27 results about "Nuclear technology" patented technology

The invention belongs to the technical field of nuclear, and relates to a high-precision full-size conical neutron conduit regulation and control system and an integration method thereof. The system is supported by a bottom storage cabinet, a marble base, a top integration bottom positioning part, two groups of horizontal and top feeding regulation and control components and a positioning block are arranged on the bottom storage cabinet, a conical standard sample block is arranged between the components, and a pneumatic system is arranged in the bottom storage cabinet; the whole process of the system is integrally spliced, an external auxiliary structure is not needed, and the operation efficiency is improved; multi-specification adaptation and size expansibility are achieved, and catheters with different opening sizes can be rapidly switched to be spliced; the positioning adsorption unit performs processing according to a set taper to meet diversified requirements; a micrometer assembly and the like ensure that the splicing form and location tolerance is less than or equal to 20 microns, and the assembly error is reduced; the temperature deformation influence is reduced through material and process innovation, and the splicing quality is guaranteed; standardized interfaces are convenient to disassemble, assemble and maintain, and the technical threshold is reduced by manual regulation.

The application provides an oxygen concentration interval checking method, and relates to the technical field of lead-bismuth fast reactor oxygen concentration interval checking, and the method comprises the following steps: establishing a lead-bismuth fast reactor model, obtaining an initial oxygen concentration interval of the lead-bismuth fast reactor model, and selecting a plurality of preset oxygen concentrations from the initial oxygen concentration interval; performing physical field calculation on the fuel rod under each preset oxygen concentration to obtain physical field parameters corresponding to each preset oxygen concentration; the physical field parameters comprise temperature distribution parameters, fuel rod geometric parameters, contact stress, cladding oxidation corrosion parameters and fission gas parameters; judging whether each preset oxygen concentration meets the use requirements of the fuel rod based on the physical field parameters corresponding to each preset oxygen concentration, and determining a qualified oxygen concentration interval based on each preset oxygen concentration that meets the use requirements of the fuel rod. The qualified oxygen concentration interval is obtained by checking the initial oxygen concentration interval, and the problems of heat transfer deterioration, inability to inhibit corrosion and inability to meet the long-term use requirements of the fuel rod during fuel rod burnup can be avoided.

The invention relates to the technical field of nuclear, in particular to a device for detecting selenium content in fruits, which comprises a D-D neutron generator and a measuring chamber, the D-D neutron generator generates neutrons, the neutrons in the measuring chamber enter a sample to generate nuclear reaction with atomic nucleuses in the sample in the measuring chamber, and the neutrons in the measuring chamber are generated; after the atomic nucleus of the measuring chamber is activated, instant gamma rays and delayed gamma rays are released; the high-purity germanium detector in the measuring chamber receives delayed gamma rays of all elements in the sample in the measuring chamber and converts each gamma photon into an electric pulse signal; the multichannel analyzer is used for converting the electric pulse signals of the measuring chamber into digital signals, dividing channel addresses according to the size of the digital signals, counting the pulse number and forming a gamma ray energy spectrum; and the computer in the measuring chamber calculates the selenium content in the sample in the measuring chamber through the gamma ray energy spectrum in the measuring chamber. The method has the advantages of low detection limit, small error and high measurement result precision, and can quickly realize the detection of the selenium content in the fruits.

The invention discloses a radiation hazard source identification method with a low counting rate, and relates to the technical field of nuclear technology application. Comprising the following steps: establishing a nuclide library containing radionuclide information; setting a peak position interval corresponding to each radionuclide based on the nuclide library; carrying out statistics on the peak position count of the peak position interval and the high-energy area count of the high-energy area, and caching the peak position count and the high-energy area count obtained through statistics as environment background data; estimating parameters according to the cached environment background data; acquiring a peak position counting rate and a high-energy region counting rate in real time; according to the parameters, the peak position counting rate and the high-energy region counting rate which are obtained in real time, the probability that the peak region is abnormal is calculated; performing early warning sequential test on the abnormal probability of the peak region; and comparing a result of the early warning sequential test with a preset early warning threshold value, and if the result reaches the early warning threshold value, outputting early warning information and identifying a corresponding radiation hazard source. According to the invention, rapid early warning and hazard source identification of abnormal radiation data are realized.

