127 results about "Nuclear chemistry" patented technology

Provided is a method for producing a solution of atropine with improved recovery. A method for producing a solution of atropine-211, comprising the steps of: irradiating bismuth with alpha rays, generating atropine-211 in the bismuth, and distilling the bismuth irradiated with the alpha rays, separating and purifying atropine-211, and dissolving it in an aqueous sodium bicarbonate solution.

The application discloses an environmental functional nanometer material, a preparation method thereof and application of the environmental functional nanometer material in treatment of fluorine-rich radioactive nuclear wastewater, and relates to the technical field of environmental functional nanometer materials. The application discloses an environmental functional nanometer material, a preparation method thereof and application of the environmental functional nanometer material in treatment of fluorine-rich radioactive nuclear wastewater, and relates to the technical field of environmental functional nanometer materials.

Described herein is a process for recovering uranium tetrafluoride (UF4) from hexafluoride uranium (UF6), the process comprising dissolving UF6 directly in an ionic liquid and recovering UF4, which can be processed to obtain UO2(s) or metallic uranium.

The invention relates to a method for separating heavy metal cations and radionuclides from aqueous solutions. The method according to the invention can be used to separate dissolved and / or suspended naturally occurring ions or radioactive ions produced by radioactive decay, nuclear transformation reactions such as nuclear fission, or, for example, by activation with neutron radiation, from aqueous solutions.

This invention relates to the field of medical radioisotope manufacturing technology, and more particularly to a method based on proton beam irradiation. 232 Th target production 212 The method for Pb. The present invention provides a method based on proton beam irradiation. 232 Th target production 212 The method for treating Pb includes the following steps: irradiation with a proton beam. 232 Th target production 228 Th; Dissolved using a mixture of acid and water 232 Th target, allowed to stand, yields the first Th solution; separation from the first Th solution or the pretreated first Th solution... 212 Pb. This invention employs proton beam irradiation. 232 The Th target material method will significantly improve 228 The content of Th was thus significantly increased. 212 The increased production efficiency of Pb fundamentally solves the problems of low output and high cost associated with traditional methods.

The present application relates to the technical field of nuclear power plant ventilation purification, and discloses an equivalent test bench for evaluating the performance of iodine removal and a verification method, the verification method comprising: obtaining three carbon samples with different iodine removal efficiencies; based on the three carbon samples, performing a first test under the French standard system, a second test under the American standard system, a third test under different humidity conditions, and a fourth test under accident conditions on the same test bench, and recording the corresponding test results; comprehensively analyzing the test results corresponding to the four tests to obtain the test equivalent relationship between the French standard system and the American standard system, the numerical relationship between humidity and iodine removal efficiency, and the change rule of iodine removal efficiency after accident radiation. The present application accurately establishes the quantitative relationship between the French standard and the American standard, humidity and iodine removal efficiency, and accident radiation and performance decay based on a unified test bench; provides a unified scientific basis for the performance evaluation of iodine adsorbers in nuclear power plants, and improves the safety and reliability of gaseous radioactive contamination protection of nuclear facilities.

This invention relates to the field of nuclear fusion reactor materials technology, specifically to a low-activation ferritic / martensitic steel CLF-1 for nuclear fusion reactor blankets, its preparation method, and its applications. The steel's chemical composition, by mass percentage, is: Cr: 8.0-9.0%, W: 1.4-1.6%, V: 0.2-0.3%, Ta: 0.06-0.1%, Mn: 0.3-0.5%, C: 0.07-0.12%, Gd: 0.03-0.06%, N≤0.02%, with the balance being Fe and unavoidable impurities of a specific content. The microstructure is tempered lath martensite, possessing excellent room-temperature and high-temperature mechanical properties, thermal properties, and radiation resistance. This steel is effectively adapted to the extreme working environment of nuclear fusion reactor blankets and is suitable for manufacturing structural components for nuclear fusion reactor blankets.

