35 results about "Fission" patented technology

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.

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 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 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 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 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.

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.

The invention discloses a multi-treatment-room boron neutron capture therapy system based on an external source driven subcritical reactor core, and particularly relates to the technical field of medical science, the multi-treatment-room boron neutron capture therapy system comprises an external neutron source, the external neutron source comprises a charged particle transport section and a charged particle targeting section, the charged particle transport section is used for accelerating and transporting charged particles, and the charged particle targeting section is used for targeting the charged particles; and the charged particle targeting section is connected with the charged particle transport section, so that the charged particles react with the target material to release primary neutrons. According to the boron neutron capture therapy system, multiplication and fission of an initial neutron source can be achieved, the neutron flux level needed by a therapy beam is greatly improved, meanwhile, the system is in a subcritical state, rapid closing of the system can be achieved by cutting off charged particle beams, the possibility of instant supercritical safety accidents is completely eradicated, and the service life of the system is prolonged. And a synchronous multi-terminal boron neutron capture treatment function can be realized, the number of cancer patients treated every year is remarkably increased, and meanwhile, the treatment cost is remarkably reduced.

This utility model relates to the field of mountain blasting technology and discloses a controlled mass fission release device for high-pressure gas in mountain blasting. It includes a cylinder with a cavity, an excitation device, and a nozzle. The cylinder has a pressure release mechanism located within the cavity to divide the cavity into a liquid storage chamber and a gas jet chamber with a nozzle. The excitation device is located in the liquid storage chamber and is used to heat the liquid gas. The cylinder is connected to a mounting ring located at the nozzle and has an axial insertion through-hole, an axially extending annular limiting groove surrounding the axial insertion through-hole, and an installation / removal guide groove extending axially along the mounting ring to connect with the annular limiting groove. The nozzle is inserted into the axial insertion through-hole, with the nozzle inlet communicating with the nozzle. The nozzle has an axial limiting block that slides into the annular limiting groove along the installation / removal guide groove and is spaced apart from the installation / removal guide groove circumferentially along the axial insertion through-hole. The axial limiting block is connected to the mounting ring by a locking element. The convenient installation and removal of the nozzle within the cylinder facilitates maintenance, thereby reducing maintenance costs.

The application discloses a uranium-radium balance correction method, device, medium and equipment, relates to the uranium detection field of sandstone type uranium mines, and comprises the following steps: obtaining an industrial uranium mine hole, performing geophysical well logging on the industrial uranium mine hole, determining a mineralization section through quantitative gamma interpretation, performing geological logging and geophysical logging on a core, collecting and analyzing uranium-radium samples in the mineralization section, performing uranium fission prompt neutron logging on the mineralization section, determining a uranium mine section depth range according to the ratio of epithermal neutrons to thermal neutrons in the time spectrum of the uranium prompt neutron logging, determining a single sample section uranium-radium balance coefficient according to the uranium-radium sample analysis result of the mineralization section, determining a single project uranium-radium balance coefficient according to the single sample section uranium-radium balance coefficient, determining a uranium-radium balance coefficient zoning map according to the single project uranium-radium balance coefficient, correcting the quantitative gamma logging curve in the uranium mine section depth range of the uranium prompt neutron logging by using the coefficients in different uranium-radium balance zones, and obtaining the actual uranium mine section depth range, thickness and grade.

A kind of pressurized water reactor PIN level power online reconstruction method based on fission matrix, according to pressurized water reactor core model, pressurized water reactor power database, in-core detector response matrix database and high-order harmonic database are sequentially constructed, then high-order harmonic database is reduced dimension and in-core detector count is obtained by critical-burnup coupling as actual measurement value, in online stage, according to high-order harmonic database and in-core detector response matrix database, combined with actual measurement value, least square method is used to reconstruct the pin-by-pin power distribution of core.The present application uses the detector response matrix and high-order eigenvector under typical conditions constructed in advance, and only needs to calculate expansion coefficient in online stage to realize real-time monitoring.The present application makes full use of the actual measurement information of in-core detector, reconstructs the pin level power distribution in reactor, uses the detector response matrix and high-order eigenvector under typical conditions constructed in advance, and can realize real-time monitoring of reactor power.

The invention belongs to the related technical field of particle transport and nuclear reactors, and discloses a self-adaptive Vilant shift acceleration method and system for solving neutron transport characteristic values. The method comprises the following steps: S1, constructing a neutron transport equation, initializing a characteristic value and a neutron flux density, and setting an external iteration number n; s2, calculating a Vilant offset and a modified fission source; s3, setting the number of internal iterations l = 0, correcting a fission source item and a Kralov subspace base vector by using the current characteristic value, and calculating a matrix vector product; s3, updating the Kralov subspace base vector and the neutron-flux density, and judging whether the neutron-flux density converges or not: if not, l = l + 1, and returning to S3; and if convergence occurs, updating the characteristic value, n = n + 1, and returning to the step S2 until an external iteration convergence criterion is met. According to the method, the problems of low iteration convergence speed and long calculation time in advanced reactor design and optimization can be solved.

