538 results about "Neutron" patented technology

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 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 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 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 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.

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 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.

The invention discloses a radiation shielding method and a radiation shielding body based on a multi-physics field synergistic effect, and the radiation shielding body is used for shielding a neutron-gamma mixed radiation field, and the radiation shielding method based on the multi-physics field synergistic effect comprises the steps: a first stage, absorbing neutrons and gamma rays at a radiation initial position through a first composite material; in the second stage, energy spectrum adaptive attenuation is carried out on the first residual radiation passing through the first stage through a second composite material, and the amount of the second composite material is gradually decreased in a gradient mode in the radiation propagation direction; and a third stage of performing path scattering on the second residual radiation passing through the second stage through a third composite material. According to the invention, the technical problem of insufficient shielding capability caused by difficulty in adapting to the dynamic shielding requirement of a neutron-gamma mixed radiation field when a traditional anti-radiation shielding body faces the neutron-gamma mixed radiation field generated in a BNCT treatment scene is solved.

The invention relates to the technical field of temperature measurement of crystal materials, and relates to a temperature measurement method for SiC crystal thermogravimetric analysis based on Si < + > ion and neutron dual irradiation, and the method specifically comprises the following steps: S1, dual irradiation of a SiC crystal; s2, annealing treatment; s3, carrying out thermogravimetric test; s4, processing thermogravimetric test data; s5, measuring the temperature of the environment to be measured; the expression form of the temperature calibration equation obtained in the step S4 is b = alpha + beta T * + gamma T * 2 + delta T *; according to the temperature measurement method for thermogravimetric analysis of the Si < + > ion and neutron dual-irradiation SiC crystal, the internal structure of the SiC crystal is changed by performing annealing treatment on the SiC crystal at different temperatures, thermogravimetric testing is performed on the SiC crystal subjected to annealing treatment at different temperatures, data are analyzed, a polynomial equation is established, and the extremely high environment temperature can be effectively tested; and the method has relatively high accuracy.

A system may include one or more downhole tools, where a first downhole tool includes a pulsed neutron generator to emit neutrons into a borehole of a geological formation and one or more gamma-ray detectors to obtain a measurement of gamma-ray emissions, of a borehole environment, induced by the emitted neutrons. The system may also include data processing circuitry to determine a Sigma value associated with the borehole environment based on the measurement and determine elemental concentrations, excluding lithium, based on one or more gamma-ray energy spectra obtained via the one or more downhole tools. The data processing circuitry may also determine an elemental Sigma contribution of the elements other than lithium based on the elemental concentrations, determine a lithium Sigma contribution based on a difference between the Sigma value and the elemental Sigma contribution, and determine a lithium concentration within the borehole environment based on the lithium Sigma contribution.

A system and a method for a pulsed neutron logging tool. The pulsed neutron logging tool may include a pulsed neutron generator for generating one or more neutrons, one or more gamma ray scintillator detectors disposed in a detection area, a shielding block that separates the pulsed neutron generator from the one or more gamma ray scintillator detectors, and a shielding layer. The pulsed neutron logging tool may further comprise an information handling system communicatively connected at least in part to the pulsed neutron logging tool. The information handling system is configured to create a gamma ray spectrum from the one or more gamma rays detected by the one or more gamma ray scintillator detectors, remove a background from the gamma ray spectrum to form a peaks-only spectrum, and compare one or more peaks-only elemental standards to the peaks-only spectrum to determine an elemental weight and yield.

The invention belongs to the technical field of marine radiation protection, and discloses a marine containment vessel shielding iterative optimization design method based on regional equal dose, which comprises the following steps: completing transport calculation of neutrons and gamma radiation fields in a containment vessel to obtain radiation distribution of the inner wall of the containment vessel; obtaining a thickness distribution initial scheme of each shielding layer of the containment according to a containment outer wall dose rate design target value; fine transport calculation of neutrons and gamma radiation in the containment and the shielding material is completed, and the radiation distribution of the outer wall of the containment in the scheme is obtained; a containment outer wall dose rate design target value is combined, and an iterative optimization algorithm is adopted to obtain a thickness distribution optimization scheme of each shielding layer of the containment; and carrying out multiple rounds of iterative optimization calculation until the thickness difference between the optimized scheme and the scheme before optimization reaches an expected target. According to the method, an optimized marine containment vessel shielding thickness distribution scheme is formed on the basis of a regional equal dose principle, the shielding weight is optimized while the radiation safety of personnel is guaranteed, and the overall resources of a ship are saved.

