446 results about "Radiogenic nuclide" patented technology

The invention discloses an ocean in-situ radioactivity monitoring method, which is characterized in that feature extraction is carried out based on a maximum likelihood expected value maximization (MLEM) algorithm, and spectrum unfolding operation is carried out by adopting a multi-layer perceptron. The method comprises the following steps: firstly, measuring FWHM related parameters of a CeBr3 detector by adopting a standard source, simulating the response of the detector to gamma rays in a marine environment based on a Monte Carlo method, and establishing a full-spectrum response matrix H in a range of 0-2048keV; based on the response matrix H, reconstructing a gamma energy spectrum y obtained by the detector by using an MLEM iterative algorithm; and the reconstructed spectrum is used as the input of a multi-layer perceptron (MLP), and qualitative and quantitative analysis of radionuclides in the energy spectrum is realized. According to the method, the discrete characteristic peak signal is convolved into the single path address corresponding to the characteristic energy, so that the accuracy of the MLP model in the spectrum unfolding process is effectively improved.

The invention discloses a nuclear accident emergency three-layer coupling network model and a construction method, and relates to the crossing field of nuclear accident emergency management and complex system risk simulation. The nuclear accident emergency three-layer coupling network model comprises an RMN layer which is used for simulating the dynamic diffusion process of radionuclides leaked by a nuclear accident in the atmosphere; the IN layer is used for constructing a spatial distribution and dependency relationship network of infrastructures; the EN layer is used for topologically defining a command or cooperation relationship of the emergency organization; according to the three-layer network coupling mechanism, when RDR of a middle grid is larger than or equal to RDRmax-1, IN-layer nodes in the middle grid at the same position are marked to be in a damaged state; and quantifying the bidirectional dependence density of the IN layer and the EN layer through the coupling strength epsilon. The method is used for predicting the direct risk of radioactive contamination in a nuclear accident and the secondary risk caused by emergency evacuation, revealing the risk evolution law, quantifying the risk diffusion scale, determining key influence factors, and providing a theoretical basis for formulating an emergency strategy.

The invention relates to a method for evaluating the radiation dose of special public nearby a nuclear facility, and the method comprises the steps: estimating the concentration distribution of radionuclides in air, surface soil and a biological medium nearby the nuclear facility through collecting the radiation dose evaluation parameters of a target special public and combining an environment quality evaluation model; based on the data, the annual effective dose of the target special public at the nuclear facility boundary and the living quarter is calculated, and a plurality of ways such as inhalation internal irradiation, air immersion external irradiation, surface deposition external irradiation and ingestion internal irradiation are considered. In addition, a dynamic calculation method is also introduced, especially for the influence of year-by-year change of the cumulant of radionuclides deposited on the earth surface, factors such as decay, resuspension and infiltration into underground water of nuclides are fully considered, and the accuracy and foresight of evaluation are improved. Accurate and comprehensive technical support is provided for radiation protection of the special public around the nuclear facility, and the requirements of the nuclear facility public for dose monitoring and risk management are met.

The invention provides a multi-nuclide automatic separation and detection system and method, and particularly relates to the technical field of radioactive substance detection. The detection system comprises a cascade separation and regeneration unit, a dynamic regulation and control unit, an online detection unit, an eluent conveying unit and a control system, and the eluent conveying unit is matched with the cascade separation and regeneration unit to separate and purify radionuclide and regenerate resin. The dynamic regulation and control unit is used for controlling a flow path between the eluent conveying unit and the cascade separation regeneration unit and feeding back changes of pressure, temperature and pH value in the flow path in real time, and the online detection unit is used for detecting concentration of nuclides in separation liquid and feeding back monitoring data to the control system. And the control system controls the operation of each functional unit and automatically plans a separation path according to a measurement requirement to form a separation-detection-optimization closed-loop process. The device is suitable for the fields of nuclear facility liquid effluent detection, environmental water radioactivity detection and the like, and has high efficiency, environmental protection property and intelligence.