Systems and methods are provided for separation / extraction of rare earth elements (REE), yttrium (Y), and / or radium (Ra) from phosphogypsum. Extraction of REE, Y, and / or Ra can be achieved by treating phosphogypsum with acidic aqueous media at ambient temperature in a reactor at low liquid:solid (L:S) ratios (e.g., L:S in a range of from 1:1 to 15:1), followed by ion exchange chromatography or solvent extraction of relevant anions and / or cations to ensure that the mineral hosts remain undersaturated while the gypsum remains saturated in the solvent. Grinding of the phosphogypsum (e.g., in a slurry mill) can be performed to ensure that minor mineral inclusions are fully accessed by the solvent.

The present application provides a suspension containing radioactive microspheres, a preparation method and application thereof. The radioactive microspheres include resin microspheres and a radionuclide loaded on the resin microspheres, the material of the resin microspheres includes at least one of polystyrene, polyethylene or polydivinylbenzene, or a crosslinked polymer thereof, the average particle size of the radioactive microspheres is 25-40 μm, and by controlling the non-spherical ratio of the radioactive microspheres within 5%, the distribution density of the microspheres in the tumor part is improved, the killing effect on the tumor is improved, and the incidence of adverse events is reduced.

The application discloses a method and device for preparing elements on demand by means of controllable nuclear fusion-nuclear transmutation with deuterium (hydrogen isotope) as the only raw material, belonging to the technical field of controllable nuclear physics and element synthesis, and suitable for laboratory small-batch research scenarios. The core is that under controllable conditions in the laboratory, 10-15 MeV high-energy neutrons are generated by deuterium nuclear fusion, and low-radioactivity / stable target materials are directionally bombarded, and stable products such as noble metals and rare earth elements are directionally prepared through nuclear transmutation and decay (1-50g per single preparation), the reaction by-products can be recycled, the fusion energy recovery rate is greater than or equal to 30%, and the external power supply accounts for less than or equal to 70%, which does not constitute unlimited energy output. The total floor area of the device is less than or equal to 10 m2, and the total power is less than or equal to 200 kW, and the device has the advantages of wide raw material sources, environmental friendliness, strong controllability and the like, and can be applied to the fields of noble metal preparation, new energy material synthesis, harmless treatment simulation of nuclear waste and the like, and fills the technical gap of global laboratory-level deuterium-based element preparation on demand.

A boron neutron capture therapy system and a treatment planning generation method thereof, the boron neutron capture therapy system comprising a neutron beam irradiation device, a treatment planning module and a control module. The neutron beam irradiation device is used to generate a therapeutic neutron beam and irradiate to a boron-containing drug-incorporated irradiated body to form an irradiated site during irradiation therapy. 10 B). The treatment planning module performs dose simulation calculation according to medical image data of the irradiated site and parameters of the therapeutic neutron beam generated by the neutron beam irradiation device, and generates a treatment plan, wherein the medical image data of the irradiated site comprises tissue-related information and boron 10 B) concentration-related information. The control module calls the treatment plan corresponding to the irradiated body from the treatment planning module, and controls the neutron beam irradiation device to perform irradiation therapy on the irradiated body according to the treatment plan. The boron neutron capture therapy system and the treatment planning generation method thereof can improve the accuracy of model establishment and dose calculation.

This invention relates to the field of shielding materials, and more particularly to a natural rubber-based ionizing radiation shielding material and its preparation method. The material comprises the following components: natural rubber: 60-80 parts; nano-boron carbide: 5-15 parts; nano-high-density metal powder: 15-23 parts; modified gadolinium oxide: 3-10 parts; plasticizer: 3-8 parts; vulcanizing agent: 1-3 parts; accelerator: 0.5-2 parts; antioxidant: 1-2 parts. This invention can effectively shield ionizing radiation while retaining the high elasticity and flexibility of natural rubber, and possesses good processing performance and low cost, thus meeting the diverse needs of different fields for ionizing radiation shielding materials.

