385 results about "Nuclide" patented technology

The invention relates to the technical field of unmanned aerial vehicle-mounted nuclear radiation detection, and provides an unmanned aerial vehicle-mounted nuclear radiation detection method and system based on deep learning, and the method comprises the steps: collecting a gamma energy spectrum signal, and carrying out the digitization of the gamma energy spectrum signal, and obtaining original energy spectrum data; preprocessing the original energy spectrum data to obtain preprocessed energy spectrum data; constructing a nuclide identification model based on a deep learning network, and performing nuclide identification on the preprocessed energy spectrum data to obtain an original nuclide identification result; performing post-processing analysis and refinement on the original nuclide identification result to obtain a nuclide identification result; and fusing the nuclide identification result with the geographic position information to obtain a nuclear radiation pollution distribution diagram, and setting a multi-level early warning mechanism based on nuclide types and intensity. According to the invention, the precision of nuclide identification in a complex environment is improved, and high-sensitivity detection of various nuclides under the condition of low signal-to-noise ratio is realized.

The invention belongs to the technical field of nuclear radiation measurement, and relates to a liquid flash TDCR multi-nuclide beta spectrum analysis method and system based on artificial intelligence. The method comprises the following steps: constructing a numerical model of a liquid flash detector spectrometer by using a Monte Carlo technology to simulate the energy spectrum response of a single nuclide under different quenching conditions; constructing a training database by adopting a parameterized data synthesis algorithm; constructing a multi-task mixed spectrum analysis neural network model; carrying out model training by utilizing the constructed database; and inputting an actual measurement spectrogram, and outputting the multi-nuclide absolute activity, the detector efficiency and the decomposition energy spectrum contribution curve in real time by using the trained neural network model. According to the method, differential distribution characteristics and quenching response curve characteristics of the beta continuous energy spectrum are creatively fused, a neural network architecture with physical mechanism constraints is constructed, and liquid flash TDCR multi-nuclide energy spectrum characteristic decoupling and accurate activity solving are achieved under the unknown quenching condition.

The invention provides an ore grade analysis method. The ore grade analysis method comprises the following steps: generating an X-ray beam by utilizing a radiation source; x-ray beams are utilized to irradiate the ore, target element nuclides in the ore are excited to generate a nuclear resonance fluorescence effect, target characteristic gamma rays are emitted, and the energy of the X-ray beams is higher than the excitation energy of the target element nuclides; when the X-ray beam is utilized to irradiate the ore, a standard sample containing a reference nuclide is irradiated, the reference nuclide is excited to emit a reference characteristic gamma ray, and the reference nuclide is expressed as a nuclide with known content and known characteristic gamma ray energy; detecting the energy of a target characteristic gamma ray and the energy of a reference characteristic gamma ray; and determining the grade of the target element in the ore based on the energy of the target characteristic gamma ray and the energy of the reference characteristic gamma ray.

The embodiment of the invention relates to the technical field of nuclear detection and measurement, and discloses a double-label nuclide analysis method and device and a program product, and the method comprises the steps: measuring a standard quenching source, and obtaining an external standard source spectrum index of the standard quenching source; drawing a correction curve by taking the external standard source spectrum index as an independent variable; measuring the double-labeled sample under external gamma source high voltage, and obtaining and calculating an external standard source spectrum index of the double-labeled sample through external gamma source spectrum stripping; determining the counting efficiency of low-energy nuclide, the counting efficiency of high-energy nuclide, the spectrum index of the low-energy nuclide sample and the spectrum index value of the high-energy nuclide sample corresponding to the double-labeled sample under the current quenching level according to the external standard source spectrum index of the double-labeled sample and the correction curve; the method comprises the following steps: measuring a double-labeled sample under double-labeled high pressure to obtain a full-spectrum counting rate and a full-spectrum sample spectrum index of the double-labeled sample; the low-energy nuclide activity and the high-energy nuclide activity of the double-labeled sample are calculated based on the parameters, so that the operability is higher, the accuracy is higher, and the application range is wider.