The invention discloses a Ga-68 activity direct measurement method which comprises the following steps: S1, preparing Ga-68 samples including a measurement sample and a blank sample; s2, establishing a mathematical model of the detection efficiency of the Ga-68 sample, including the detection efficiency epsilon D that double tubes of a single particle conform to D and the detection efficiency epsilon T that three tubes of the single particle conform to T; s3, establishing measurement equations based on the relationship between the measured total count and the mathematical model of the Ga-68 sample detection efficiency, including the measurement equations corresponding to the total detection efficiency epsilon total, D of the sample under the condition that double tubes conform to D and the measurement equations corresponding to the total detection efficiency epsilon total, T of the sample under the condition that three tubes conform to T; and S4, calculating the Ga-68 activity based on the measured sample count and radionuclide detection efficiency. The method provided by the invention meets the requirements of simple and convenient sample preparation, recyclable nuclide and low uncertainty.

The present invention belongs to the technical field of radiopharmaceutical chemistry and nuclear medicine diagnosis and treatment and relates to a radiolabeled FAPI complex, a preparation method therefor, and a use thereof. The general structural formula of FAPI capable of realizing radionuclide labeling is as shown below: wherein R is a nitroimidazole group or -COOH; X is a chelating group or a chelating structure chelated with a radionuclide; L1 and L2 are linking groups; and A is an adjusting atom of an FAP-targeting group, and all the A atoms are H atoms or F atoms. The synthesis method for a radionuclide-labeled precursor of the present invention is simple and suitable for various nuclide labels, the labeling method is convenient, and the label has high radiochemical purity and good stability. The radiolabeled FAPI complex shows high affinity and specificity for FAP in a cell experiment and a high tumor uptake and target to non-target ratio in a tumor-bearing mouse, and is a potential FAP-targeting radioactive diagnosis and treatment drug.

To provide a radiopharmaceutical production apparatus and the like capable of reducing the generation amount of radioactive waste.SOLUTION: A radiopharmaceutical production apparatus 51 comprises: a cassette 20 which has a reagent supply port, a radionuclide supply port, a drug solution outlet, a reaction unit, a purification column, and flow paths respectively connecting the reagent supply port and the reaction unit, the radionuclide supply port and the reaction unit, the reaction unit and the purification column, and the purification column and the drug solution outlet; and a cassette mounting unit to which the cassette 20 is attachable and detachable.SELECTED DRAWING: Figure 3

Systems and methods for generating radionuclides, such as radium-224, are disclosed herein. Systems herein may include a first cartridge having a first opening, a second opening, and a chamber therebetween having a first resin having affinity for thorium-228 and bismuth-212; a second cartridge having a first opening, a second opening, and a chamber therebetween having a second resin having affinity for thorium-228, a third cartridge having a first opening, a second opening, and a chamber therebetween comprising a third resin having affinity for thorium-228 and bismuth-212, and a fourth cartridge having a first opening, a second opening, and a chamber therebetween having a fourth resin having affinity for lead-212; wherein a continuous flow path is formed from a top of the first cartridge though the second cartridge, through the third cartridge, and to a bottom of the fourth cartridge during system use.

The invention discloses a method and device for reusing boric acid in waste liquid of a nuclear power plant, and the method comprises the following steps: carrying out the pretreatment of radionuclides in the waste liquid of the nuclear power plant, and obtaining first radioactive waste liquid; carrying out nanofiltration on the first radioactive waste liquid to respectively obtain separated first produced water and first concentrated water; the first produced water is subjected to first boron concentration through reverse osmosis, separated second produced water and second concentrated water are obtained, the second concentrated water is a boric acid concentrated solution, and the second concentrated water is reused in the nuclear power plant; the second produced water is subjected to second-stage boron concentration through second-stage reverse osmosis, third produced water and third concentrated water which are separated are obtained, and the third produced water is produced water meeting the preset discharge condition. According to the method, deep purification treatment is conducted on the waste liquid of the nuclear power plant, resource recycling and reuse of boric acid are achieved while the emission requirement is met, and the reuse concentration of boric acid of different reactor types can be met by adjusting technological parameters.