The embodiment of the application relates to the technical field of nuclear reactor design, and particularly relates to a reactor, which comprises: solid fuel arranged in a reactor body, mixed gas input into the reactor body and reacted with the solid fuel to generate gaseous fuel, a reaction degree control element arranged to be capable of controlling the reaction degree of the mixed gas and the solid fuel, and a nuclear reactivity control element arranged to be used for controlling the nuclear reactivity of the gaseous fuel and the solid fuel. The embodiment of the application inputs the mixed gas into the reactor body, makes the mixed gas react with the solid fuel to generate the gaseous fuel, can simplify the supply mode and supply equipment of the reactor fuel in the mode of producing the gaseous fuel on line, improves the economy of the reactor, simultaneously, controls the reaction degree of the mixed gas and the solid fuel by using the reaction degree control element, controls the nuclear reactivity of the gaseous fuel and the solid fuel by using the nuclear reactivity control element, and is favorable for guaranteeing the reliability and stability of the reactor operation.

This application relates to the field of nuclear reactors, specifically to a reactor comprising: a reactor vessel, a core assembly, a support member, a shielding assembly, a coolant heat exchange and delivery assembly, and a reactivity control assembly. The reactor vessel forms a top opening; the support member provides support for the core assembly within the reactor vessel, and a coolant flow channel is formed between the support member and the reactor vessel for coolant flow; the shielding assembly is disposed within the reactor vessel to provide shielding for the core assembly; the coolant heat exchange and delivery assembly is configured to drive coolant to flow within the coolant flow channel to the core assembly to cool the core assembly, and to guide and cool the coolant flowing out of the core assembly; the reactivity control assembly controls the reactivity of the reactor; the support member also supports the coolant heat exchange and delivery assembly, and together with the coolant heat exchange and delivery assembly, closes the top opening. The reactor components of this application are relatively concentrated and compact, achieving a small reactor design.

This invention discloses a method for selectively separating uranium and radium, belonging to the field of solid waste treatment technology. The method utilizes modified rapeseed straw hydrothermal carbon (HTC-AA-Na). The hydrothermal carbon preparation method includes placing pretreated rapeseed straw in an aqueous acrylic acid solution for carbonization to obtain hydrothermal carbon HTC-AA rich in carboxyl functional groups; then, HTC-AA is placed in a NaOH solution and stirred, and the H groups in the carboxyl groups are removed through cation exchange. + Replaced with Na + HTC-AA-Na was obtained. When NaNO3 was added to a mixed solution containing uranium and radium, HTC-AA-Na selectively adsorbed uranium while hardly adsorbing radium; when EDTA-2Na was added to the same solution, HTC-AA-Na selectively adsorbed radium while hardly adsorbing uranium, thus achieving selective separation of uranium and radium. This invention provides a green and sustainable technical path for the harmless treatment of uranium tailings.

To provide a radioactive nuclide production system and a radioactive nuclide production method which improve cooling performance of a target material including a material nuclide, suppress contamination of impurities to the target material, and are capable of efficiently producing a radioactive nuclide.SOLUTION: A radioactive nuclide production system 1 comprises: a particle beam irradiation device 10 which generates a particle beam 11; and a target 20 which generates radioactive nuclide by irradiation of the particle beam 11. In the target 20, a target material 21 including a material nuclide generating the radioactive nuclide is mounted on a target cooling plate 22 cooling the target material 21. In a radioactive nuclide production method, the target material 21 including the material nuclide forms the target 20 mounted on the target cooling plate 22 cooling the target material 21, and the target 20 is irradiated with the particle beam 11 or bremsstrahlung while cooling the target material 21 by the target cooling plate 22 to produce radioactive nuclide.SELECTED DRAWING: Figure 1

