318 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 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 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 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 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.

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 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 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 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.

The invention provides a reactor nuclide density prediction method, a reactor nuclide density prediction system and electronic equipment. The method comprises the following steps: constructing a snapshot matrix of a reactor nuclide number density sequence; using a power function to constrain nuclides in the snapshot matrix to obtain a constraint matrix; wherein the index of the power function is a score; performing DMD prediction on the constraint matrix to obtain nuclide density with constraint; and reducing the nuclide density with constraints according to the power function, and determining the nuclide density according to a reduction result. According to the method, physical driving calculation and data driving modeling are combined, a constraint function related to nuclide behaviors is introduced into nuclide evolution snapshot data, and through modal extraction and inverse transformation recovery, the capturing capability of future nuclide behaviors is improved, and the prediction precision of the number density of reactor nuclides is improved. The DMD method based on the snapshot constraint can more accurately predict the nuclide number density evolution trend at the future moment, and has the advantages of stronger generalization ability and precision.

The present disclosure relates to a method for monitoring failure of coated particles in fuel elements in a core of a pebble-bed high-temperature gas-cooled reactor, which is related to the technical field of nuclear reactor engineering and includes the following steps: S11, calculating an inventory of a short-lived noble gas fission nuclide; S12, obtaining a ratio of a release rate to a birth rate of the short-lived noble gas fission nuclide based on a temperature of the fuel elements using a Booth diffusion and release model; S13, deriving a theoretical expression for an activity concentration of the short-lived noble gas fission nuclide in a primary circuit using a migration model of the nuclide in the primary circuit; S14, obtaining an experimental measurement value of the activity concentration of the short-lived noble gas fission nuclide in the primary circuit at a sampling moment by gas sampling; S15, optimally calculating a failure fraction of the coated particles in the fuel elements and a share of uranium contamination in the matrix graphite in the core based on the theoretical expression and the experimental measurement value. The present disclosure can provide key parameters for the performance and status of the fuel elements in the core, which are required for radiation safety studies, source term calculations and accident analysis of the pebble-bed high-temperature gas-cooled reactor.

The invention belongs to the field of radioactive chemical synthesis, and particularly relates to carbon-14 labeled tetrahydroprogesterone as well as a preparation method and application thereof. The carbon-14 labeled tetrahydroprogesterone provided by the invention has a structure as shown in a formula I; the carbon-14 labeled tetrahydroprogesterone provided by the invention has the advantages that the labeled sites are determined, the radiochemical purity and the chemical purity are relatively high, the nuclide carbon-14 is firmly labeled in tetrahydroprogesterone molecules and is not easy to fall off, and the requirements of tracer experiments in pharmacokinetics can be met.

Gamma camera imaging techniques, which include: Select (S101) from energy spectra recorded by a gamma camera on one or more radioactive materials, one or more energy ranges of each radioactive material, as one or more monitored energy ranges of the radioactive material; Performing (S102) an image reconstruction on the monitored energy ranges of each radioactive material among the one or more radioactive materials; Performing (S103) a normalization of the images obtained by image reconstruction; and Performing (S104) a superposition on the normalized images to produce a composite image, characterized by the fact that the selection includes: Sweeping through the energy spectra to find peaks and comparing the peaks with energies of branches of respective radioactive materials recorded in a predetermined nuclide bank to determine the one or more radioactive materials; Determine, for each radioactive material among the one or more radioactive materials, a portion of branches of the radioactive material with a relatively higher branch ratio or lower energy than one or more branches of the radioactive material to be monitored; and Determine, for each branch of the one or more branches, an energy range of the radioactive material centered on a central energy of the branch and having a width of half a height width of a peak corresponding to the branch as one of the one or more monitored energy ranges of the radioactive material.

This invention discloses a method for precipitation separation and ceramic solidification of divalent strontium and tetravalent uranium in radioactive waste liquid, relating to the field of radioactive waste liquid treatment and disposal. U in high-level radioactive waste liquid is in the form of hexavalent uranyl (UO2). 2+ U exists in the form of hydrogenation catalytic reduction, and the present invention uses hydrogenation catalytic reduction to prepare U. 4+ Sr 2+ and U 4+ ionic radius and Ln 3+ Due to the significant differences, CePO4·0.667H2O cerium phosphate, which has a stronger nuclide-containing capacity, was selected as the co-precipitation enrichment agent for Sr. 2+ and U 4+ Separation precipitant. Inexpensive raw materials such as UO2(NO3)2·6H2O, Sr(C2H3O2)2, NH4H2PO4, and Ce(NO3)3·6H2O are selected. By controlling the Sr... 2+ :U 4+ Ion ratio, reaction time, temperature, pH, and preparation of well-developed Sr crystals 2+ / U 4+ The precipitate of cerium phosphate solid solution with high solid content is avoided by using spark plasma solid-state sintering technology to prevent grain agglomeration and adhesion during high-temperature dehydration and crystal transformation of cerium phosphate. Furthermore, pressure sintering can promote further growth of nanocrystals and promote U... 4+ and Sr 2+ Homogenization within the monazite ceramic solidified body enables a one-step preparation of monazite ceramic solidified body from a lanthanite solid solution.