The invention relates to the technical field of radionuclide separation, in particular to an Eu (III) / An (III) separation method based on a zirconium phosphonate-based metal organic framework adsorbent, which comprises the following steps: adding the zirconium phosphonate-based metal organic framework adsorbent into an Eu (III) and / or An (III) nuclide-containing aqueous solution with the acid concentration of 0.001-3M to carry out adsorption reaction, and then carrying out solid-liquid separation, wherein the zirconium phosphonate-based metal organic framework adsorbent is a zirconium phosphonate-based metal organic framework material exchanged by metal cations M; the metal cation M is selected from one of alkali metal ions, alkaline earth metal ions or divalent transition metal ions. According to the invention, the interlayer chemical properties of the material are regulated and controlled through ion exchange, so that the adsorbent has ultrahigh selectivity, rapid adsorption kinetics, high adsorption capacity and excellent structural stability on trivalent actinide / simulated nuclides in extremely high-acid, high-salt and intense-radiation environments; the key problems of low separation efficiency and poor stability in a high-acid complex system in the prior art are solved, and the method is suitable for treating high-level liquid waste.

Systems and methods for generating radionuclides, such as radium-224. Systems herein may include a first cartridge having a first opening, a second opening, and a chamber therebetween having a first resin having affinity for thorium-228 and bismuth-212; a second cartridge having a first opening, a second opening, and a chamber therebetween having a second resin having affinity for thorium-228 and bismuth-212, a third cartridge having a first opening, a second opening, and a chamber therebetween comprising a third resin having affinity for thorium-228 and bismuth-212, a fourth cartridge having a first opening, a second opening, and a chamber therebetween having a third resin having affinity for lead-212; wherein a continuous flow path is formed from a top of the first cartridge though the second cartridge, through the third cartridge, and to a bottom of the fourth cartridge during system use.

Disclosed example radiographic source exposure devices include: a radiographic source capsule having a radionuclide; a radiographic shield; a channel within the radiographic shield, the channel having a first end and a second end; and a replaceable tube configured to guide a radiographic source capsule between a stored position and an exposed position in which at least a portion of the radiographic source capsule is exposed to an exterior of the radiographic shield.

The invention provides an online monitoring device and method for radionuclide in gaseous effluent of a nuclear power plant, and particularly relates to the technical field of radionuclide detection. The device comprises a gas path system, an optical system and an electric control system, wherein the gas path system is used for collecting gas to be detected; the optical system comprises a laser ring-down cavity and a laser detection module, the laser ring-down cavity is communicated with the gas path system and used for containing gas to be detected, the laser detection module comprises a light source module and a photoelectric detector, the light source module is used for generating laser beams, and the laser beams can enter the laser ring-down cavity and are transmitted in the laser ring-down cavity; the photoelectric detector is arranged at a corresponding position of the laser ring-down cavity and is used for detecting the intensity of the laser beam transmitted from the laser ring-down cavity; the electric control system is electrically connected with the laser detection module and used for controlling and processing transmission signals of the laser detection module. According to the invention, on-line monitoring of radionuclides can be realized, and the problem that secondary hazardous wastes are generated by a traditional method is solved.

The invention discloses a comprehensive evaluation method for soil pollution in a decommissioned uranium mine area, relates to the technical field of environmental evaluation, and solves the technical problem that a soil pollution evaluation method is not clear due to lack of soil pollution data in an existing decommissioned uranium mine area. According to the technical scheme, the method is characterized in that the influence of radionuclides, environmental ecology, environment and biological effects in the decommissioned uranium mine area soil on the soil quality is comprehensively considered, key influence factors influencing the uranium mine area soil quality can be screened out, and the problems that existing data is lack, and a research method is not clear are solved; and a basis is provided for subsequent sustainable utilization of soil and environmental health.