This invention discloses a monolithic, high-performance integrated nuclear and chemical gas detector, belonging to the field of nuclear and chemical detection technology. The invention includes a nuclear radiation detection module, a chemical gas detection module, and a central control and processing module. The nuclear radiation detection module and the chemical gas detection module are integrated and both connected to the central control and processing module. The nuclear radiation detection module includes a sampling module and a nuclear data output module, while the chemical gas detection module includes a chemical gas sampling module, a gas analysis module, and a gas data output module. This invention achieves integrated detection of both nuclear radiation and chemical gases. Through modular design and central control coordination, it enables simultaneous acquisition and analysis of nuclear radiation and chemical gases, achieving efficient and integrated monitoring of nuclear and chemical substances. Simultaneously, the integrated detection modules significantly reduce size and power consumption, making it more suitable for portable applications and improving the efficiency and comprehensiveness of early warning in hazardous scenarios.

The invention refers to a method of separating, and thus recovering, gadolinium from a solution comprising gadolinium-based contrast agents via electrolysis of the solution, thereby dissociating gadolinium ions from the gadolinium-based contrast agents, and then by precipitating the gadolinium ions to obtain gadolinium in solid form, which can be later separated from the solution.

This invention discloses a method and system for the operation and maintenance of a shielded pump in a nuclear reactor experimental setup. By monitoring the stress of devices related to the shielded pump within the experimental setup and utilizing a stress assessment model for the shielded pump, the invention enables a clear, comprehensive, and real-time diagnosis of the pump's operating status. This allows for more rational intervention in preventative partial and full disassembly inspections of the shielded pump, enhancing the targeted nature of maintenance interventions, saving manpower and time costs, and increasing the safety and economy of shielded pump operation and maintenance.

This invention provides a suspension containing radioactive microspheres, a method for preparing the same, and its applications. The radioactive microspheres comprise resin microspheres and a radionuclide loaded on the resin microspheres. The resin microspheres are made of at least one of polystyrene, polyethylene, or divinylbenzene, or a cross-linked polymer thereof. The average particle size of the radioactive microspheres is 25 μm to 40 μm. By controlling the non-spherical ratio of the radioactive microspheres to within 5%, the distribution density of the microspheres in the tumor region is increased, thereby enhancing the tumor-killing effect; furthermore, it can reduce the incidence of adverse events.

This invention belongs to the field of disinfectant technology, specifically relating to a TCCA instant effervescent tablet and its preparation method. A TCCA instant effervescent tablet, by weight percentage, is prepared from the following components: 35-50% trichloroisocyanuric acid, 8-15% urea, 10-18% anhydrous citric acid, 8-14% sodium bicarbonate, 3-8% modified starch disintegrant, 2-6% disodium hydrogen phosphate, 1-4% sodium dodecyl sulfate, 0.5-2% disodium ethylenediaminetetraacetate, 1-3% hydroxypropyl methylcellulose, 0.5-1.5% magnesium stearate, with talc to bring the total to 100%. This invention addresses the problems of existing disintegrants by preparing a modified starch disintegrant through grafting and cross-linking modifications. Experimental results show that effervescent tablets using the modified starch disintegrant of this invention have a disintegration time of only 65-112s in pure water, and the increase in disintegration time in hard water is small. After accelerated testing at 54℃ / 14d, the effective chlorine reduction rate is only 2.2-5.5%, and the increase in disintegration time is small. This indicates that the grafted and cross-linked dual-modified structure has good stability and can maintain excellent disintegration performance after heat storage.

This invention discloses a method for uranium extraction from low-concentration uranium liquid using flotation extraction. The method uses an extractant emulsion saponified with alkali as the flotation reagent, which is added to the low-concentration uranium liquid. The pH of the solution is adjusted to 4.0-6.5, and the mixture is thoroughly mixed. An organic receiving phase is spread on the surface of the solution, and air bubbles are introduced to allow the solution to float to the surface. A uranium-loaded organic phase is obtained by separation. The obtained uranium-loaded organic phase can be directly back-extracted, or this uranium-loaded organic phase can be repeatedly used as the organic receiving phase to process the low-concentration uranium liquid multiple times, with uranium continuously enriching into the organic phase to obtain a uranium-rich organic phase. Finally, the uranium-rich organic phase is back-extracted to obtain a high-concentration clean uranium liquid. This invention has the advantages of simple operation, low reagent consumption, and high resource recovery rate, making it suitable for industrial application.