The invention discloses a nuclear leakage radioactive substance diffusion situation inversion method and device, and relates to the technical field of nuclear emergency response. The method comprises the following steps: acquiring initial source item parameters and meteorological data, and obtaining a nuclide concentration predicted value through a Gaussian puff model according to the initial source item parameters and the meteorological data; acquiring a nuclide concentration observation value corresponding to the nuclear leakage area, and performing Kalman filtering fusion processing according to the nuclide concentration prediction value and the nuclide concentration observation value to obtain a nuclide optimal concentration field and analysis error distribution; adjusting parameters of a Gaussian diffusion mode according to the analysis error distribution to obtain an adjusted Gaussian diffusion mode, and obtaining target source item parameters based on the adjusted Gaussian diffusion mode and the optimal concentration field; and performing iteration by taking the target source item parameter as an initial source item parameter until a preset requirement is met, and taking the source item parameter obtained when iteration is ended as a final source item parameter. The method can dynamically respond to the sudden change of atmospheric stability, and improves the prediction precision of the nuclear hazard area range.

The invention discloses a radioactive waste liquid treatment method and system for the nuclear medicine department, and belongs to the technical field of waste liquid treatment of the nuclear medicine department. The method comprises the steps that waste liquid is discharged into a pre-storage container, concentration detection and nuclide type detection are conducted, and the waste liquid with the concentration detection result lower than a specified limit value is directly discharged into a medical waste water pipeline; other waste liquids are classified and discharged into the corresponding decay tanks according to the half-life periods of nuclide types, the storage time of the waste liquids in the decay tanks is different according to different half-life periods, and the waste liquids in the corresponding decay tanks are discharged in time after the storage time is reached. According to the invention, the amount of the waste liquid actually entering the decay tank is reduced from the source end, and the waste liquid is discharged into the decay tank in a classified manner based on nuclide types, so that the storage time of the waste liquid corresponding to nuclides with short half-life periods in the decay tank can be shortened, the discharge efficiency of most of the waste liquid is improved, and the amount of the waste liquid in the decay tank is effectively reduced.

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 fuel element damage monitoring method and device based on delayed neutron energy spectrum measurement. The method comprises the following steps: acquiring a delayed neutron zone for fuel element damage monitoring; according to the delayed neutron partition and the delayed neutron partition multi-column energy spectrum measurement principle, a moderator for fuel element damage monitoring is obtained; according to the moderation body and a discrimination threshold automatic identification algorithm, determining a discrimination threshold value of fuel element damage monitoring; and according to the discrimination threshold value, a few-channel spectrum unfolding algorithm and an energy spectrum peak searching algorithm, identifying neutron energy spectrum change conditions and delayed neutron precursor nuclides of fuel element damage monitoring. According to the invention, on-line real-time measurement of the delayed neutron energy spectrum can be realized, and the neutron energy spectrum change condition can be determined; according to the invention, when the delayed neutron precursor nuclide is identified, the occurrence of effective fuel element damage is judged, the online monitoring of the fast reactor fuel element damage is realized, and the rapidness and timeliness of fuel damage monitoring are improved.

The invention discloses a monocrystalline silicon neutron irradiation monitoring method and system based on a vanadium self-powered detector, and belongs to the field of nuclear engineering. The monocrystalline silicon neutron irradiation monitoring method based on the vanadium self-powered detector comprises the following steps: calculating a corresponding relation between the burnup depth and nuclide components of a fuel assembly and a corresponding relation between axial burnup distribution of the fuel assembly and fuel circulation and the burnup depth; calculating to obtain a corresponding relationship between the thermal neutron fluence rate of the irradiation pore channel and the burnup cycle and the burnup depth; and taking a position where the thermal neutron fluence rate of the irradiation pore channel does not exceed a given threshold value as a detector mounting area, and calibrating a corresponding relation between a vanadium self-powered detector signal mounted in the detector mounting area and the thermal neutron fluence rate of the irradiation pore channel by using an activated foil, thereby monitoring the thermal neutron fluence rate of the monocrystalline silicon by using the vanadium self-powered detector.

