808 results about "Neutron" patented technology

The invention provides a near-earth orbit neutron and gamma ray detector group, detection equipment and a detection method, and the detection equipment is a neutron / gamma spectrometer which is small in size, small in mass, low in power consumption and high in speed and is designed based on a multilayer single / double sensitive area Si detector and a cadmium zinc telluride (CZT) detector array. The neutron / gamma spectrometer can serve as an independent load to be carried on a small satellite, and measurement of thermal neutrons, fast neutrons below 20 MeV and gamma rays is achieved. Different layers of the detector group in the detection equipment are matched to form a plurality of anticoincidence groups, so that the accuracy of energy spectrum inversion is improved. According to the detection method, a stacking and discarding algorithm is provided, so that interference caused by signal overlapping is avoided, and the system resolution of the whole neutron / gamma spectrometer is improved.

A heat pipe cooled nuclear reactor core (“reactor core”) includes an array of fuel, an array of primary heat rejection heat pipes, and an array of moderator cooling heat pipes or thermosiphons. Each moderator cooling heat pipe in the array of moderator cooling heat pipes is surrounded by metal hydride moderator materials, slowing neutrons from a fission energy range to a lower energy range in the reactor core.

The invention provides a solid material doping detection method, device and system. The method comprises the following steps: acquiring first intensity before fast neutrons transmit a to-be-detected material and second intensity after the fast neutrons transmit the to-be-detected material, and gamma-ray energy spectrum data of the to-be-detected material; obtaining a neutron macroscopic cross section of the to-be-detected material according to the first strength and the second strength; judging whether doping exists or not according to the neutron macroscopic cross section of the to-be-detected material and the neutron macroscopic cross section of a preset undoped material; and according to the gamma-ray energy spectrum data of the to-be-detected material, obtaining the element category and content in the to-be-detected material. According to the method, the doping condition of the solid material to be detected is quantitatively and nondestructively judged, whether the doping possibility exists in the material or not is quickly judged by detecting the change of the penetration yield of fast neutrons, and deeper analysis is carried out based on gamma-ray energy spectrum data; the detection process is carried out in a full-automatic mode, human intervention is not needed, and many defects of manual detection are overcome.

The invention relates to the field of theoretical calculation simulation of a nuclear reactor core state, in particular to a simulation method for dynamic reactor core power distribution measurement and a related product, and the method comprises the following steps: obtaining accurate burnup data; establishing a steady-state reactor core theoretical model based on accurate burnup; solving a three-dimensional neutron kinetic equation by taking the steady-state reactor core theoretical model as an initial state and taking a relation curve of reactor core operation power and operation time as input to obtain a dynamic reactor core theoretical model taking time as a function; inputting the power and time relation corresponding to the power and time relation corresponding to the power plant test plan into the dynamic reactor core theoretical model to obtain a database; according to the method, the steady-state theoretical model is constructed by utilizing accurate burnup data, dynamic reactor core simulation is carried out through a transient reactor core neutronics calculation model and the like, a dynamic reactor core model completely consistent with an actual test state can be generated, and actually measured data of dynamic reactor core power distribution can be directly processed and accepted.

The invention provides a samarium / europium / gadolinium-containing zirconium hydride block as well as a preparation method and application thereof, and relates to the technical field of radiation shielding materials. The zirconium hydride block containing samarium / europium / gadolinium comprises the following elements in percentage by mass: 95% < = Zr < = 99.5%, 0% < = Gd < = 5%, 0% < = Sm < = 5%, 0% < = Eu < = 5% and 0% lt. Gd + Sm + Eu < = 5%. The preparation method comprises the following steps: respectively hydrogenating sponge zirconium, metal gadolinium, metal europium and metal samarium to respectively obtain corresponding hydride powder; weighing zirconium hydride powder, gadolinium hydride powder, europium hydride powder and samarium hydride powder according to a required proportion, and uniformly mixing under a protective atmosphere; and carrying out pressure sintering on the mixed powder, and carrying out secondary hydrogenation to obtain the zirconium hydride block containing samarium / europium / gadolinium. According to the invention, the absorption of the material to low-energy-segment neutrons can be obviously improved; and the preparation of the zirconium hydride block containing samarium / europium / gadolinium and the accurate regulation and control of doped elements are realized.

