110 results about "Fission" patented technology

The invention relates to a business data processing method and device, computer equipment, a storage medium and a computer program product. The method comprises the steps of obtaining a fission sharing path and initial user information of service data, wherein the fission sharing path comprises multiple pieces of fission user information and a data sharing mode between users; constructing a graph network model of the service data according to the information of the multiple fission users and the data sharing mode; screening fission user nodes of the graph network model according to the initial user information to obtain an initial node; constructing a directed propagation path of service data according to the out-degree of the initial node and a node connection relationship between fission user nodes; and if the number of the nodes in the directed propagation path is not less than a preset fission threshold, generating fission sharing information of the service data according to the directed propagation path. By adopting the method, the data structure can be simplified, and the data volume can be compressed, so that the fission sharing information generation efficiency is improved.

The invention provides a method for quantitatively evaluating the storage condition of magma hydrothermal deposit based on thermochronology, which comprises the following steps: determining a sample sampling point based on field geological survey, collecting typical samples, picking out zircon and apatite monominerals, and preparing a U-Pb isotope test target and a fission track test optical slide; carrying out fission track dating on the zircon and the apatite to measure the U-Pb isotope age and (U-Th) / He isotope age of the zircon and the apatite; thermal history evolution process simulation is carried out by integrating test data, and finally, the deposit preservation condition is quantitatively calculated and evaluated according to simulation. According to the method, a scientific quantitative full-chain research scheme is provided for augmentation, denudation and thermal history evolution research of the magma hydrothermal deposit, and the technical problem of quantitative evaluation of the storage condition of the magma hydrothermal deposit is solved.

The invention discloses a uranium-radium balance correction method, device, medium and equipment, and relates to the field of sandstone type uranium mine uranium exploration, and the method comprises the steps: obtaining an industrial uranium mine hole, carrying out the geophysical logging of the industrial uranium mine hole, determining a mineralized section through quantitative gamma interpretation, and carrying out the geological logging and geophysical exploration logging of a rock core, uranium and radium samples are collected and analyzed in the mineralization section; uranium fission instant neutron logging is carried out on the mineralization section, and the uranium ore section depth range is determined according to the ratio of the epithermal neutron time spectrum to the thermal neutron time spectrum of uranium fission instant neutron logging; according to the uranium radium sample analysis result of the mineralization section, determining a single sample section uranium radium balance coefficient, and according to the single sample section uranium radium balance coefficient, determining a single engineering uranium radium balance coefficient; determining a uranium radium balance coefficient partition map according to the single-project uranium radium balance coefficient; and correcting the quantitative gamma logging curve in the uranium ore section depth range of uranium instantaneous neutron logging by utilizing the coefficients in different uranium-radium balance regions to obtain the actual depth range, thickness and grade of the uranium ore section.

This invention relates to the monitoring and diagnosing of nuclear power plants for its thermal performance using the NCV Method. Its applicability comprises any nuclear reactor such as used for research producing a useful output. Its greatest applicability lies with conventional Pressurized Water Reactor and Boiling Water Reactor nuclear plants generating an electric power. Its teachings of treating fission as an inertial process, a phenomena which is self-contained following incident neutron capture, allows the determination of an absolute neutron flux. This process is best treated by Second Law principles producing a total fission exergy. This invention also applies to the design of fusion thermal systems regards the determination of its Second Law viability and absolute plasma flux.

The application provides a heat pipe micro reactor steady-state multi-physical coupling method and system based on a fission response function, which integrates a fission response function physical program, a self-developed Unifoam finite element thermal program and a heat pipe program, realizes steady-state joint solution of three physical fields of nuclear neutronics, solid thermal engineering and heat pipe two-phase flow through a coupling iterative driving script, and obtains accurate power field and temperature field distribution of the heat pipe micro reactor under a steady-state working condition. The method pre-constructs a high-fidelity fission response function database through a Monte Carlo method, does not need to repeatedly perform a complex particle transport process during operation, and only needs to complete neutronics solution through matrix multiplication, so that the simulation accuracy and the calculation efficiency are effectively balanced.