The application discloses a kind of nuclear power plant power operation mode pipeline deposition source item background evaluation system and method, this method includes: S1.the activity ratio ε of activated corrosion product nuclide i is measured i , pipe surface dose rate D and the activity concentration C of nuclide i in primary coolant i ;S2.modeling and calculating the effective dose rate D of pipe surface by activated corrosion product in deposition contribution d ;S3.modeling and calculating the surface activity of activated corrosion product nuclide i in deposition source item;S4.modeling and calculating the total count rate CPS or total absorbed dose rate Gy / h of pipeline deposition source item to detector.The nuclear power plant power operation mode pipeline deposition source item background evaluation system and method of the application, the activity ratio of activated corrosion product nuclide i, pipe surface dose rate, the activity concentration of nuclide i in primary coolant are measured, modeling and calculating the total count rate CPS or total absorbed dose rate Gy / h of pipeline deposition source item to detector, realize the evaluation of deposition source item background under the power operation mode of unit.

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.

The present invention relates to a method for evaluating the solidification treatment of radioactive waste and a method for evaluating the feasibility of alternative nuclides. The method for evaluating the feasibility of alternative nuclides includes: obtaining at least one actinide nuclide in the solidification treatment of radioactive waste; preparing a single crystal of trivalent coordination salt of the actinide nuclide in a simulated coordination environment for the solidification treatment of radioactive waste; preparing a single crystal of trivalent coordination salt of the lanthanide nuclide in the same simulated coordination environment; comparing the crystal structures of the single crystals of trivalent coordination salts of the actinide nuclide and the lanthanide nuclide; and obtaining the evaluation result of the feasibility evaluation according to the comparison result. By preparing single crystals of trivalent coordination salts of actinide and lanthanide nuclides under the same conditions, comparing the crystal structures of the single crystals, and based on the comparison result, it can be determined whether the lanthanide nuclide can replace the actinide nuclide for use in the research on the evaluation of the solidification treatment of radioactive waste.

The present application belongs to the technical field of radiation protection and pressurized water reactor source term calculation, and particularly relates to a pressurized water reactor pellet-cladding gap fissile product source term fast calculation method. The method comprises the following steps: step 1: establishing a first-order kinetic balance equation of the fissile product in the fuel rod; step 2: using Chebyshev rational approximation method to solve the accumulation of the nuclides in the decay chain corresponding to different mass numbers in the fuel rod; step 3: calculating the release production ratio of the fissile product nuclides; step 4: judging the release path of the target nuclides from the fuel pellet to the pellet-cladding gap; and step 5: judging the fissile product source term in the pellet-cladding gap according to the release path of the fissile product. The present application has the beneficial effect that the fissile product source term in the fuel rod is calculated in detail, and the fissile product source term in the pellet-cladding gap can be obtained, which is used for the release source term calculation when the fuel cladding is damaged subsequently.

The present invention provides a radionuclide production system and a radionuclide production method with which it is possible to produce radionuclides efficiently by improving the cooling of a target material containing a raw material nuclide and reducing contamination of the target material with impurities. A radionuclide production system comprises a particle beam irradiation device that generates a particle beam, and a target that generates radionuclides through irradiation with the particle beam. In the target, a target material containing a raw material nuclide that generates radionuclides is mounted on a target cooling plate for cooling the target material. The radionuclide production method comprises: forming a target in which a target material containing a raw material nuclide is mounted on a target cooling plate for cooling the target material; and producing radionuclides by irradiating the target with a particle beam or bremsstrahlung radiation while cooling the target material using the target cooling plate.

The present invention belongs to the technical field of nuclide adsorption materials, and specifically discloses a Ca-ZrP-1,8-octanediamine composite material, its preparation method, and use. The composite material is synthesized by intercalation and ion exchange reaction using hydrothermal ZrP as a substrate. Compared with the prior art, the process for preparing the Ca-ZrP-1,8-octanediamine composite material is simple, economical, and does not produce intermediate products that pollute the environment. The prepared Ca-ZrP-1,8-octanediamine composite material has better adsorption performance than pure zirconium phosphate, stable performance in a wide pH range and a high correlation coefficient, and is more effective for Sr at pH = 1. 2+ The adsorption capacity reaches 34.52 mg / g, which is much higher than that of pure ZrP and can be used for the adsorption of radionuclide strontium.

The application provides a pulse neutron element scanning and analyzing system for extraterrestrial exploration, which comprises a shell, a support, a neutron generator, a shielding body and a gamma spectrometer, the support is installed in the shell, the neutron generator, the shielding body and the gamma spectrometer are all installed on the support, and the shielding body is located between the neutron generator and the gamma spectrometer. The neutron generator, the shielding body and the gamma spectrometer are integrally installed on the support, the controllable neutron source is combined with the gamma spectrometer, the stratum is activated by the controllable pulse neutron generator, the element components are detected by gamma analysis, or the gamma generated by the natural radioactive nuclides and the cosmic background radiation is directly detected, element component analysis is carried out, the element measurement precision is improved, the detection time is shortened, more kinds of elements can be measured, and the application provides strong support for the measurement of more kinds of elements, and is suitable for lunar geology and resource exploration.