The application provides a radionuclide nanohydrogel microneedle and a preparation method and application thereof, and belongs to the field of nanobiomedicine. 32 P-labeled calcium phosphate, which is formed by a biomimetic reaction of Ag2S@Ca 32 P nanoparticles, and a base layer which is a base matrix and comprises hyaluronic acid 32 P nanohydrogel microneedle can release Ag2S nanoparticles in response to a weakly acidic tumor environment, and under the irradiation of near-infrared light, the photothermal conversion efficiency can effectively kill bacteria, effectively improve the tumor hypoxic environment, enhance the treatment effect of radionuclides, play an antibacterial and wound healing promoting role in melanoma defect repair, and maximize the inhibition of tumor growth and tumor recurrence.

A dual-radionuclide production target includes a first radionuclide target that produces a first radionuclide when irradiated by a particle beam from a radionuclide production system, and a second radionuclide target that produces a second radionuclide when irradiated by the particle beam that has passed through, and been degraded by, the first radionuclide target. With this design, the first radionuclide target serves two purposes: it reduces the beam energy to a suitable energy for isotope production on the second radionuclide target, and in-doing so becomes activated with a medically relevant radionuclide that can be easily extracted. In this way, in a single production run can produce two or more radionuclides.

The utility model relates to a dynamic measuring device for activity of medical radioactive waste liquid, in particular to the field of radiation monitoring, which comprises a sample chamber and a detector. The sample chamber is provided with a liquid level detection device, an overflow detection device, a sample feeding / discharging pump and cleaning equipment; the sample chamber comprises a protection cavity and a Marinin cup-shaped sample box; the detector is positioned above the Marinin cup-shaped sample box; the liquid level detection device comprises first liquid level detection equipment arranged at the bottom end of the sample chamber and movable second liquid level detection equipment; the overflow detection device is used for detecting whether materials in the sample chamber overflow the sample chamber; the medical radioactive waste liquid decay pool is connected with the sample chamber through the sample injection / discharge pump; and a control valve is arranged between the sample feeding / discharging pump and the sample chamber. The measuring device provided by the utility model can accurately measure the activity of radionuclides in the waste liquid, and can realize that the measuring result can be traced to the national social public measurement standard through accurate measurement, so that the reliability is guaranteed.

The invention discloses a method and equipment for identifying a radon release dominant channel in a granite surrounding rock underground space, a medium and a product, and relates to the field of data processing. The method comprises the following steps: according to content information of granite surrounding rock radionuclides, obtaining space distribution characteristics of radon parent nuclides in different types of surrounding rocks, and combining the characteristics to obtain radon background characteristics of an underground space; the method comprises the following steps: acquiring the distribution characteristics of faults, fractures or flowing water fractures at high radon concentration points in underground space air, and identifying the fracture occurrence information and radon precipitation dominant channel information corresponding to potential radon precipitation dominant channels by using the distribution characteristics; and based on radon precipitation dominant channel information and in combination with underground space ventilation conditions, screening to obtain radon measurement instrument point locations to be arranged by the radon continuous monitoring system. According to the invention, accurate identification of the radon release dominant channel is realized, technical support is provided for design of a radon continuous monitoring system in an underground space, and scientific support is provided for radon protection in the underground space.

Methods for processing metal components from the nuclear industry are described. In some embodiments, a method includes processing a component having a surface contaminated with radionuclides by dissolving at least a portion of the surface with an acid solution to produce a waste solution including the radionuclides, and separating the radionuclides from the waste solution. In some embodiments, a method includes processing a component including a target radionuclide by immersing at least a portion of the component in an electrolyte bath, applying an electric current to cause electrodissolution, and electroplating the target radionuclide from the solution onto a cathode. In some embodiments, a method includes processing a metal composite component by dissolving it to produce a solution including dissolved metal and particulate material, and separating the particulate material.