The invention provides a reactor core fuel assembly source distribution calculation method, system and equipment and a medium, and relates to the technical field of nuclear reactors. The reactor core fuel assembly source distribution calculation method provided by the invention comprises the following steps: acquiring input parameters, wherein the input parameters comprise power distribution, fission nuclide share, fission spectrum, average release particle number of each fission, and average release energy and source bias factor of each fission; calculating a cumulative distribution function of source particles based on the input parameters; and sampling the source particles based on the cumulative distribution function, and determining distribution parameters of the source particles. The source particles are sampled based on the cumulative distribution function of the source particles, the distribution parameters of the source particles are determined, the shielding calculation efficiency based on the MC method is improved, and support is provided for new heap design.

The invention relates to a modeling method and system for a non-uniform reactor core three-dimensional neutron physical model, a storage medium and electronic equipment, and the method comprises the steps: carrying out the material component analysis of a fuel assembly of a reactor, and obtaining a plurality of different segments; performing homogenization treatment on each of the plurality of different segments to obtain a plurality of different uniform segments; performing calculation based on the components in the uniform segment and the thermal parameters to obtain section parameters of the uniform segment; modeling based on the section parameters and in combination with the geometric parameters of the reactor core to obtain a multi-group neutron diffusion equation; solving the multi-group neutron diffusion equation to obtain the neutron flux of uniform segments; and calculating according to the neutron flux in combination with the fission cross section to obtain the three-dimensional power distribution of the reactor core. The method can be suitable for modeling of the reactor core composed of non-uniform components, and the model precision is high.

The present invention provides a method and device for calculating a transient splicing fission matrix of a core model, the method comprising: obtaining a pre-established fission matrix database; wherein the fission matrix database comprises fission matrices corresponding to a uniform model of the core model under various material properties; selecting corresponding target fission matrix elements from the fission matrix database based on the current state of the target material properties of the core model, splicing a transient splicing fission matrix corresponding to the current state of the target material properties based on the target fission matrix elements; and correcting the transient splicing fission matrix based on the distribution boundary information of the target material properties in the current state to obtain a corrected transient splicing fission matrix. The present invention can simulate the transient process of the enrichment, control rods and temperature of the core model under a changing state, so that the calculated transient splicing fission matrix is ​​more in line with the actual situation, and improves the calculation accuracy and calculation rate of the transient splicing fission matrix.

The present application discloses a method, device and related equipment for determining the apparent age of zircon fission tracks, the method comprising: determining an annealing model of zircon fission tracks, the annealing model recording the constraint relationship between track length ratio, time and temperature; determining the track length ratio of the zircon sample in each time interval within a target simulation period based on the annealing model and a preset thermal history curve, the target simulation period being the period included in the thermal history curve; determining the track density of the zircon sample in each time interval within the target simulation period based on the track length ratio of the zircon sample in each time interval within the target simulation period; determining the apparent age of the fission track of the zircon sample based on the track density of the zircon sample in each time interval within the target simulation period. The present application calculates the apparent age of the fission track of the zircon sample by forward simulation of the fission track, which has certain practical significance in the application of low-temperature chronology.

The present application proposes a method and device for detecting the safe operation of a pebble bed high-temperature gas-cooled reactor, wherein the method comprises: performing core diffusion calculation on the pebble bed high-temperature gas-cooled reactor based on the nodal method to obtain the multi-group average neutron flux density of each nodal block of the pebble bed core; obtaining the multi-group neutron flux density and multi-group defect factor of each batch of fuel balls, as well as the multi-group neutron flux density and multi-group defect factor of the TRISO fuel particles in each batch of fuel balls; determining the power distribution of the TRISO fuel particles of the pebble bed high-temperature gas-cooled reactor based on the multi-group average neutron flux density of each nodal block of the pebble bed core, the fission energy generation cross section of the TRISO fuel particles, the multi-group defect factor of each batch of fuel balls and the TRISO fuel particles in each batch of fuel balls, and then judging whether the pebble bed high-temperature gas-cooled reactor is operating safely, thereby providing a more reliable data basis for the nuclear design and safety analysis of the pebble bed high-temperature gas-cooled reactor.