The invention relates to the field of shielding materials, and provides lead-free rubber with ionizing radiation shielding performance and a preparation method thereof, and the lead-free rubber is mainly prepared from barium titanate / boron nitride composite filler, barium tungstate / boron carbide composite filler, tungsten powder, bismuth powder, a rubber matrix and an auxiliary agent according to a proportion. The barium titanate particles are uniformly loaded on the surface of the flaky boron nitride, so that the gamma ray and X ray shielding performance is improved, and meanwhile, the boron nitride provides excellent neutron shielding capability; barium tungstate particles are loaded on the surfaces of the boron carbide nanowires to form a multi-scale shielding network, and the shielding effect of neutrons and ionizing radiation is remarkably enhanced. The rubber matrix endows the material with excellent flexibility and mechanical properties. By scientifically optimizing the component proportion and the preparation process, the problems that a traditional high-density shielding material is prone to agglomeration, poor in flexibility and the like are solved, and the material shows excellent performance in the fields of gamma ray, X ray and neutron shielding and is suitable for the fields of nuclear energy, medical protection, industrial radiation protection and the like.

The invention relates to the technical field of CLYC neutron-gamma rays, and provides a CLYC neutron-gamma ray screening detection method and device based on deep learning, and the method comprises the steps: carrying out the preprocessing of an original pulse waveform outputted by a CLYC detector, and obtaining clean standard waveform data; constructing a one-dimensional convolutional neural network model, respectively capturing the features of the waveform at different time scales, and identifying the most discriminative feature region in the waveform through an attention mechanism; performing adaptive adjustment on the original hyper-parameter of the one-dimensional convolutional neural network model to obtain an optimized hyper-parameter; establishing a direct mapping relation from an original pulse waveform to a neutron-gamma particle type to obtain a trained model; and carrying out compression optimization processing on the model, and deploying the model to an embedded platform to carry out neutron-gamma ray discrimination. According to the method, automatic extraction of deep discrimination features of the CLYC pulse waveform and accurate identification of key feature areas are realized, and the discrimination accuracy and noise interference resistance of neutron-gamma rays are improved.

The invention discloses a hybrid neutron gamma detector array and a signal processing method, and relates to the technical field of nuclear physical detection.During detection, eight EJ200 scintillators arranged in an array are used for energy deposition of gamma rays and fast neutrons and generating photons, and meanwhile three EJ426 thermal neutron scintillation screens are used for capturing thermal neutrons and generating photons; the response of gamma rays, fast neutrons and thermal neutrons is realized at the same time, and the detection precision is greatly improved. Meanwhile, if the three EJ426 thermal neutron scintillation screens are alternately stacked front and back, the total reflection of the EJ200 scintillators and the semi-light-proof performance of the EJ426 thermal neutron scintillation screens are utilized to block photon propagation in the X direction and only enable photons to be propagated in the Y direction and the Z direction, and each EJ200 scintillator corresponds to a four-channel photon collection structure to perform high-efficiency collection of the photons, so that the high-efficiency collection of the photons is realized. Therefore, high-precision detection and high-efficiency photon collection are realized, and support is provided for subsequent intense radiation field monitoring or high-precision positioning.

The invention discloses a GPU accelerated multi-particle coupling Monte Carlo calculation method and device. The method comprises the following steps: asynchronously migrating geometric material parameters subjected to CPU serial processing and a basic database to a pre-allocated memory space in a GPU; operation parameters of the GPU are determined, and Monte Carlo simulation of the multi-particle coupling transportation process is realized in the GPU based on the operation parameters and the neutron nuclear reaction database; on the basis of an atom addition function of a CUDA parallel computing architecture, energy transfer data, obtained through Monte Carlo simulation, of particles and voxel grids in a multi-thread environment is efficiently summarized, energy deposition of the particles is determined, and a high-resolution three-dimensional energy deposition spatial distribution data set is generated. According to the method, high precision can be maintained, and meanwhile, the parallel computing capability of the GPU is fully utilized to realize multi-particle coupling transportation synchronous simulation, so that the computing efficiency of a complex radiation field is improved, and the dual requirements of different fields on high precision and real-time performance are met.

