21 results about "Fission" patented technology

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 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.

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.

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.

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 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.

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.

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.

This invention relates to a method and system for diagnosing faults in microfission ionization chambers, comprising: monitoring microfission ionization chambers in the same pressurized water reactor nuclear power unit; and when any microfission ionization chamber experiences an anomaly in calibration coefficient or plateau tilt, obtaining the fault type of the microfission ionization chamber to be diagnosed; if the fault type is an excessive calibration coefficient, then a preset calibration coefficient diagnostic method is used to diagnose the microfission ionization chamber to be diagnosed; if the fault type is an abnormally large or excessive plateau tilt, then a preset plateau tilt diagnostic method is used to diagnose the microfission ionization chamber to be diagnosed. This invention uses a preset calibration coefficient diagnostic method based on the physical meaning of the calibration coefficient, providing more comprehensive diagnostic evidence. Simultaneously, the use of a preset plateau tilt diagnostic method solves the problem of insufficient evidence when judging solely based on an abnormally large or excessive plateau tilt. Moreover, the diagnostic method of this invention has clear logic and sufficient evidence, enabling effective diagnosis.

The invention relates to a method and a system for calculating fission stock and impurity activity release amount of a reactor core. The method comprises the following steps: constructing a model of the reactor core and a fuel element; controlling the fuel element model to circularly flow in the reactor core model, and dividing the whole process time of the circular flow into a plurality of time steps; the fission stock and the impurity activity release amount of each fuel element model in each time step are calculated, and calculation results are obtained; when the refueling batch is one time and an effective multiplication factor is calculated through a reactor core physical calculation program and meets a set threshold value, a calculation result is stored in a first array; when the number of the refueling batches is at least two and the fuel element model burnup and the unloading burnup meet the preset conditions, the calculation result of the circular flow is stored in a second array; and obtaining calculation results of the fission stock and the impurity activity release amount of the reactor core through the first array or the second array. The problems that the accumulation amount and the impurity activation rate of fission products cannot be accurately calculated, and the historical track of the fuel element is difficult to track are solved.

The application discloses a poster fission generation method and device and a medium. The method comprises the following steps: constructing a poster fission template of a fission poster, wherein the poster fission template comprises a picture, text and a preset element region; performing key feature extraction on the poster fission template to obtain key region information and reserved position information of the poster fission template; adjusting a to-be-fissioned poster content according to size information in the reserved position information to obtain adjusted to-be-fissioned poster content; and embedding the adjusted to-be-fissioned poster content into the poster fission template based on the reserved position information and the key region information to generate a fission poster.

The invention relates to a Mo-99 target optimized production method based on fission-activation co-production, which comprises the following steps: establishing a high-fidelity full-core model, and constructing a model which takes an LEU target as a matrix and is doped with highly enriched Mo-98 powder in different proportions; the preparation method comprises the following steps: preparing a dispersion-type target of a uniform mixture, adjusting a proportional internal fission-external activation-type layered target and an internal activation-external fission-type layered target by changing the radius of a core part, performing RMC simulation to obtain the Mo-99 yield and specific activity of the dispersion-type target, and further obtaining a yield optimization strategy.