Some embodiments of the invention disclose a modeling method for a fission gas swelling behavior of UO2 fuel under a high-burn-up condition, and the rule of irradiation swelling caused by irradiation bubbles in a high-burn-up structure in the UO2 fuel is accurately determined by coupling grain refinement and bubble evolution. The method comprises the steps of constructing a total free energy density function; based on a recrystallization grain nucleation theory, establishing a grain refinement nucleation model; based on an Allen-Cahn equation, a phase field dynamics evolution equation of high-burn-up structure evolution is constructed; solving a total free energy density function and a phase field dynamics evolution equation, and carrying out visualization processing on a numerical solution; and based on the final morphology of the high-burnup structure obtained after visualization processing, determining the effective thermal conductivity, porosity and grain size in the high-burnup structure in the UO2 fuel, and determining the rule of irradiation swelling caused by irradiation bubbles in the high-burnup structure in the UO2 fuel.

The invention relates to the technical field of nuclear reactors, in particular to a fission fragment pumping test method based on a three-dimensional model and a related product, and the method comprises the following steps: establishing a three-dimensional analysis model of a fission fragment pumping unit; based on the three-dimensional analysis model, a Monte Carlo method is adopted to calculate and obtain calculated neutron flux at a preset position and calculated fission fragment deposition energy in a gain medium in the fission fragment pumping unit; determining a conversion factor according to the ratio of the calculated fission fragment deposition energy to the calculated neutron flux; at the preset position, the actual measurement neutron flux in the test is obtained through measurement; determining fission fragment deposition energy of the fission fragment pumping unit according to the conversion factor and the measured neutron flux; by establishing the three-dimensional analysis model and calculating the conversion factor of the external neutron flux and the internal deposition energy in advance, the problem of direct measurement of the interior of the gain medium in an extreme test environment is solved.

The application discloses a data object risk prediction method, comprising: sending a unique identifier of a to-be-predicted data object to a second server, the unique identifier being used for instructing the second server to query a plurality of related data corresponding to the to-be-predicted data object according to the unique identifier, a counting statistical result of each fission label corresponding to the to-be-predicted data object is determined according to the multiple pieces of target annotation data and the multiple pieces of related data, the multiple pieces of target annotation data are obtained after the first server conducts data fission processing on the multiple pieces of initial annotation data, and the initial annotation data comprise initial labels; the target labeling data comprises fission labels, and one initial label corresponds to one or more fission labels; receiving a counting statistical result corresponding to each fission label sent by the second server; determining a counting statistical result corresponding to each initial label according to a corresponding relationship between the initial labels and the fission labels; and performing risk prediction on the to-be-predicted data object based on the counting statistical result corresponding to each initial label.

The invention discloses a fission yield experiment data screening method, and relates to the technical field of nuclear data evaluation.The method comprises the steps that S1, fission yield experiment data are obtained from an EXOR database, and the experiment data are grouped; s2, fitting each group of fission yield experimental data by adopting a Zp model, and determining model parameters; the fission yield is calculated according to the determined model parameters, so that a yield calculation value is obtained; s3, calculating the deviation between the fission yield experiment data and the fission yield calculation value obtained in the step S2; and S4, screening the experimental data according to the deviation obtained in the step S3. According to the method provided by the invention, the evaluation efficiency can be effectively improved, the research period is shortened, and technical support is provided for further establishing an experimental database.

Fission reactions within a molten salt reactor may produce gaseous fission products. These fission products may be removed through a variety of methods. However, proper removal of such fission products requires a thorough understanding of the proportion of gaseous space to fluid space within the component being interrogated, known as the void fraction. The present invention provides a means for determining the void fraction within such a component utilizing an enrich cadmium zinc telluride detector to produce collimated radiation data. The collimated radiation data may then be used to compute a mean void fraction across the domain being interrogated.