Techniques are described for producing actinium-225 by arranging a converter material at least partially around a target material, where the target material includes radium-226. The converter material comprises a compound of lithium and hydrogen at least partially enriched with lithium-6 and deuterium, and is configured to convert thermal neutrons to fast neutrons, which are then incident on the radium-226. If the fast neutrons have a kinetic energy of at least 6.4 MeV, they will initiate a (n,2n) reaction that converts radium-226 into radium-225, which then decays into actinium-225. The process can desirably be performed within an environment rich in thermal neutrons, such as within a fission reactor.

A reactor and method commonly applicable for high-pressure in-situ DSC and neutron tests of a gas hydrate relates to the field of characterization of physical and chemical properties of the gas hydrate. The reactor can not only carry out a high-pressure and low-temperature in-situ DSC test of the hydrate but also be suitable for a neutron diffraction test of the hydrate. The reactor can be adapted to an existing high-pressure and low-temperature in-situ DSC device without the need to re-develop a whole set of system, thus greatly reducing the replacement cost of the device. Owing to the sectional design, the flexibility and the applicability of the reactor can be ensured. Researchers can conveniently transport the hydrate in a pressure-maintaining manner. Even at a long distance, with the assistance of a liquid nitrogen tank or a vehicle-mounted refrigerator, it can be ensured that the hydrate may not be decomposed during transportation.

The invention belongs to the technical field of production of isotopes by a sodium-cooled fast reactor, and aims to solve the problems that the moderation effect is limited, the reactor core disturbance is relatively large and fast neutron energy spectrum is not beneficial to the production of the ultra-uranium isotopes during the production of the ultra-uranium isotopes. The invention provides a method for producing a super-uranium isotope by using a sodium-cooled fast reactor and a sodium-cooled fast reactor core. According to the method, moderator assemblies and super-uranium isotope production assemblies are arranged on a first circle of a radial conversion area in a flower arrangement mode, the super-uranium isotope production assemblies are arranged on a second circle of the radial conversion area, and the side, facing the center of the reactor core, of the first circle is coated with a thermal neutron absorption material; the reactor core is provided with a radial conversion area, moderator assemblies and super-uranium isotope production assemblies are arranged on a first circle of the radial conversion area, the super-uranium isotope production assemblies are arranged on a second circle of the radial conversion area, and thermal neutron absorbing materials are arranged on the side, facing the center of the reactor core, of the first circle. The super-uranium isotope is concentrated in the radial conversion region for production, the moderation effect is good, the disturbance to the reactor core is small, and the productivity is high.

The utility model provides a rotary neutron target. The rotary neutron target comprises a neutron generation layer, a hydrogen diffusion layer and a cooling device, the cooling device comprises a cooling substrate and a rotating shaft, the cross section of the cooling substrate is circular, and the rotating shaft is vertically and fixedly connected to the circle center of the cooling substrate; the cross sections of the neutron generation layer and the hydrogen diffusion layer are of a first annular structure and a second annular structure respectively, and the first annular structure and the second annular structure are the same in shape and area. The hydrogen diffusion layer, the neutron generation layer and the cooling substrate are coaxially arranged, the hydrogen diffusion layer is arranged above the cooling substrate, and the neutron generation layer completely covers the hydrogen diffusion layer. According to the structure, all protons enter the hydrogen diffusion layer, and heat generated by the neutron generation layer and the hydrogen diffusion layer on the rotating neutron target is rapidly taken away, so that the heat is uniformly distributed, and high temperature of the target material caused by proton aggregation is avoided; according to the structure, proton energy deposition is basically in the hydrogen diffusion layer, the neutron generation layer is effectively protected, and the service life of the neutron target is prolonged.

The invention discloses a composite shielding collimator suitable for a complex mixed radiation field, belongs to the technical field of radiation shielding and measurement, and mainly solves the problems of low signal-to-noise ratio, high space background and mixed energy spectrum caused by the mixed radiation field of high-energy neutrons, gamma rays and the like in a fusion device and the problem that traditional equipment is heavy and difficult to flexibly adjust. The composite shielding collimator adopts a cylindrical shielding structure formed by aluminum oxide, boron-containing polyethylene, boron carbide and lead from outside to inside, and background interference and secondary radiation are remarkably inhibited by moderating and absorbing neutrons step by step and effectively shielding gamma rays; the neutron and gamma ray detector is integrally integrated on a rotatable supporting mechanism, unification of excellent radiation shielding performance and detection flexibility is achieved, and the neutron and gamma ray detector is mainly used for high-precision neutron and gamma ray diagnosis in complex radiation environments such as a magnetic confinement fusion device.

