204 results about "Nuclide" patented technology

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 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 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 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.

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 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 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.

The present application relates to a kind of natural conditions under the estimation method of superuranic nuclide resuspension release source term.This application obtains the environmental characteristic parameter of the place of contaminated area, calculates the related parameter of superuranic nuclide pollution condition, calculates the surface covering layer factor of contaminated area soil, calculates the resuspension release source term of superuranic nuclide, can start from the characteristics of soil environment, nuclide characteristics, natural condition three aspects, based on the mass load theory of particulate matter, using the release flux of particulate matter in contaminated area soil and the level of radioactivity concentration in soil, the resuspension release source term of superuranic nuclide is calculated, realizes the resuspension release amount of superuranic nuclide in soil environment pollution under natural conditions is scientifically and accurately evaluated.

The invention discloses a radiation hazard source identification method with a low counting rate, and relates to the technical field of nuclear technology application. Comprising the following steps: establishing a nuclide library containing radionuclide information; setting a peak position interval corresponding to each radionuclide based on the nuclide library; carrying out statistics on the peak position count of the peak position interval and the high-energy area count of the high-energy area, and caching the peak position count and the high-energy area count obtained through statistics as environment background data; estimating parameters according to the cached environment background data; acquiring a peak position counting rate and a high-energy region counting rate in real time; according to the parameters, the peak position counting rate and the high-energy region counting rate which are obtained in real time, the probability that the peak region is abnormal is calculated; performing early warning sequential test on the abnormal probability of the peak region; and comparing a result of the early warning sequential test with a preset early warning threshold value, and if the result reaches the early warning threshold value, outputting early warning information and identifying a corresponding radiation hazard source. According to the invention, rapid early warning and hazard source identification of abnormal radiation data are realized.

The invention discloses a fission spectrometer used for fission fragment independent yield measurement. The fission spectrometer comprises a flight time detector and a screen grid ionization chamber fixedly connected to an outlet of the flight time detector. The flight time detector and the screen grid ionization chamber are provided with pump interfaces used for being connected with a vacuum pump for vacuumizing, and the flight time detector and the screen grid ionization chamber are each provided with a signal interface and a high-voltage electrical interface which are used for receiving detection signals. The signal interface and the high-voltage interface are respectively connected with the data acquisition system and the high-voltage system through lines; the method is suitable for independent yield measurement of fission fragments of multiple fission nuclides and multiple neutron energy points; according to the integrated device, online measurement of neutron-induced heavy nuclear fission can be achieved, errors caused by chemical treatment and step-by-step measurement in a traditional measurement technology are avoided, the mass resolution of fission fragments can be smaller than 1 amu, and the independent yield of the fission fragments can be obtained while excellent mass resolution is obtained.

The invention relates to a high-radioactive waste disposal simulation method and system based on multi-field coupling. The method comprises the following steps: constructing a three-dimensional finite element model of a high-radioactive waste repository and a surrounding geological medium based on physical field simulation; simulating temperature change around the high-radioactive waste repository, underground water flow velocity change and radionuclide concentration change; based on a geochemical calculation model, performing chemical reaction calculation on the distribution of the radioactive concentration of the high-radioactive waste in underground water and rock-soil media, including dissolution, precipitation, adsorption and ion exchange of the high-radioactive waste; and carrying out iterative coupling on the physical field simulation and the geochemical calculation model, and outputting a high-radioactive waste disposal simulation result which comprises a radionuclide concentration three-dimensional distribution diagram, a migration flux vector diagram and temperature field distribution. Through iterative coupling of a physical field and a chemical field, high-precision simulation prediction of nuclide migration behaviors in a high-radioactive waste long-term disposal environment is realized, and a simulation result is close to an actual migration process.

The invention belongs to the technical field of nuclear energy accident treatment, and relates to a high-temperature gas cooled reactor accident condition typical nuclide chemical form research method and system and a readable medium. Inputting the data into an accident source item calculation model to obtain the content change condition of each radionuclide and impurity element along with time; obtaining a chemical form evolution result along with temperature and a chemical form evolution result along with time through a thermodynamic method; determining a Latin hypercube sampling range, constructing an EMSAC input file, and running EMSAC calculation to obtain an element chemical form database in a multi-parameter variation range; and obtaining a correlation analysis result of the element chemical form database. The method solves the problems that in the prior art, accident source items cannot be processed, chemical forms evolve along with time and dimension explosion occurs during multi-parameter multi-change range calculation, accident source item whole-process calculation is completed for the first time, and theoretical reference is provided for subsequent radiation safety evaluation of nuclear facilities.