The invention relates to the technical field of geological exploration, and discloses a deep exploration evaluation method based on multi-parameter geology and thermal history simulation. Mineralized quartz and granite samples in a target area are collected, the samples are sent into a laboratory to be processed, including cleaning, cutting and grinding, the collected samples are analyzed, and the analysis result is obtained. The method specifically comprises the following steps: analyzing the metallogenic depth of an inclusion, analyzing the invasion depth of a rock mass, analyzing the invasion age of a sample, and obtaining the cooling age of the sample by combining fission tracks of zircon and apatite. The method is used for analyzing a collected sample, and specifically comprises the following steps: analyzing the metallogenic depth of an inclusion, analyzing the invasion depth of a rock mass, analyzing the invasion age of the sample, obtaining the cooling age of the sample by combining fission tracks of zircon and apatite, and carrying out comprehensive calculation based on the denudation thickness, the invasion depth and the formation depth of an ore body. Therefore, the storage condition of the ore body is finally obtained, deep mineralization favorable information is obtained, and deep mineral product prediction is realized.

The utility model discloses a lead-copper-free PPR (pentatricopeptide repeats) anti-freezing fission reducing adapter substitute, and belongs to the technical field of pipeline adapters. Comprising a conversion head body, a first threaded hole is formed in the top face of the conversion head body, a second threaded hole is formed in the bottom face of the conversion head body, a conical cavity is formed in the conversion head body and communicates with the first threaded hole and the second threaded hole, and a connecting block is installed in the middle of the outer side of the conversion head body; connecting rings are installed at the positions, close to the two ends, of the outer side of the adapter body, corrugated pipes are installed on the sides, close to the connecting blocks, of the connecting rings, the corrugated pipes are connected with the connecting blocks in an adsorption mode, heat preservation cotton adheres to the inner walls of the corrugated pipes, and a vacuum cavity is formed in the adapter body; according to the technical scheme, heat loss can be reduced through the heat preservation cotton, the heat preservation effect on fluid in the pipeline is achieved, and the fluid freezing risk is reduced; the vacuum cavity utilizes the vacuum heat insulation principle to further enhance the heat preservation effect and prevent the internal fluid from being frozen due to the too low surface temperature of the adapter body.

The present disclosure relates to a parameter determination method and apparatus for a uranium fission prompt neutron logging model, and a storage medium, which include: determining starting time, terminating time and a time window of a uranium fission prompt neutron of each neutron type by constructing the uranium fission prompt neutron logging model, setting quantitative parameters and exploration areas of the model, simulating and calculating flux time spectra of a plurality of neutron types under different variable parameters and comparing the time spectrum of the uranium fission prompt neutron with a background time spectrum; determining an accumulated flux of each neutron according to the starting time, terminating time and time window; setting a reference pore diameter, a reference ore-bearing model diameter and a reference ore-bearing model thickness; and determining the model pore diameter, ore-bearing model diameter and ore-bearing model thickness in respective.

The present invention discloses a calculation method, system, medium and device for the generation, diffusion and release of fission gases in UO2 fuel pellets. The method models the diffusion process of fission gases generated in the fuel pellets based on the Booth diffusion equation, adopts a two-dimensional geometric model, and calculates the concentration distribution of fission gases in the pellet grains at each moment through the finite element analysis method. Based on the concentration distribution, the volume release rate of fission gases released from the grains to the grain boundaries is calculated through the concentration gradient on the surface of the pellet grains, and finally the volume release amount of fission gases released from the grains to the grain boundaries and the volume release amount of fission gases accumulated on the grain boundaries are calculated. Based on the traditional Booth diffusion equation of UO2 fuel, the present invention develops a calculation model for the generation, diffusion and release of fission gases in UO2 fuel pellets, and at the same time considers the influence of the geometric shape of the fuel pellets and the fuel temperature transition. Therefore, the concentration distribution of fission gases in the fuel pellets obtained by using the present invention will be more accurate.

The embodiment of the specification provides a neutron detector, comprising: a housing, the interior of which is vacuum; the at least one detection electrode group is arranged in the shell, each detection electrode group comprises a coating electrode, and the surface of the coating electrode is covered with a fission material layer; and the collecting electrode is used for collecting electrons escaping from the fission material layer after the neutrons are injected into the fission material layer.