An anti-proliferation technique is disclosed to reduce the likelihood of nuclear proliferation due to the use fissionable fuel salts. The technique includes doping the fuel salt with one or more elements (referred to herein as activation dopants) that, upon exposure to neutrons such as would occur in the fuel salt when a reactor is in operation, undergo a nuclear reaction to, directly or indirectly, form highly active “protecting isotopes” (of the same element as the activation dopant or a different element). A sufficient mass of activation dopants is used so that the Figure of Merit (FOM) of the fuel salt is decreased to below 1.0 within some target number of days of fission. This allows the FOM of the fuel salt to be controlled so that the fuel becomes too dangerous to handle before to the creation of a significant amount of weaponizable isotopes.

The invention discloses a curing agent performance data set construction method capable of reducing the number of experiments, a medium and a system, and the method comprises the following steps: S1, setting initial preparation parameters, preparing a solidified soil sample, and testing to obtain performance data of the solidified soil sample; s2, fission generation is carried out based on the preparation parameters of the previous round, and at least two groups of new preparation parameters are generated by increasing and decreasing the numerical range of the key variables; s3, repeatedly preparing and testing new parameters to obtain new performance data; s4, drawing a performance oscillogram of multiple groups of experimental data, comparing performance changes and identifying local peak points; s5, dynamically adjusting a fission coefficient according to a performance waveform change trend to gradually reduce a fission step size so as to improve the peak value interval search precision; s6, the steps S2-S5 are repeated, and fission is stopped until the difference value between the performance data obtained through fission and the target performance is smaller than a set threshold value; and S7, integrating the data set, and solving the problems of many experiment times, long period and high cost when the initial data set is obtained in the prior art.

The invention provides a safety control module and method for multi-source data acquisition of an electric energy metering box. The safety control module comprises an energy flow anomaly topological system, a dielectric fission entropy increase sensing system and a thermal runaway phase change blocking system. The method comprises the following steps: fusing a microscopic deformation vibration spectrum, internal gas ion mobility and electromagnetic field eddy current distribution into a three-dimensional energy flow density topological graph through unsteady-state field reconstruction; generating an arc fault embryonic period early warning signal according to the fractal dimension mutation rate of the three-dimensional energy flow density topological graph; capturing a relaxation abnormal point in the dielectric constant frequency domain response curve, calculating a material molecular chain fracture entropy increment through relaxation peak offset of the relaxation abnormal point, and generating a medium fission level; and positioning a fission grade ultralimit region in the three-dimensional energy flow density topological graph, calculating a thermodynamic phase change critical path by combining the energy flow gradient, injecting a reverse vortex field along the thermodynamic phase change critical path to offset energy accumulation, and releasing the nanoscale aerogel to block the oxidation chain reaction.

A hybrid nuclear reactor that is operable to produce a medical isotope includes an ion source operable to produce an ion beam from a gas, a target chamber including a target that interacts with the ion beam to produce neutrons, and an activation cell positioned proximate the target chamber and including a parent material that interacts with the neutrons to produce the medical isotope via a fission reaction. An attenuator is positioned proximate the activation cell and selected to maintain the fission reaction at a subcritical level, a reflector is positioned proximate the target chamber and selected to reflect neutrons toward the activation cell, and a moderator substantially surrounds the activation cell, the attenuator, and the reflector.

This invention discloses a fusion plasma combustion-assisted method for fusion-fission coupling, belonging to the field of nuclear fusion technology. The method includes: evaluating the parameters of ultrafine fission particles to determine the type of fission material, the total injection volume, particle size, and injection velocity; preparing ultrafine fission particles of the appropriate size based on the evaluation results; and injecting the prepared ultrafine fission particles into the fusion plasma at the determined injection velocity, causing the fission material to undergo a fission reaction in the fusion neutron environment, and using the kinetic energy of the resulting fission fragments to heat the fusion plasma. This invention utilizes the energy release at the megaelectron volt level of the fission reaction for heating, which is an improvement over existing kiloelectron volt level technologies, effectively reducing the fusion ignition conditions and accelerating the fusion research process.

The invention discloses a rapid calculation method for fission delayed nuclear radiation fluence or dose under the impact wave influence, which solves the problem of low particle fluence or dose calculation speed in a fission delayed nuclear radiation environment under the impact wave influence in the prior art, and considers the complex distribution of air density under the impact wave disturbance. The method can be used for calculating the fission delayed nuclear radiation environment under the impact wave influence, and can provide support for nuclear radiation environment evaluation and personnel and equipment nuclear radiation shielding design in a specific scene. In addition, calculation is carried out based on the mapping relation between the air quality thickness and the nuclear radiation environment, and compared with simulation of a traditional Monte Carlo method, the calculation time is obviously shortened, and the calculation speed is effectively increased.

A method for determining real-time thermal power of a nuclear reactor based on number of gamma rays counted comprising measuring a number of gamma-ray counts by a gamma-ray detector that is placed outside a biological shield at a primary cooling circuit of a fission nuclear power plant so that maintenance of the detector is possible during normal operation of the fission nuclear power plant, and determining, by a computing device in real-time, a thermal power of the fission nuclear power plant based on the number of gamma-ray counts measured by the gamma-ray detector.