactive waste. 99 Tc has a good adsorption effect.

The invention discloses a method for preparing clinical medicinal grade iodine 131 by separating medical radioactive waste, and relates to the technical field of pharmaceutical engineering, the clinical medicinal grade iodine 131 is prepared by pre-treating the used molybdenum-technetium generator medical radioactive waste liquid through selective precipitation, microorganism-electrochemical catalytic reduction, oxidative distillation, chromatographic refining and degerming filtration blending. The method has the advantages that the purity and the yield of the iodine 131 separated from the medical waste liquid are remarkably improved by establishing a set of multi-stage synergistic purification process comprising pretreatment, selective precipitation, microorganism-electrochemical catalytic reduction, oxidative distillation and chromatographic refining; wherein filtration acidification and carrier addition in the pretreatment stage effectively guarantee the stability of an initial material, and subsequent precipitation and reduction operations deeply remove isotope impurities such as technetium and molybdenum and radioactive interference nuclides, so that the problems of low product purity and high impurity residue caused by single purification step in the traditional process are thoroughly solved.