The present disclosure provides a neutron transport simulation method, related device and computer program product, relating to the technical field of computer technology and nuclear physics. The method is implemented in the following manner: CSG modeling is performed on a reactor core to construct a CSG tree; N projection planes are generated around the reactor core, and an incident point array is generated on all the projection planes; a blank Segment list is constructed, and recursive characteristic line tracking is performed on all the incident points starting from the root node of the CSG tree, and Segment data is sequentially filled into the blank Segment list; GPU layer and GPU internal thread layer load balancing is performed; the characteristic line, FSR and material data are transmitted to each GPU; each GPU performs transport scanning calculation; the global FSR angular flux is calculated; each GPU calculates the fission moment and the fission source; the K effective value is calculated and whether it converges is verified. The technical solution of the present disclosure can break the CSG geometry construction restriction in the existing neutron transport simulation method, reduce unnecessary characteristic line generation to save computing resources, and improve the iteration solving speed to speed up the simulation.

The invention belongs to the technical field of nuclear fuel simulation design, and provides a U3Si2-Cu dispersion type nuclear fuel with a stable interface at a high temperature and a simulation construction method aiming at the problem that an interface reaction is easily generated between U3Si2 fuel particles and an Al matrix of the existing U3Si2-Al fuel at the high temperature. A U-Si-Xe-M quaternary potential function is developed by combining a gradient optimization algorithm of multi-target simulated annealing with an MEAM model, M is any element of Al, Cu, Mg and W, atomic scale information of irradiation defects and fission gas in U3Si2-M fuel at an interface is obtained by means of computational simulation, and finally U3Si2-Cu dispersion type nuclear fuel is obtained. The U3Si2-Cu dispersion type nuclear fuel provided by the invention has relatively high interface stability at a high temperature.

The embodiment of the invention provides an automatic fission molybdenum-99 separation method and device which are applied to a fission molybdenum-99 separation device, and the fission molybdenum-99 separation device comprises a separation system, a liquid supply system and a control system. The separation system comprises a first separation device, a second separation device and a third separation device which are sequentially connected. The liquid supply system is connected with the separation system, and the control system is used for controlling the liquid supply system to convey the stored media to the separation system according to a set sequence. The automatic separation method of the fission molybdenum-99 comprises the steps that the first separation device is filled with aluminum oxide, the second separation device is filled with anion exchange resin, and the third separation device is filled with activated carbon. The first separation device is used for carrying out primary separation on the target dissolving liquid to obtain primary purified liquid; the separation system comprising the first separation device, the second separation device and the third separation device is arranged, the liquid supply system and the control system are arranged, and the control system controls medium injection of the liquid supply system, so that the whole separation process can be completely automatic.

Additive manufacturing compositions include low-absorbing particles or non-absorbing particles that have an absorbance for wavelengths of 300 nm to 700 nm that is equal to or greater than 0 Au and is less 1.0 Au, such as 0.001 Au absorbance≤0.7 Au. Slurries including such particles and an uranium-containing particle and that are used in additive manufacturing processes have an increased penetration depth for curative radiation. Removal of low-absorbing particles or non-absorbing particles during post-processing of as-manufactured products results in pores that create porosity in the as-manufactured product that provide a volume accommodating fission gases and / or can enhance wicking of certain heat pipe coolant liquids. Low-absorbing particles or non-absorbing particles can be functionalized for improved properties, for example, with fissionable material for improved ceramic yields, with burnable poisons or stabilizers for increased homogeneity, with stabilizers for localized delivery of the stabilizer, or with combinations thereof.

The invention relates to a fission matrix-based pressurized water reactor PIN-level power online reconstruction method, which comprises the following steps of: sequentially constructing a pressurized water reactor power database, an in-reactor detector response matrix database and a high-order harmonic database according to a pressurized water reactor core model; the method comprises the following steps: performing dimension reduction processing on a high-order harmonic database, obtaining in-pile detector counting as an actual measurement value through critical-fuel coupling, and reconstructing the pin-by-pin power distribution of a reactor core by combining the actual measurement value according to the high-order harmonic database and an in-pile detector response matrix database in an online stage by utilizing a least square method. According to the method, the detector response matrix constructed in advance and the high-order feature vector under the typical condition are utilized, and real-time monitoring can be achieved only by calculating the expansion coefficient in the online stage. According to the method, the actual measurement information of the detector in the reactor is fully utilized, the PIN-level power distribution in the reactor is reconstructed, and the reactor power can be monitored in real time by utilizing the detector response matrix constructed in advance and the high-order feature vector under the typical condition.