The application relates to a split type reactor base cold neutron source and reactor, belonging to the nuclear technology field. The split type reactor base cold neutron source is arranged in a reactor and comprises a vacuum cylinder, a coolant circulating assembly and a cold package assembly. The vacuum cylinder has a vacuum cavity with a ring-shaped cross section. The coolant circulating assembly comprises a coolant input pipe and a coolant output pipe which are communicated to a coolant cavity defined by the inner wall of the vacuum cylinder for coolant input and output. The cold package assembly is a single-layer wall structure and is arranged in the coolant cavity and cooled by the coolant. The cold neutron source has the advantages of simple structure, good cooling effect, convenience in maintenance and repair, high reliability and the like.

The invention provides an accelerated aging test method for a nuclear power non-metallic hose, electronic equipment and a medium, and relates to the technical field of nuclear, the method comprises the following steps: carrying out an accelerated aging test on the nuclear power non-metallic hose, regulating and controlling the strength of multiple aging factors through an aging test platform, and obtaining the strength of the multiple aging factors at different time t; the comprehensive aging rate of the nuclear power nonmetal hose is determined according to the strength of the multiple aging factors; obtaining a difference function between the comprehensive aging rate and a target aging rate; and according to the difference function and the aging factor intensity adjustment function corresponding to each of the aging factors, continuously updating the intensity of the aging factors, and obtaining first aging characterization data of the nuclear power nonmetal hose at different time t in the accelerated aging test process.

The invention relates to the field of nuclear technology and automatic detection, and provides a high-precision automatic sample change detection device for mass neutron imaging experiments and an experiment method. The device comprises a sample library system and an automatic sample switching system, the sample library system can store and manage at least 200 sample boxes, and the automatic sample switching system adjusts the sample boxes to an experiment area and returns the sample boxes; the two systems work cooperatively, batch parallel processing of hundreds of samples is achieved, the problem of difficulty in automatic sample replacement of a large-batch sample experiment is solved, the experiment efficiency is improved, and the consistency and reliability of experiment data are ensured; the device runs fully automatically, experimenters are liberated, and the utilization efficiency of human resources is improved; and the device is suitable for samples of different sizes, wide in applicability, capable of improving the experiment efficiency by 70%, convenient for classified storage and management of the samples, suitable for various neutron line stations, and high in universality and economical efficiency.

The present application relates to the nuclear technology field, specifically to a method for separating thorium from a thorium and lanthanum mixture. The method comprises the following steps: in a three-electrode system, pulse constant current controlled potential electrolysis is performed on a molten salt system containing mixed ions of thorium and lanthanum series elements, and the deposition rate difference of thorium and lanthanum series element ions during pulse constant current electrolysis and the electrolytic reaction limitation of the control potential on the lanthanum series element ions are utilized to achieve efficient separation of thorium from the thorium and lanthanum mixture. The molten salt system containing mixed ions of thorium and lanthanum series elements is obtained by adding the thorium and lanthanum mixture into a chloride molten salt, mixing uniformly, and passing in chlorine gas mixture to chlorinate the mixture and dissolve it into the chloride molten salt. The method has the advantages of simple process, strong radiation resistance, no additional radioactive waste, and low cost, and is suitable for industrial application of molten salt reactor spent fuel reprocessing.

The invention relates to the technical field of nuclear, and discloses a site dose rate measuring device and method. According to the site dose rate measuring device provided by the invention, a physical barrier is constructed through the sleeve pre-embedded in the outer wall of the structure and the closed blind end of the sleeve, a high-dose site is isolated from an external overhaul site, and during measurement, a probe of a monitor only needs to be inserted from the outside through a measuring channel opened by a shielding compensation structure; a worker does not need to carry equipment to enter or get close to a high-dose place, temporary perforation is not needed during measurement, and aerosol disturbance, equipment surface contamination diffusion (cross contamination) and direct external irradiation and internal irradiation risks borne by the worker which are possibly caused by the fact that the worker enters the place during operation are completely eradicated. Meanwhile, as the blind end extends into the high-dose place, the detector completely penetrates through the structural wall body, the radiation field of the place to be detected is directly detected, the dose rate of the real place which is not additionally attenuated by the wall body is measured, and the data is more accurate.