The invention provides an online monitoring device for multiple nuclides in liquid effluent, and particularly relates to the technical field of radioactive substance detection. The online monitoring device comprises a shell, a cascade separation module, a fluid control module, a detection module and a control system, the cascade separation module, the detection module and the control system are arranged in the shell, the cascade separation module comprises a plurality of resin columns and a plurality of separated liquid collecting bottles communicated with the resin columns, and the separated liquid collecting bottles are communicated with the resin columns. The fluid control module comprises a liquid storage device, an injection pump and a control valve, the liquid storage device is respectively communicated with the plurality of resin columns through pipelines, the injection pump and the control valve are respectively arranged on the pipelines and are used for controlling the flow direction and the flow velocity of fluid in the liquid storage device, and the detection module is used for detecting the radioactive activity of nuclide; the control system is used for controlling operation of the cascade separation module, the fluid control module and the detection module. According to the invention, synchronous monitoring of multiple nuclides is realized, the multi-nuclide monitoring efficiency is improved, and the monitoring cost is reduced.

The invention discloses a nuclear power plant reactor core fuel assembly uranium contamination fission product calculation method and system, and the method comprises the following steps: monitoring the characteristic nuclide of a loop coolant of a reactor core fuel assembly in real time, and judging whether the surface of the reactor core fuel assembly has uranium contamination; if the uranium contamination exists, calculating the uranium contamination amount of the uranium contamination distributed in the primary loop coolant on the surface of the reactor core fuel assembly; and calculating uranium contamination fission product distribution of the reactor core fuel assembly according to the uranium contamination amount. According to the method, the uranium contamination fission product distribution in the primary loop coolant is collected, so that the risk of erroneous judgment of the damage of the reactor core fuel assembly caused by uranium contamination can be remarkably reduced.

The invention discloses a nuclear power plant reactor core fuel assembly damage analysis and calculation method and system. The method comprises the following steps: calculating a typical nuclide equivalent attenuation coefficient according to fission product source item characteristics; calculating a typical nuclide release rate according to the typical nuclide equivalent attenuation coefficient; according to the equivalent attenuation coefficient of the typical nuclide and the release rate of the typical nuclide, calculating the reachable equilibrium specific activity of the typical nuclide; by analyzing typical nuclides, equilibrium specific activity can be achieved, and reactor core fuel assembly damage analysis is carried out. According to the method, the equilibrium specific activity can be reached through typical nuclides to carry out reactor core fuel assembly damage analysis, and the operation reliability of the reactor core fuel assembly is analyzed.

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.

A portable small protective workstation for alpha nuclide operation comprises a protective cabin, a transparent window and glove holes are formed in the front side face of the protective cabin, radiation-resistant operating gloves are embedded in the glove holes in a sealed mode, a tray inlet is formed in the side face of the protective cabin, and a drawer type bracket is connected to the tray inlet in a sliding mode. The drawer type bracket comprises a tray layer, a cushion layer and a cover plate, the tray layer and the cushion layer are both vertically and fixedly connected to the side face of the cover plate, the tray layer is located above the cushion layer, a high-adsorbability radioactive water absorption pad is fixedly installed on the cushion layer, a plurality of through holes are formed in the bottom face of the tray layer, and the tray layer and the cushion layer are movably installed in the protection cabin body through a tray inlet. The cover plate abuts against the outer side face of the protection cabin in a sealed mode. An air filter box is arranged above the protective cabin body, and a filter layer is arranged in the air filter box. According to the utility model, the defects in the prior art are overcome, and the organic unification of radiation protection, pollution control and experiment convenience for operators is realized.

The invention discloses a compensation method and system for continuously, rapidly and accurately measuring the radon concentration in air, and the method comprises the steps: taking the number of nuclides in a measurement cavity at the initial moment of a current sampling period as an initial value, and obtaining a time-varying function of the number of nuclides in the measurement cavity in combination with a gas exchange and nuclide decay model; determining the theoretical total count of alpha particles in the current measurement period, and establishing a relational expression between the count of the alpha particles and the number of nuclides; based on the relational expression and the actual measurement count, the number of environmental radon atoms is obtained, and the radon concentration is calculated in combination with a radon decay constant; and finally, combining the number of radon atoms at the current sampling point and a model containing gas exchange and nuclide decay to obtain the nuclide number at the end of sampling, taking the nuclide number as the initial state of the next sampling period, switching the sampling points, and repeating the steps to realize continuous, rapid and accurate sampling. According to the method, through loop iteration, the balance waiting time is effectively saved, and continuous, rapid and accurate measurement of the radon concentration in the environment by a single radon measuring instrument is realized.