The invention discloses a Gd-Nb co-doped perovskite and a preparation method and application thereof, and belongs to the technical field of solidification of highly radioactive nuclear waste. The preparation method of the Gd-Nb co-doped perovskite in the present invention comprises the following steps: grinding and mixing a calcium source, a zirconium source, a titanium source, a gadolinium source, a niobium source, and an oxygen source, performing SPS sintering at 1100-1250°C, and performing annealing and decarbonization after sintering to obtain the Gd-Nb co-doped perovskite. The present invention synergistically processes the actinide nuclide substitute Gd and the fission element Nb, has a short synthesis cycle, and the prepared Gd-Nb co-doped perovskite has a high density.

The present application relates to a kind of trivalent actinide and rare earth fission fragment nuclide separation device and separation method, belong to actinide and fission product separation technical field, solve the technical problem that separation exists decontamination low, recovery rate is low, its separation device includes liquid storage tank unit, chromatography separation unit, waste liquid collector, product collection unit, delivery pump.Liquid of liquid storage tank unit enters chromatography separation unit by delivery pump, the liquid that chromatography separation unit flows enters waste liquid collector or product collection unit.Liquid storage tank unit, chromatography separation unit, product collection unit, waste liquid collector are connected by pipeline, on pipeline, set valve.Its separation method includes preparation before separation, the steps of adsorption, washing, leaching of trivalent actinide and rare earth fission fragment nuclide.The present application is used for the separation of trivalent actinide and rare earth fission fragment nuclide.

The invention discloses a continuous energy characteristic line neutron and photon transport calculation method applied to a nuclear-related device. The method comprises the following steps: 1, carrying out continuous energy deterministic theory neutron transport calculation, photon transport calculation and neutron and photon transport calculation on a nuclear-related facility; 2, generating characteristic line information on the basis of the geometric model of the nuclear facility, and performing source iteration flux initialization; 3, calculating the fission source of each energy section of the nuclear-related facility; 4, calculating a scattering source of each energy section of the nuclear-related facility; 5, calculating the average angular flux of the nuclear facility and the flux density of the neutron standard; 6, judging whether the flux of the ith energy section of the nuclear-related facility is converged or not, and if not, returning to the step 4; judging whether all energy segments are scanned or not, and if not, returning to the step 4; judging whether the flux of each energy section converges or not, and if not, returning to the step 4; and 7, calculating an effective value-added factor keff of the nuclear-related facility, judging whether the keff is converged or not, and if the keff is not converged, returning to the step 4. According to the method, neutron and photon transport calculation with a wide application range and high efficiency is realized.

The invention relates to the technical field of oil exploration and development, in particular to a paleostructure denudation amount recovery method, medium and equipment, and the method comprises the following steps: obtaining an apatite sample of a researched area; testing the apatite sample by adopting an external detector method, and calculating the fission track age; and according to the age and length data of a single particle of the fission track, the structure-thermal history evolution process of the sample is simulated by combining the actual geological characteristics of the sampling well. On the basis of drilling and logging data, analysis and test data and seismic data, an apatite fission track method is used for calculating the distance between the corresponding depth and the respective unconformity surface of the ancient ground temperature gradient curve extending to the ancient ground surface temperature position in each period, and then the lifted denudation amount in each period is obtained.

The application discloses a method for selectively separating molybdenum from a nitric acid solution containing uranium and fission element. The method comprises the following steps: chemically grafting 8-hydroxyquinoline to silica gel by using a silane coupling agent to prepare a hydroxyquinoline grafted silica gel material, and then using the material as an adsorbent to efficiently and selectively adsorb molybdenum from a nitric acid solution containing uranium and fission element. The method breaks the limitation that known silica gel grafts and single quinoline ligands cannot well realize the selective separation of molybdenum and other elements, adjusts the spatial configuration of the 8-hydroxyquinoline ligand by grafting 8-hydroxyquinoline on the silica gel, provides the selective adsorption effect of the 8-hydroxyquinoline ligand on molybdenum and uranium and the like, and improves the saturated adsorption amount of molybdenum.