A radiation irradiation system and a control method therefor. On the basis of weight proportions of elements in a human body and reaction intensities the elements with a neutron and a photon, the element that has influence on simulation calculation results of the neutron and the photon in an application scenario of the radiation irradiation system is screened out, and during a simulation process, only the screened-out element is simulated, so that the calculation speed can be greatly improved, and the calculation time can be reduced.

Disclosed are a device and method for water-pumping-injection single-sphere neutron spectrum timing measurement. The device includes a double-layer nested shell, a central detector, a water pump, a water injector, a sprayer and a timing analysis system. The central detector is located at a center of an inner shell, and a single-layer moderation cavity for containing moderator liquid is formed between the inner shell and an outer shell. The sprayer, the water pump and the water injector cooperate to continuously adjust a water flow state and a mist sedimentation effect in the moderation cavity, and a moderator layer with a variable thickness is formed to meet measurement needs of different energy neutrons. The central detector is configured for detecting the thermal neutrons and the prompt gamma-ray signals. The device is compact and lightweight overall, and easy to transport and deploy on site, which facilitates promotion and application.

A nuclear engine system includes: a pump configured to pump a propellant; a fuel element including a set of nuclear fuel particles; a moderator configured to surround the fuel element and defining a set of moderator coolant channels configured to cool the moderator; a reflector including a neutron-reflecting material and a reflector coolant channel arranged within the reflector to cool the reflector, the reflector configured to at a first time, operate in a closed configuration to reflect neutrons to the fuel element to increase an energy flux, and at a second time, operate in an open configured to leak neutrons out of the engine system to decrease the energy flux; and a thrust nozzle configured to outlet propellant from the fuel element to produce thrust including a nozzle coolant channel arranged within a wall of the thrust nozzle configured to cool the nozzle.

The invention relates to a moderator of a compact single-ball neutron spectrometer, and the moderator is characterized in that an inner layer is of a cylindrical structure, and a plurality of thermal neutron detectors are installed in the inner layer; the thermal neutron absorption layer sleeves the outer side of the inner layer and is made of boron-containing polyethylene; the scattered neutron moderation layer sleeves the outer side of the thermal neutron absorption layer; the inner layer comprises a detector placing block and a thermal neutron absorption block, and the thermal neutron absorption block is provided with a storage groove; a plurality of thermal neutron detectors are placed in the detector placement block along the axis direction, the first thermal neutron detector is placed in the storage groove, and the thermal neutron absorption block is made of boron-containing polyethylene; the diameter of the inner layer is 8 cm, the outer diameter of the thermal neutron absorption layer is 14 cm, and the outer diameter of the scattered neutron moderation layer is 20 cm. The purposes that the size is controlled by controlling the diameter of the cylindrical structure and the scattering shielding effect is enhanced by arranging the three layers of moderator structures in a sleeving mode are achieved, and therefore the technical effects that the moderator is small in size, convenient to transport and high in scattering shielding effect are achieved.

The invention provides a nuclear-grade hafnium plate preparation method and a nuclear-grade hafnium plate, and belongs to the technical field of nuclear fission reactor control rod neutron absorber material processing.The nuclear-grade hafnium plate preparation method comprises the steps that nuclear-grade sponge hafnium is adopted as a raw material, the oxygen content of a hafnium cast ingot is controlled within a specific range through a smelting technology, and the more obvious strengthening effect is achieved; and the strength of the finished hafnium plate can meet the nuclear pole requirement. Meanwhile, the deformation resistance of the material is increased, so that the problem that the high-strength hafnium plate is difficult to machine is further solved through optimal combination of forging upsetting, hot rolling, annealing, one-pass cold rolling, intermediate annealing, two-pass cold rolling and finished product heat treatment parameters, and the nuclear-grade hafnium plate with various properties meeting the technical requirements is prepared; the tensile strength and the welding strength of the hafnium plate are improved, the process is stable, and the yield reaches 95% or above.

The utility model provides a large-size curved-surface thin-wall neutron beam window structure, which is applied to a high-resolution neutron diffractometer and aims to provide a high-vacuum environment for an experimental sample and a scattered neutron channel and ensure high transmittance of neutrons and mechanical strength and stability of a device at the same time. The neutron beam window structure is an integrated curved surface type structure, and the thickness of the beam window is 1.5 mm. According to the technical scheme, through strict quality detection and fatigue test, the stability and durability of the neutron beam window in actual use are ensured. In addition, potential problems can be found and processed in time through regular performance state monitoring, so that the service life of the beam window is prolonged, the maintenance cost is reduced, and more reliable and efficient support is provided for scientific research activities such as neutron diffraction experiments.