The application discloses a wide-energy gamma ray standard surface source and a preparation method thereof, and aims at solving the technical problems that the existing radioactive sources cannot meet high-precision measurement due to short service life, uneven distribution of radioactive sources, and self-absorption, etc. The wide-energy gamma ray standard surface source comprises a radioactive source carrier, a sealed shell and a sealed cover. The radioactive source carrier is laid on the bottom of the sealed shell, and radioactive nuclides are attached to the radioactive source carrier. 226 Ra 226 The activity of Ra is not less than 5*10 4 Bq. The sealed cover is tightly pressed on the radioactive source carrier, and the gap between the sealed cover and the sealed shell is sealed by sealing glue, so that the radioactive source carrier is sealed between the sealed shell and the sealed cover.

The invention provides a method, system, equipment and medium for identifying the number and size of reactor fuel cladding damage, and the method comprises the steps: S1, obtaining the unit crevasse release amount reference data of each nuclide based on transient release modeling experiments of different groove sizes; s2, obtaining the total release amount of nuclides based on the obtained unit crevasse release amount reference data of each nuclide; and S3, constructing a matrix reversibility constraint, and realizing the unique solution inversion of the damage parameter. The method can improve the accuracy and efficiency of fuel cladding damage detection, provides reliable method support for safe operation of a reactor, and can be widely applied to monitoring and evaluation of the safety of the fuel assembly of the pressurized water reactor nuclear power station.

The application discloses a kind of uranium nuclide automatic extraction device and method, wherein, device includes dissolving target device, acid regulator, extractor, purifier, waste liquid tank, product tank, control system and shielding box, dissolving target device, acid regulator, extractor, purifier, waste liquid tank, product tank are connected by conveying pipe and valve, liquid conveying power is provided by peristaltic pump, control system is set in shielding box, and relevant components in shielding box are electrically connected.The application also describes a method for extracting a uranium nuclide solution substantially free of impurities from an irradiated thorium target material.The uranium nuclide automatic extraction device of the present application can achieve rapid and efficient extraction and purification of uranium nuclides from irradiated thorium target material.The device has the characteristics of simple structure, compactness and automation operation, which greatly improves the production efficiency and reduces the radiation dose received by the operator.

The application discloses a kind of on-line detection device and method for resin adsorption process element, device includes shell and from top to bottom sequentially arranged in the inside of shell neutron source system, shielding body module, gamma detector system and for controlling the work and detection result analysis of each component control system;The application generates continuous neutrons or pulsed neutrons by neutron source system, generates prompt gamma rays by the interaction of continuous neutrons and resin atomic nuclei, realizes the quantitative analysis of non-fissionable elements by analyzing the energy and intensity of gamma rays, and then generates prompt fission neutrons by pulsed neutrons to excite fissile nuclides, and the quantitative analysis of fissile nuclides is carried out by analyzing the intensity of prompt fission neutrons generated.The whole detection process is simple and convenient, and the detection will not damage the test sample, and in-situ measurement and on-line analysis can be realized.

The application discloses a method for determining complete transformation of complex Am-based nitrate solution under a radioactive system, and belongs to the technical field of radioactive waste liquid treatment. The method for transforming the radioactive Am-based nitrate solution comprises the following steps: concentrating the radioactive Am-based nitrate solution to obtain nitrate concentrate solid; and then calcining the nitrate concentrate solid at 700-750 DEG C. Meanwhile, the application also provides a method for determining complete transformation of complex Am-based nitrate solution under a radioactive system. The application solves the problem of how to transform the long-life and high-toxicity transuranium element stream into an oxide; the method is beneficial to temporary storage of the element, facilitates reuse of the element in the future, and provides a method for identifying whether the transformation of the element nitrate solution is complete or not.

The invention discloses a nuclear reaction cross section data generation method and system based on machine learning. The nuclear reaction cross section data generation method based on machine learning comprises the following steps: acquiring data of nuclide characteristic indexes; the data of the nuclide characteristic indexes are used to predict a nuclear reaction cross section by using a pre-trained ensemble learning model, the ensemble learning model comprises a plurality of base learners and a secondary model, and the base learners are used to predict according to the nuclide characteristic index data; the secondary model is used for integrating the prediction results of the base learners to obtain the nuclear reaction cross section. The invention aims to optimize the generation efficiency and accuracy of nuclear reaction cross section data by combining various machine learning and deep learning technologies.

The invention belongs to the technical field of nuclear radiation protection, and particularly relates to an advanced fast reactor special alpha nuclide human body protection filter box and a protection device. A tank cover is connected with the upper end of a tank body, a rubber pad is arranged in the tank cover, the tank cover and the tank body are externally connected through a connecting rope, a special alpha nuclide filtering unit is arranged in the tank body, air particle filtering paper is arranged below the special alpha nuclide filtering unit, and a rubber plug is additionally arranged on the inner side of the tank cover; sealant is poured into the seam of the special alpha nuclide filter unit and the air particle filter paper. The filtering efficiency of special alpha nuclide aerosol is larger than 99.99% through the gradient filtering design, the breathing resistance is reduced to 180 Pa or below by adopting high-specific-surface-area impregnated carbon (larger than 1000 m < 2 > / g), and the filtering efficiency and the performance stability of the filtering box are enhanced in the mode that the air particle filtering paper and the impregnated carbon are stacked.