The present invention belongs to the field of biomedicine and provides radioactive silica microspheres coated with biomedical polymer materials, as well as a preparation method and application thereof. The method comprises the following steps: S1, mixing silica microspheres, water, and a radioactive nuclide to obtain a radioactive silica microsphere mixture; S2, mixing the radioactive silica microsphere mixture with an alkaline solution and reacting to obtain radioactive silica microspheres; S3, mixing the radioactive silica microspheres with a biomedical polymer solution and coating the mixture to obtain radioactive silica microspheres coated with biomedical polymer materials. The preparation method provided by the present invention is simple, and the prepared radioactive silica microspheres coated with biomedical polymer materials have a high radionuclide labeling rate, a low escape rate, good radiostability, and excellent biosafety, and can be used for the embolization treatment of all solid tumors.

1. The name of the design product: radiation measuring device. 2. The use of the design product: the design product is used for monitoring the radiation dose field of the environment radiation and analyzing the radionuclide deposition characteristics. 3. The design points of the design product: in shape. 4. The picture or photo that best indicates the design points: design 1 perspective view. 5. Design 1 is designated as the basic design.

This invention relates to the field of neutron source technology, and particularly to a controllable isotope neutron source device based on ion migration and a method for generating neutrons. The device includes a casing, inside which are a target material module, a target ion migration system, a safety module, and an alpha particle source term. The target material module provides target nuclide ions. The alpha particle source term includes alpha radioactive nuclide material and target nuclide intercalation material. The safety module is composed of target nuclide intercalation material. The target ion migration system is connected to an external control field. Target nuclide ions migrate through the target ion migration system within the safety module and the alpha particle source term. This invention enables active controllability of the isotope neutron source, significantly improving safety. By ensuring sufficient contact between alpha particles and target nuclides to form a molecular intercalation structure, the (α,n) reaction efficiency is greatly improved. Furthermore, this neutron source can be used in combination with multiple modules according to application requirements, effectively increasing the neutron output flux.

The present application relates to a kind of ice fresh water product in the defrosting device in the determination of gamma radionuclide, solve the problem that ice fresh water product in the determination of gamma radionuclide, the surface of sample box with ice fresh sample is condensed water and frost problem.The circumferential outside of the sample box of the device is provided with water receiving groove;The upper surface of the upper cover of the sample box is provided with transmission top plate, the bottom surface of the sample box is provided with transmission bottom plate, the upper surface of the transmission top plate is provided with driving roller in the middle, the left and right sides of the transmission top plate are provided with upper deflection roller, the left and right sides of the bottom end of transmission bottom plate or sample box are provided with lower deflection roller, water-absorbing cloth is sequentially wound around driving roller, left upper deflection roller, left lower deflection roller, right lower deflection roller, right upper deflection roller and connected to driving roller again.The sample box of frozen water product gamma radionuclide determination can be defrosted on surface, avoid sample box outer wall to form condensed water and frost, cause influence to the internal environment of detector.

The invention discloses a coated yttria-90 loaded radioactive microsphere and a preparation method and application thereof, and relates to the technical field of radioactive microsphere preparation, and the preparation method comprises the following steps: preparing a polyvinyl alcohol microsphere by using a microfluidic technology; the preparation method comprises the following steps: labeling polyvinyl alcohol microspheres with yttrium-90 radionuclide to obtain yttrium-90 loaded radiation microspheres; and coating the polylactic acid-glycolic acid copolymer on the surface of the yttria-90 loaded radioactive microsphere to obtain the coated yttria-90 loaded radioactive microsphere. The composite radioactive microspheres with uniform particle size, stable radioactivity and controllable characteristics can be obtained, and the composite radioactive microspheres are expected to be applied to transcatheter arterial radioembolism (TARE) treatment of liver cancer.

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.

The invention discloses an electrochemical denucleation decommissioning facility oxide layer method based on a synergistic acidic electrolyte, and relates to the technical field of cleaning. The method comprises the following steps: by taking a polluted metal component as an anode and an inert electrode as a cathode, immersing the polluted metal component into a synergistic acidic electrolyte of a low-concentration acidic medium, and stripping an oxide layer on the surface of the polluted metal component until a clear grain boundary is exposed at the temperature of lower than 40 DEG C in a high-voltage direct-current electrolysis mode. By developing a synergistic enhanced acidic electrolyte system, high decontamination rate is realized in a low-concentration sulfuric acid medium, radionuclide stripping is accurately controlled to avoid matrix pitting corrosion damage, and high-valence nuclides in a compact oxide layer are deeply removed, so that the decontamination efficiency is improved. Finally, an efficient, low-damage and stable-process radioactive contamination metal treatment scheme is provided for nuclear facility decommissioning engineering. The method is suitable for removing radioactive contamination oxidation layers on the surfaces of the inner walls of stainless steel pipelines and reactor containers in nuclear facility decommissioning projects, and radionuclides are enriched in the oxidation layers.