The invention discloses a vanadium-based neutron transparent alloy material and a preparation method thereof, the alloy is a disordered solid solution of a body-centered cubic structure formed by a single crystal phase, the chemical composition is V < 1-x > Mx, x is less than 0.1, V is a metal element vanadium, and M is selected from one or more of Al, Nb, Ta, Fe, Sn or Mo. According to the alloy, neutron diffraction peaks generated by metal vanadium can be effectively eliminated while a metal vanadium body-centered cubic crystal structure is not damaged, the neutron transparent alloy material is successfully prepared, the process is simple and feasible, the formula has diversified choices, multiple novel double-element vanadium-based neutron transparent alloys are prepared, and the neutron transparent alloy material has good application prospects. In addition, the multi-element vanadium-based transparent neutron alloy is prepared for the first time, all the materials have excellent chemical and mechanical properties, and compared with an existing Ti2.08Zr alloy, the multi-element vanadium-based transparent neutron alloy has better high-temperature chemical stability.

The invention relates to the technical field of yield control, in particular to a neutron generator yield control method and system. The method comprises the following steps: establishing a polynomial model by using historical yield data to predict a target efficiency attenuation factor; dynamically determining neutron event boundaries in the pulse integral value and the pulse width two-dimensional characteristics based on the attenuation factor, and counting the neutron event boundaries; the yield deviation and the change rate are calculated, and when the deviation exceeds a threshold value, feedforward compensation is carried out on the target yield; and selecting a three-dimensional lookup table, an input deviation, a change rate and a current high voltage according to the interval to which the attenuation factor belongs, and cooperatively determining the adjustment amount of the ion source current and the accelerator high voltage. According to the scheme, the measurement and control problems caused by aging of the target piece can be solved, and the precision, the response speed and the long-term stability of yield control are improved.

A method and systems for taking downhole measurements. The method may include selecting a pulsing scheme for taking one or more measurements using a pulsed neutron logging tool, selecting a neutron burst width for the pulsing scheme based at least in part on a neutron tube utilized by the pulsed neutron logging tool to form at least in part a neutron burst train, and selecting a decay window in which the one or more measurements are performed by the pulsed neutron logging tool. The method may further include selecting a starting time for a late gate in which the one or measurements are taken, disposing the pulsed neutron logging tool into a borehole, performing the neutron burst train with the pulsed neutron logging tool, and performing the one or more measurements with the pulsed neutron logging tool during the late gate.

A neutron generator with an ion source within a housing may be used for generating neutrons for neutron logging downhole in a wellbore. The ion source within the housing of the neutron generator may include a hot cathode, an ion source cylinder, a first grid separated from the ion source cylinder, and an extractor separated from the ion source cylinder, the extractor having a second grid.

The invention relates to a fuel element damage monitoring method and device based on delayed neutron energy spectrum measurement. The method comprises the following steps: acquiring a delayed neutron zone for fuel element damage monitoring; according to the delayed neutron partition and the delayed neutron partition multi-column energy spectrum measurement principle, a moderator for fuel element damage monitoring is obtained; according to the moderation body and a discrimination threshold automatic identification algorithm, determining a discrimination threshold value of fuel element damage monitoring; and according to the discrimination threshold value, a few-channel spectrum unfolding algorithm and an energy spectrum peak searching algorithm, identifying neutron energy spectrum change conditions and delayed neutron precursor nuclides of fuel element damage monitoring. According to the invention, on-line real-time measurement of the delayed neutron energy spectrum can be realized, and the neutron energy spectrum change condition can be determined; according to the invention, when the delayed neutron precursor nuclide is identified, the occurrence of effective fuel element damage is judged, the online monitoring of the fast reactor fuel element damage is realized, and the rapidness and timeliness of fuel damage monitoring are improved.

The invention discloses a moderated neutron energy spectrum prediction method based on deep learning, and the method comprises the steps: firstly obtaining a data set of neutron energy spectrum parameters through Monte Carlo simulation calculation, and carrying out the preprocessing of the data, including feature extraction, normalization, noise reduction and other means; then, a training set and a test set are constructed, a deep neural network model is constructed based on the multi-layer perceptron, and training is carried out through a small-batch training method; according to the model, a residual block is adopted to replace a hidden layer to enhance the learning ability of the model, and a dropout layer is utilized to prevent overfitting of the model. In the training process, model parameters are updated by calculating a mean square loss function and back propagation, so that the purposes of reducing prediction errors and improving prediction precision are achieved. The method can accurately predict the moderated neutron energy spectrum, and has an important value for application of artificial intelligence in the nuclear medicine field.