The invention provides an intelligent treatment system for radioactive waste liquid. The intelligent treatment system comprises the following steps: acquiring basic working data of a radiation subject, analyzing estimated waste liquid radiation activity and estimated waste liquid generation amount in a specified time period to construct a waste liquid collection measure for waste liquid collection, and calling a corresponding waste liquid detection mode according to the estimated waste liquid radiation activity, the waste liquid is sampled and detected in a waste liquid detection mode, the nuclide activity value of the waste liquid is determined, the waste liquid is conveyed to a treatment area of a corresponding grade based on the nuclide activity value to be subjected to active treatment, nuclide activity value change information of the waste liquid is determined according to supervision data in the active treatment process, and after the active treatment of the waste liquid is completed, the waste liquid is processed. According to nuclide activity value change information, after-treatment radioactivity of a generated treatment object is analyzed, the generated treatment object is stored, and an intelligent treatment system integrating waste liquid collection, radioactivity monitoring, separation treatment and environment-friendly emission is constructed.

Disclosed in the present invention is a fuel rod for a water-cooled reactor. The fuel rod comprises a cladding tube, an upper end plug and a lower end plug that match at two opposite ends of the cladding tube, and fuel pellets stacked in the cladding tube, wherein the fuel pellets comprise at least one first pellet; the first pellet is formed by at least two pellet units connected in an enclosed manner in a circumferential direction; and a through hole penetrating two opposite ends of the first pellet is formed in the first pellet, and is filled with a filler containing burnable poison nuclides. In the fuel rod for a water-cooled reactor of the present invention, by means of providing a hole in a pellet and filling the hole with the filler containing the burnable poison nuclides, a gas generated by the pellet during the operation of the reactor can be absorbed, thereby reducing the internal pressure of the cladding tube, reducing the risk of fuel rod failure, preventing the leakage of radioactive substances, increasing the uranium utilization rate, and providing a relatively large design margin.

The invention relates to the technical field of radiotherapy, and discloses a method for generating a boron neutron capture therapy plan, computer equipment and a storage medium. Comprising the following steps: based on current state data of neutrons, determining first section data, second section data and a current transport distance of the neutrons at a current position, and judging whether the neutrons collide or not; wherein the first cross section data is used for describing the macroscopic cross section of nuclides except boron, and the second cross section data is used for describing macroscopic cross section neutrons obtained through calculation according to the boron concentration of the current grid and the microcosmic cross section of the nuclide boron; and under the condition that neutrons do not collide and the neutrons are transmitted from the current grid to the next grid through the surface, if the two grids are made of the same material, updating the second cross section data based on the boron concentration of the next grid to calculate the dose, and finally generating a boron neutron capture therapy plan by using a dose distribution calculation result. Therefore, only the cross section data related to the boron concentration is updated in the penetrating surface, and the calculation efficiency is optimized.

The invention discloses a radioactive liquid activity detection method, device and equipment, and relates to the field of nuclear energy. The method comprises the following steps: acquiring a to-be-detected water sample, and respectively inputting the to-be-detected water sample into a detector to obtain a detector original pulse signal and a detector fast signal corresponding to the to-be-detected water sample; performing FPGA time window coincidence processing based on the detector original pulse signal and the detector fast signal to obtain a new energy spectrum corresponding to the to-be-detected water sample; obtaining a preprocessed theoretical energy spectrum template, and inputting the theoretical energy spectrum template and the new energy spectrum into the peak decomposition network model to obtain a characteristic peak area list of the target nuclide; and obtaining an efficiency curve of the detector, and obtaining an activity detection result of the target nuclear element in the to-be-detected water sample based on the characteristic peak area list of the target nuclide and the efficiency curve. According to the method, the final result is determined by combining the theoretical energy spectrum template, the original pulse signal of the detector and the fast signal of the detector, so that the spectrum unfolding error of the overlapping peak can be reduced, and the recognition accuracy of the mixed system is effectively improved.