The invention discloses a < 68 > Ge separation and purification method and an automatic system, belongs to the technical field of radionuclide separation, and aims to solve the problems of large amount of waste liquid, complex operation and difficulty in continuous production in the conventional < 68 > Ge separation method. According to the separation method in which the actinide resin column and the glucose gel resin column are connected in series, selective adsorption, impurity removal and desorption of < 68 > Ge are completed step by step by accurately controlling the acidity of the solution, an eluting agent and an eluent; finally, a set of continuous-automatic separation system is provided, the design that two sets of separation columns are connected in parallel is adopted, work-regeneration circulation alternation is achieved through program control, and full-automatic continuous production is achieved. The method has the beneficial effects that the 68Ge recovery rate is high, the product purity is high, the waste liquid amount is remarkably reduced, meanwhile, automatic operation and continuous production are realized, and a core technology and equipment support is provided for localization of 68Ge.

The invention relates to the technical field of radionuclide data analysis, in particular to a half-life period parameter generation method and device, equipment and a storage medium. Analyzing the sampling data set based on a random forest regression model to obtain a first half-life period correction coefficient; analyzing the sampling data set based on a multiple linear regression method, a preset humidity correction function and a preset storage time length correction function to obtain a second half-life period correction coefficient; according to a preset dynamic coupling relation updating model, the first half-life period correction coefficient and the second half-life period correction coefficient, generating a real-time half-life period parameter; by integrating multi-source heterogeneous data support model training, real-time half-life parameters are generated, a reliable tool is provided for food safety supervision and the like, industry development is assisted, and economic and social values are achieved.

The application provides a kind of time-frequency feature enhanced deep learning nuclide automatic identification method, comprising the following steps: nuclear radiation detector measures radionuclide spectrum, carries out spectrum two-dimensionalization;The original one-dimensional spectrum is respectively logarithmically transformed and two-dimensionalized;The obtained two-dimensional image is synthesized into two-channel image;The image obtained is extracted using a deep learning neural network model;Model training is performed to realize metric feature embedding and multi-label classification, and a trained model is obtained;Load the trained model, input the original one-dimensional spectrum data, and output the nuclide categories and their existence probabilities that may be contained in the spectrum.The application can extract deep metric features with nuclide characteristic peak time-frequency feature constraints, enabling automatic and accurate identification of complex multi-nuclides;It has high automation, strong model generalization ability and high reliability, avoids the problem of repeated conditional parameters in traditional peak searching and matching methods, and alleviates the problem of poor reliability and generalization ability of existing deep learning nuclide identification technology.

This invention discloses a method for estimating radioactive release during the preparation of radioactive metal waste, comprising the following steps: (1) conducting gamma dose rate surveys and delineating the distribution of radioactive pollution in the radioactive source area; (2) setting up monitoring points in the radioactive metal waste areas with different pollution levels, and calculating the radioactive residue in the metal waste based on the monitoring data; (3) calculating the total release of each nuclide in the metal waste based on the above-mentioned radioactive residue in the metal waste and the obtained concentration data of radionuclides in the metal waste. By conducting gamma dose rate surveys and delineating the distribution of radioactive pollution in the radioactive source area; setting up monitoring points in the delineated areas to obtain the radioactive residue in the metal waste; and calculating the total release of each nuclide using formulas, the process is simple, easy to operate, and can obtain effective data on the total release of nuclides, providing data support for the preparation of radioactive contaminated metal waste and supporting the assessment of the dose caused by contaminated metal waste to the public.