The invention relates to an intelligent design method of a radiation shielding material based on cross-scale error feedback lock step, which comprises the following steps: determining a material component set and acquiring an atomic-scale graph structure of the material component set, and extracting intrinsic performance parameters such as heat conductivity and attenuation coefficient through an atomic-scale prediction module; and then, constructing a design variable vector in combination with a multi-physical field condition, calculating equivalent performance parameters of the composite material, establishing a cross-scale lock step mechanism, and triggering a molecular dynamics and Monte Carlo recalculation module to carry out correction when the deviation between a predicted value and a mesoscopic recalculation result exceeds a limit. Based on the correction parameters, a candidate material scheme is generated by using a multi-objective optimization algorithm; and performing gamma-ray and neutron shielding performance simulation on the candidate schemes through a Monte Carlo radiation transfer model accelerated by a deep neural network, and comparing with an experimental result to realize self-adaptive material design driven by cross-scale errors. According to the method, the design precision and efficiency of the radiation shielding material are remarkably improved.

The invention provides an on-site neutron energy spectrum and dose parameterization real-time measurement method. The method comprises the following steps: representing an on-site neutron energy spectrum by using a three-section type function model; constructing simulated field neutron energy spectrums of various moderated levels, measuring by using a detector, and calculating a thermal fast ratio; simulating and calculating a simulated field neutron energy spectrum of the test point through a Monte Carlo model, and obtaining model parameters; establishing a corresponding relation between the model parameters and the thermal fast ratio; a detector is used for measuring the fast neutron pulse counting rate and the thermal neutron pulse counting rate on site, and according to the corresponding relation between the model parameters and the thermal fast ratio, the values of the model parameters are determined in a table look-up or interpolation mode; and calculating a field neutron energy spectrum according to the model parameters, and calculating the field neutron dose by using the field neutron energy spectrum and the neutron fluence-dose conversion coefficient.

A downhole tool is provided with a neutron generator configured to emit neutrons into a geological formation. The downhole tool includes one or more neutron detectors configured to detect neutrons that return to the downhole tool after interacting with the geological formation. The downhole tool also includes one or more gamma ray detectors configured to detect gamma rays from the geological formation that form when neutrons are inelastically scattered by the geological formation. Measurements from a combination of detectors of at least one of the one or more neutron detectors and at least one of the one or more gamma ray detectors are used to determine formation density. A first formation density determined using a first combination of detectors is used to compensate a second formation density determined using a second combination of detectors.

The invention relates to a neutron spectrum measurement method and system, and the method comprises the steps: obtaining a neutron signal obtained through the detection of an incident neutron beam through a plurality of detectors; after the neutron signal is processed, a processed signal is obtained; obtaining the neutron counting rate of each detector based on the processing signal; the initial neutron counting rate of the jth energy interval in the incident neutron beam is set as # imgabs0 #, the initial neutron counting rate is substituted into a formula for iterative calculation, and when the absolute value of the difference value of # imgabs1 # and # imgabs2 # is smaller than a preset first threshold value, # imgabs3 # is the neutron counting rate of the jth energy interval in the incident neutron beam; and calculating the neutron counting rates corresponding to the n energy intervals in the incident neutron beam corresponding to each detector in sequence to obtain the neutron energy spectrum of the incident neutron beam. The measuring method is simple in measuring process and high in accuracy.

The invention relates to the technical field of steel smelting scrap steel briquette treatment, and particularly provides a neutron generator-based scrap steel briquette component online detection device and method. Neutrons generated through nuclear reaction of a neutron generator are absorbed and activated with elements such as Fe in the scrap steel, gamma rays with characteristic energy are emitted, then the energy intensity of the gamma rays is observed, contrasted and calibrated from the response of a detector, and therefore the content composition and the content proportion of all the elements, mainly the Fe element, in the scrap steel briquette are calculated, analyzed and determined. The device effectively avoids the low-quality employment risk in the detection of the scrap steel briquettes, solves the problems that the detection of the internal quality of the scrap steel briquettes is difficult, time-consuming and labor-consuming, remarkably improves the working efficiency and effect, and has remarkable economic benefits and popularization prospects.