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 provides a preparation method of a material for selective adsorption of iodine-131 nuclide, and belongs to the technical field of nuclear medicine radioactive wastewater purification treatment.The preparation method comprises the steps that graphite powder, concentrated sulfuric acid, phosphoric acid, potassium permanganate and hydrogen peroxide serve as raw materials to synthesize graphene oxide in a membrane uniform distribution type microreactor; taking the graphene suspension and Zn (NO3). 6H2O as raw materials to synthesize nanoporous graphene in a membrane uniform distribution type microreactor; and sequentially filtering the graphene oxide solution, the nanoporous graphene solution and the graphene oxide solution on a basement membrane to obtain the graphene oxide / nanoporous graphene / graphene oxide two-dimensional separation membrane. According to the preparation method of the material for selective adsorption of the iodine-131 nuclide, provided by the invention, the prepared material for selective adsorption of the iodine-131 nuclide is good in selectivity, high in adsorption capacity and strong in stability.

The invention discloses a nuclear power plant fuel operation accident source item result determination method and device and a nuclear power plant fuel operation accident alarm method and system, and relates to the technical field of nuclear industry. The method for determining the fuel operation accident source item result of the nuclear power plant comprises the following steps: acquiring a target decontamination factor corresponding to a target nuclide in a first time period and the total concentration of aerosol iodine retained in a spent pool of the nuclear power plant in a second time period; determining a first release amount corresponding to the target nuclide in the first time period according to the target decontamination factor; according to the total concentration of iodine in the aerosol, determining the share of iodine volatilized in the form of elements in the spent fuel pool in the second time period; determining a second release amount of the element iodine in a second time period according to the iodine share; and generating a fuel operation accident source item result of the nuclear power plant according to the first release amount and the second release amount. According to the embodiment of the invention, the method can improve the accuracy of determining the fuel operation accident source item result of the nuclear power plant.

The invention relates to the technical field, and discloses a radioactive aerosol real-time monitoring method and system, and the method comprises the steps: collecting real-time multi-modal data of aerosol monitoring in a target region, and obtaining the fusion features of the real-time multi-modal data; performing nuclide identification on the fusion features of the real-time multi-modal data by using a pre-trained identification model to obtain an identification result; on the basis of the recognition result, semantic attributes of nuclides of known types are obtained by judging the semantic attributes of the nuclides, and semantic attributes of nuclides of unknown types are speculated by using a zero sample learning model; and based on the attribute of the nuclide corresponding to the real-time multi-modal data, generating a suggested decision by using a predefined decision support model. Through real-time monitoring and rapid identification, risk early warning can be given out in time, a scientific basis is provided for emergency response, and the problem that sudden radiation events cannot be handled in time is solved.

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.

The invention provides a radioactive substance transportation monitoring system, and relates to the technical field of radioactive substance transportation safety monitoring. The radioactive substance transportation monitoring system comprises a transfer protection box and a monitoring unit integrated on the transfer protection box, and the monitoring unit is composed of a double-detector fusion module, a three-dimensional data linkage analysis module and an active emergency disposal module which are sequentially connected through a circuit. The double-detector fusion module comprises a GM counting tube and a scintillator detector which are arranged inside and outside the box so as to realize alpha, beta, gamma and neutron multi-nuclide coverage; the three-dimensional data linkage analysis module performs comprehensive judgment based on internal and external doses of the box, temperature and humidity and vibration data, and accurately identifies protection failure or article leakage; and the active emergency disposal module automatically starts gate closing and adsorption measures in an emergency. According to the system, the problems of monitoring blind areas, reduced protection performance, frequent false alarm, emergency passivity and the like are solved, and the transportation safety and the intelligent level are improved.

The application discloses a kind of determination methods and program products of core reactivity state, the method includes: the key state parameter in the process of core reaction is collected, key state parameter includes at least one of core power distribution parameter, coolant temperature, coolant density, control rod position, boric acid concentration and burnup;Determine the burnup step of fuel, the evolution process of fissile nuclide and fission product in core is modeled based on the burnup step of fuel and burnup model, to obtain the reactivity parameter related to the burnup of fuel, the burnup model is used to describe the nuclide transmutation and consumption logic of fuel during the operation of core;Corresponding digital twin is constructed with the reactivity state of the core, the reactivity state of the core is mapped by the digital twin based on the reactivity parameter and the key state parameter driving, can realize high-precision, real-time, full life cycle dynamic digital mapping and intelligent detection of core reactivity state.

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.