A container for transporting and / or storing radioactive materials, comprising a body delimiting a cavity for housing the radioactive materials, the container including at least one radiological protection enclosure formed by enclosure wall elements, each enclosure delimiting with its elements a single space wherein a radiological protection device (28) comprising a plurality of prefabricated blocks arranged in succession along a given direction, so as to face each other, is disposed. At least several blocks are each held on at least one of the elements by at least one holding member cooperating with this block and with the at least one of the associated enclosure wall elements.

The application discloses a radioactive solid waste discharge statistical analysis method and system based on multi-source tracking single data fusion. The method comprises the following steps: S1, providing a tracking single standard template and collecting radioactive solid waste discharge data; S2, settling and summarizing the radioactive solid waste discharge data, and converting different types of radioactive solid waste into a unified estimated prepared volume according to a settlement coefficient based on a rule engine; S3, generating a curve graph and trend analysis according to the settlement and summary result; S4, recording all operation processes; and S5, abnormal early warning and threshold management. The application is not dependent on semantic analysis, adopts a template+rule engine mode, is more stable and simple to maintain, provides a trend graph and prediction function, assists nuclear power plant resource planning, and has strong scalability and supports waste type addition and refinement.

The application discloses a method for preparing uranium tetrafluoride based on a reduction hydrogen fluoride reaction, and is applied to a system for preparing uranium tetrafluoride based on a reduction hydrogen fluoride reaction; steps are as follows: 1, thermal decomposition of ammonium fluoride; 2, separation of ammonia and hydrogen fluoride in mixed gas; 3, reduction hydrogen fluoride reaction of uranium oxide; 4, recycling of tail gas. The method has the advantages that: the uranium oxide is reduced and hydrogen fluoridated by adopting a three-stage fluidized bed structure to prepare uranium tetrafluoride, compared with an existing 'uranium purification-uranium conversion' process, continuous control of production can be realized, and the process flow is greatly shortened.

This invention provides a passive pressurized water reactor (PWR) nuclear island main building and nuclear power plant. The main building comprises a reactor building, a nuclear auxiliary building, and a nuclear island electrical building. The reactor building provides containment space for the nuclear fission reaction; the nuclear auxiliary building houses radioactive materials and non-radioactive materials preferably placed close to radioactive areas; the nuclear island electrical building houses non-radioactive materials. In a top view, the main building has a rectangular outline. The nuclear auxiliary building and the nuclear island electrical building surround the reactor building, with the reactor building located in the central region of the rectangle. The nuclear auxiliary building and the nuclear island electrical building are located on either side of the reactor building, and their outer outlines together form a rectangle. This passive PWR nuclear island main building and nuclear power plant improves the compatibility of the building layout structure.

This invention belongs to the field of scintillator material preparation technology, specifically relating to a method for purifying high-purity cesium iodide and its application in the preparation of cesium iodide scintillator screens. The method involves sequentially mixing and heating cesium iodide-containing waste and residual materials to dissolve them, filtering and removing impurities, adsorption and removing impurities, concentrating and temperature-controlled crystallization, centrifuging, drying, and vacuum packaging. This process comprehensively removes solid particulate impurities, elemental iodine, and various trace metal ions from the raw materials, effectively preventing oxidation and deterioration during purification. The method is simple to operate, highly industrially feasible, and can purify waste materials to a high purity level that meets the requirements of high-performance applications. It allows for the reuse of waste materials at the same purity level, reducing production costs, alleviating raw material supply pressure, and complying with environmental protection requirements. Applying the high-purity cesium iodide obtained by this method to the preparation of cesium iodide scintillator screens using a vacuum thermal evaporation process ensures the excellent performance of the scintillator screens, making them suitable for high-end fields such as medical imaging and non-destructive testing.