The invention discloses a reactor core neutron field ex-situ explicit inversion method based on a graph structure, and belongs to the technical field of reactor core safety monitoring and control. According to the method, reactor core space coordinates, energy spectrum grouping, material physical properties and prior information are fused through an explicit graph structure, a lightweight graph convolutional network architecture is adopted, and accurate inversion of a high-dimensional reactor core neutron field is achieved. And meanwhile, the inversion precision under the high-dimensional resolution can be remarkably improved, and reliable technical support can be provided for safe operation and intelligent monitoring of an advanced reactor type. Compared with an existing deep learning black box model, the method has the advantages that the occurrence probability of local high deviation can be remarkably reduced at different reactor core change positions, and more uniform and stable prediction performance can be realized in the whole space. And under the condition that the reactor core has multi-region change, better inversion performance is still shown. When the middle and low complexity changes, the fine multi-region features of the reactor core can be effectively and stably captured, and the method has better overall error suppression capability.

A method includes: during a first time period, pumping propellant: to a set of moderator coolant channels to cool the moderator; through a set of fuel elements to heat the propellant to a first outlet temperature; out of a thrust nozzle to produce a first thrust; and locating the reflector at a first closed position to reflect neutrons toward the set of fuel elements to maintain a fission rate of the set of fuel elements above a threshold fission rate. Additionally, during a second time period: locating the reflector at a second open position to reduce the fission rate of the set of fuel elements to below the threshold fission rate; and pumping the propellant through the fuel elements to heat the propellant to a second outlet temperature less than the first outlet temperature and out of the thrust nozzle to produce a second thrust less than the first thrust.

The invention relates to a device and a method for measuring neutron shielding capability of concrete at high temperature, the device comprises a neutron emitter, a shielding device and a neutron receiver, the neutron emitter is located at a first side of the shielding device, and the neutron receiver is located at a second side, opposite to the first side, of the shielding device; the shielding device comprises a support and samples, the samples are arranged on the support, the samples comprise three test samples and a measurement environment background sample, the three test samples are a control group, a cylindrical concrete test piece and a detection piece group in sequence, and the cylindrical concrete test piece group is arranged on the support. An ultrahigh heat-conducting polymer film layer and a stainless steel alloy sleeve layer are sequentially arranged on the outer side of the cylindrical concrete test piece, and the detection piece comprises a temperature sensor and a strain gauge. The neutron shielding capability is measured by adopting the measuring device, so that the neutron shielding capability of the new-generation molten salt reactor nuclear energy system warehouse concrete is effectively measured, and the thickness design of the neutron shielding concrete in engineering construction is facilitated.

The invention belongs to the field of nuclear radiation detection systems, and provides a portable neutron gamma detection system based on a NaIL crystal, the portable neutron gamma detection system comprises a detection device, a tablet computer, a support and a handle, the support and the handle are installed on the detection device through a plurality of groups of fastening bolts and spring washers, the tablet computer is fixed on the support through a buckle, and the handle is fixed on the support through a buckle. The detection device comprises a shell, a polyethylene moderator, a NaIL crystal, a photomultiplier, a front-end electronics part, a rear-end electronics part and a wireless communication antenna. The device is simple in structure and easy to hold and use or carry by a backpack, only one detection element is used, the structure of the detection device is greatly simplified, the front-end electronics part and the rear-end electronics part only need to consider processing of one path of signals, the equipment cost and the design difficulty are remarkably reduced, and the reliability is remarkably improved.

The invention discloses a neutron and gamma ray integrated aluminum-based nuclear shielding material. The aluminum-based nuclear shielding material comprises, by mass, 1%-10% of boron carbide, 5%-30% of high-entropy alloy powder and 60%-94% of aluminum alloy powder. The invention further discloses a preparation method of the aluminum-based nuclear shielding material. The preparation method comprises the steps that firstly, boron carbide, high-entropy alloy powder and aluminum alloy powder are mixed and then subjected to ball milling; step 2, crushing; 3, cold pressing and hot pressing sintering; and 4, hot rolling and solid solution aging heat treatment. According to the neutron and gamma ray integrated aluminum-based nuclear shielding material, the gamma ray shielding efficiency and the mechanical property of the aluminum-based nuclear shielding material are improved by introducing the high-entropy alloy powder, the neutron and gamma ray integrated aluminum-based nuclear shielding material high in shielding efficiency and excellent in comprehensive mechanical property is obtained, and the neutron and gamma ray integrated aluminum-based nuclear shielding material is suitable for the technical fields of small reactors, movable reactors, nuclear submarines and the like.