The invention discloses a reactor loop nuclide on-line monitoring system and method, relates to the technical field of reactors, and can solve the technical problem that whether a fuel element cladding is damaged or not cannot be accurately monitored on line at present. The system comprises a sampling module, a control module and a monitoring module, the sampling module comprises a main loop and a target loop which are connected in parallel; the control module is used for controlling a loop coolant to flow in the main loop and the target loop so as to stop the loop coolant in the target loop as a loop coolant to be detected, and the loop coolant in the main loop keeps flowing; and the monitoring module is used for monitoring the activity of target gamma radionuclide in the loop coolant to be detected.

1. Name of the product in this design: Handheld radionuclide identifier. 2. Application of this design: This design is used for the detection of radioactive materials, dose rate measurement, and identification of radionuclides. 3. The key design feature of this product is its shape. 4. The image or photograph that best illustrates the design's key points: a 3D model. 5. The bottom surface of this product is a part that is not easily seen or cannot be seen during use, so the bottom view is omitted.

Nuclear emergency 131 I Isotope measuring device and its remote calibration method, the method at least includes: in the standard environment: to 131 I Isotope at least two distance conditions of detection efficiency measurement, and establish 131 I Isotope and the detection efficiency relationship of the first reference isotope;In the remote calibration site and in the condition of energy calibration: measure the measured detection efficiency of one of the distance conditions of the first reference isotope with known activity, and determine 131 I Isotope measured detection efficiency in at least one distance condition;Compare 131 I Isotope measured detection efficiency with the corresponding detection efficiency is consistent or not.The present application completes the remote calibration of 131 I Isotope measuring device by using the reference isotope with long half-life and the detection efficiency relationship of the reference isotope and 131 I Isotope, without transporting high activity standard source to the site, solves the problem that 131 I Isotope is not easy to save due to short half-life.

The application belongs to the technical field of irradiation effect of nuclear fuel and material, and particularly relates to a fissile nuclide Zr quantitative analysis method based on SIMS. 90 Zr + Ion implantation, coating treatment of the surface of the CeO2 sample after ion implantation, optimization adjustment of SIMS analysis parameters, depth profiling and concentration calibration of fissile nuclide Zr in the CeO2 sample. 90 The method can quickly test and obtain the concentration distribution of fissile nuclide Zr in the depth direction of the CeO2 simulation fuel, the relative sensitivity factor and the detection limit. 90 The method can be used for the characterization analysis of fissile nuclide Zr in the irradiated nuclear fuel, and the obtained relative sensitivity factor of fissile nuclide Zr can be used for the concentration calibration of the secondary ion intensity of fissile nuclide Zr in the irradiated nuclear fuel at different radial positions along with the depth change curve, so as to obtain the concentration distribution of fissile nuclide Zr along the radial direction in the irradiated nuclear fuel. 90 The method can be used for the characterization analysis of fissile nuclide Zr in the irradiated nuclear fuel, and the obtained relative sensitivity factor of fissile nuclide Zr can be used for the concentration calibration of the secondary ion intensity of fissile nuclide Zr in the irradiated nuclear fuel at different radial positions along with the depth change curve, so as to obtain the concentration distribution of fissile nuclide Zr along the radial direction in the irradiated nuclear fuel. 90 The method can be used for the characterization analysis of fissile nuclide Zr in the irradiated nuclear fuel, and the obtained relative sensitivity factor of fissile nuclide Zr can be used for the concentration calibration of the secondary ion intensity of fissile nuclide Zr in the irradiated nuclear fuel at different radial positions along with the depth change curve, so as to obtain the concentration distribution of fissile nuclide Zr along the radial direction in the irradiated nuclear fuel.

The invention provides a method for determining the migration distance of nuclide in a fractured rock mass by considering multistage decay, and relates to the technical field of radioactive waste deep geological disposal safety evaluation. According to the method, a random discrete fracture network rock mass model is constructed based on actual site geological information or statistical laws, a unified pipe network method is adopted to discretize a fracture-matrix system into a flow network formed by nodes and pipelines, and a coupling discrete control equation set of multi-nuclide migration is established under the framework; convection, dispersion, diffusion, adsorption retardation and multistage decay processes are synchronously depicted. Spatial and temporal distribution data of nuclide concentration are obtained through numerical solution, the farthest migration distance of mother nuclide is extracted, fracture connectivity density parameters are introduced to carry out quantitative analysis on fracture structure influence, and then an empirical relational expression used for predicting the maximum migration distance is obtained through fitting. According to the method, the nuclide migration ability can be rapidly evaluated under the complex fracture condition, and a quantitative basis is provided for radioactive waste disposal safety evaluation.