57 results about "Fission" patented technology

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.

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 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 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.

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 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 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.

This invention relates to a method for preparing an ultrathin 4π fission source based on a graphene substrate, belonging to the field of fission source preparation. The method involves preparing a graphene-toughened carbon film, growing graphene of a predetermined thickness on the film, and transferring the resulting substrate material. Fissionable material is then transferred onto the graphene substrate using a combination of solution drying and vacuum self-transfer technology, resulting in an ultrathin fission source capable of emitting fission fragments in the 4π direction. By transferring fission nuclides to a nanometer-thick graphene substrate, the fission products released from the fission source are emitted in the 4π direction, and the average energy loss of the fission fragments penetrating the substrate is far less than their own kinetic energy. This yields a high-quality fission source with low energy loss and minimal self-absorption, achieving efficient and stable preparation of an ultrathin 4π fission source, providing a relatively "perfect" fission source for fission physics experiments or related detector calibration.

This invention discloses a method for predicting hydrocarbon charging based on the superposition of ancient and modern structures, comprising the following steps: Step 1: Outputting a current structural contour map and an uncertainty thermogram that combine spatial accuracy and uncertainty quantification; Step 2: Constructing a structural change index to quantify the destruction / preservation weights of each phase on the trap; Step 3: Integrating apatite fission track and (U-Th) / He thermochronological data to output a paleostructural contour map with error distribution and a paleoriver probability field; Step 4: Generating a three-dimensional volume of sand body probability distribution, coupling a digital core pore network model to simulate the critical threshold for hydrocarbon charging, and generating a sand body connectivity index map to constrain the distribution of high-quality reservoirs; Step 5: Superimposing multiple parameters in the superposition area of ​​ancient and modern structural high points, applying the SHAP algorithm to analyze the contribution of reservoir-controlling elements to focus on the main controlling factors and lock in favorable target areas for hydrocarbon enrichment; Step 6: Generating a three-dimensional probability volume of favorable target areas and combining it with an economic evaluation module to generate an exploration priority ranking.

The present invention belongs to the field of calculation technology of fission product nuclear variation, and specifically relates to an analytical calculation method for fission yield variation in a strong neutron field, comprising the following steps: Step S1, determining initialization parameters, including product nuclear ID, burnup depth R, neutron energy E n In step S2, the spectral average cross section is read from the database; in step S3, the change in product nuclear density due to a nuclear reaction channel i is calculated relative to the change in the independent yield; in step S4, the contributions of each nuclear reaction channel calculated in step S3 are summed to obtain the change in product nuclear density. The analytical calculation method for strong neutron field fission yield variation provided by the present invention can rapidly calculate the change in strong neutron field fission yield and reveal the physical mechanism of yield variation, providing important technical support for strong neutron field research in nuclear testing.

Fissile target materials are provided. The fissile target materials can include a target substrate and a capturing layer operably interfacing with at least one surface of the target substrate. Fission fissile target materials are also provided that can include a target substrate comprising at least one fissile atom and a capturing layer operably interfacing with at least one surface of the target substrate. The capturing layer can include at least one fission product.

The invention relates to a fractured sandstone uranium mine prompt neutron logging numerical simulation method based on a disperse multi-angle random statistical theory. The method comprises the steps of firstly establishing a geological model of the fractured sandstone-containing uranium mine based on a dispersion multi-angle theory; then the simulation pulse neutron source emits neutrons to the stratum, the neutrons and uranium in the cracks are subjected to fission reaction, and instant neutrons are released; then simulating the movement and reaction process of neutrons in the stratum; and finally, recording logging response by a neutron flux counter, and obtaining epithermal neutron count and thermal neutron count according to the energy range. Through tests, the fractured sandstone uranium mine prompt neutron logging numerical simulation method based on the scattered multi-angle random statistical theory can effectively simulate prompt neutron logging responses of sandstone uranium mines under fractures of different angles, and epithermal neutron counting and thermal neutron counting are obtained.

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 nuclear fuel burnup detection, and particularly discloses a micro-zone burnup test method based on 98Mo quantitative analysis aiming at the problem of low accuracy of burnup detection of different micro-zones of nuclear fuel in an existing nuclear fuel burnup measurement method. The micro-zone burnup test method comprises the following steps: S1, taking a nuclear fuel sample, and determining the Mo content of the sample by adopting EPMA; s2, performing isotope test on fission nuclide 98Mo and isotope thereof in the sample by adopting SIMS; s3, the content C < 98Mo > of the 98Mo is calculated; s4, calculating a micro-area fuel consumption value Bu of the nuclear fuel according to the following formula; wherein B is a microcell burnup value, 9.380 is a coefficient, m < 98Mo > is the relative atomic weight of 98Mo, m < UO2 > is the relative atomic weight of UO2, and FY is the yield of nuclide obtained by fission of 235U into 98Mo. According to the method, the damage to the nuclear fuel sample is small, the micro-zone burnup in the micrometer size range of the nuclear fuel can be efficiently detected, the micro-zone burnup value of the nuclear fuel and the micro-zone burnup distribution at different positions of the fuel are obtained, the measurement range is more accurate, and the accuracy is higher.

The invention discloses a fission track detection method and device under a microscope, and relates to the technical field of geochronology, and the method comprises the steps: obtaining a fission track image to be detected under the microscope; performing fission track positioning on a fission track image to be detected through the trained Mask R-CNN model, and outputting masks and boundary coordinates to which a plurality of fission tracks belong; performing ellipse fitting on the contour of the associated fission track by adopting each boundary coordinate, and determining an ellipse long axis and an ellipse short axis of each fission track; correspondingly calculating a track overlapping coefficient based on the elliptical long axis and the elliptical short axis of each fission track; and color adjustment is performed on the associated masks according to a comparison result of each track overlapping coefficient and an overlapping threshold value, each mask is mapped to the fission track image to be detected, and an identification fission track image is output. Based on the above scheme, through combination of Mask R-CNN model identification and ellipse fitting overlapped fission track determination, efficient and accurate identification of fission tracks is realized on the whole.