The embodiment of the invention discloses a composite shielding plate and a preparation method thereof, and the composite shielding plate comprises a cladding which is provided with an accommodation space, and the accommodation space is in a vacuum state; the shielding material is arranged in the accommodating space; wherein the shielding material comprises a neutron moderation base material, and the cladding is made of a boron-containing metal material; or the shielding material comprises a neutron moderation base material and a neutron absorption base material, the cladding is made of a metal material, the neutron moderation base material comprises rare earth hydride, and the neutron absorption base material comprises boron-containing ceramic. The composite shielding plate provided by the embodiment of the invention not only comprises the neutron moderation base material which is high in hydrogen content and excellent in temperature resistance so as to effectively moderate fast neutrons, but also is combined with the neutron absorption base material containing boron atoms so as to efficiently absorb thermal neutrons, so that the complete neutron moderation-absorption integrated shielding performance of the composite shielding plate is realized; and stable shielding performance can be provided at the temperature of 300-1000 DEG C.

Passivation regions and device configurations are described herein. The passivation regions can be configured to seal against diffusion of an objective material from an underlying region into and / or through the passivation region. The passivation regions can also be configured to seal against diffusion of an externally sourced or ambient substance into and / or through the passivation region towards the underlying region.

The invention relates to a method for detecting boron distribution in an oxidation corrosion product deposition layer. The total mass of boron in the oxidation corrosion product deposition layer is detected by utilizing an instantaneous gamma activation technology; the method comprises the following steps: obtaining a geometric boundary of a deposition layer of a corrosion product according to neutron photography, constructing a Compton cone in the geometric boundary obtained by neutron photography according to instant gamma rays generated by the oxidation corrosion product and Compton scattering information generated by the action of a Compton photography unit, and reconstructing the origin of the gamma rays; and carrying out pixelated image reconstruction by adopting a random origin set algorithm to obtain a distribution image generated by gamma rays so as to obtain the proportional distribution of the boron element in the oxidation corrosion product deposition layer. According to the obtained total boron element concentration and boron element proportion distribution, boron element concentration distribution in the oxidation corrosion product deposition layer is calculated, and the defects that only boron element distribution in the limited depth of the surface of the corrosion product deposition layer can be measured and internal element distribution information cannot be obtained in an existing method are overcome.

The invention provides a method for determining radial power density distribution of a fuel rod and a product, and relates to the technical field of nuclear control. The method comprises the following steps: acquiring the atomic density and neutron fluence rate of each nuclide in each radial ring; determining a volume power density function of each radial ring according to the atomic density and the neutron fluence rate; performing normalization processing on the volume power density function to obtain a radial power density distribution factor; determining the product of the average power of the axial section and the radial power density distribution factor corresponding to the radial ring as the local power corresponding to the radial ring; and determining the radial power density distribution of the fuel rod by using the local power. According to the method, the plurality of local powers are determined according to the atomic density and the neutron fluence rate of each nuclide, and the radial power density distribution of the fuel rod is determined according to the plurality of local powers, so that the radial power density distribution of the fuel rod can be determined without being limited to a fixed nuclide category; and the accuracy of performance analysis and design of the new reactor type fuel rod can be improved.

The invention provides a doped yttrium hydride neutron moderation material and a preparation method and application thereof, and belongs to the technical field of radiation shielding materials. At least one of samarium, europium and gadolinium is doped in yttrium hydride in the form of hydride, and the mass ratio of metal yttrium to metal gadolinium to metal europium to metal samarium is controlled, so that the neutron moderation effect of the neutron moderation material can be improved. According to the invention, at least one of gadolinium hydride, europium hydride and samarium hydride is mixed with yttrium hydride under the protective atmosphere, so that the mass content of each element in the doped yttrium hydride neutron moderation material can be accurately controlled. According to the invention, the mixture is sintered and molded, and the sintering parameters of sintering and molding are controlled, so that the components of the mixture are sintered in a compact state, cracks generated by thermal expansion and cold contraction can be reduced, and the doped yttrium hydride keeps high density and has fewer cracks and defects.