The invention discloses a low-flow-resistance frame type control rod assembly and belongs to the technical field of nuclear. The control rod assembly mainly comprises an upper connecting base, a lower connecting base and a supporting pipe, the supporting pipe is fixedly connected with the upper connecting base and the lower connecting base through connecting rib plates arranged at the two ends respectively, and axial gaps are formed between the upper end face of a pipe body and the lower end face of the upper connecting base and between the lower end face of the pipe body and the upper end face of the lower connecting base. The axial gap provides a flow passage for the coolant. The low-flow-resistance frame type control rod assembly can effectively reduce the flow resistance of a coolant during axial flow, improve the balance of flow distribution of a pool type reactor core and improve the safety of the reactor.

The application discloses a mobile radionuclide production device and a use method thereof, and belongs to the technical field of nuclear technology and nuclear medical technology.The device comprises a movable carrier, a radionuclide generation module, a shielding and protection module and an intelligent control module which are integrated on the movable carrier.The radionuclide generation module comprises at least one commercially available and regulatory approved radionuclide generator.The shielding and protection module is a fixed shielding structure arranged around the radionuclide generation module.The intelligent control module is electrically connected with the radionuclide generation module and is used for controlling the production process of the radionuclide.The mobile radionuclide production device and the use method thereof can realize on-site and rapid production of the radionuclide, and solve the problems of inflexibility of fixed facilities and large attenuation of long-distance transportation of short half-life radionuclides.

A simplified neutron source cold package, a reactor-based cold neutron source, and a reactor are disclosed, belonging to the field of nuclear technology. The simplified neutron source cold package includes a moderator input tube, a moderator output tube, a cold package cavity, and a coolant circulation pipeline. The moderator input tube includes an inflow section parallel to the moderator output tube and a curved recirculation section. The cold package cavity is connected to the recirculation section, and its maximum cross-sectional area covers the azimuth range of the neutron conduit arrangement. The coolant circulation pipeline includes an input jacket, an output jacket, and a cold package cavity jacket, with a vacuum space between the cold package cavity jacket and the input jacket.

The invention relates to the field of nuclear technology, and more particularly to technology for producing radioactive isotopes for medicinal purposes, especially actinium-225, as well as other radionuclides, formed during the charged-particle (proton) irradiation of thorium in an accelerator. A key issue for achieving a high yield of radionuclides during proton irradiation of a target is that of manufacturing the target from sufficiently thick metallic thorium enclosed in a hermetic metallic niobium casing which is resistant at high temperature to chemical and radiation damage. Effective water cooling during irradiation is achieved by assuring good contact of a niobium foil (proton beam entrance and exit windows) with the thorium, as well as with other niobium parts of the casing. For this purpose, during production of the target, diffusion welding is carried out at a temperature of 1250-1430°С and a pressure of 30-150 kg / cm2 on the surface area of the target, wherein a hollow channel is formed in advance in the niobium body of the target to compensate for greater expansion of the thorium during welding. The target may be additionally coated from the outside with a protective layer of from 40 to 100 μm of another metal resistant to water during irradiation.

A cold pack with enhanced heat dissipation capacity, a reactor-based cold neutron source, and a reactor belong to the field of nuclear technology. The cold pack with enhanced heat dissipation capacity includes a moderator input pipe, a moderator output pipe, a cold pack cavity, and a coolant circulation loop. The moderator input pipe includes an inflow section and a curved return section. The cold pack cavity is connected to the return section and located on one side of the inflow section, with its maximum cross-section covering the azimuth area of ​​the neutron conduit. The coolant circulation loop includes an input jacket, an output jacket, and a coolant cavity. The coolant cavity encloses the cold pack cavity, the return section, and part of the inflow section, with the space between the cold pack cavity and the inflow section located within the coolant cavity. This cold pack provides a larger volume of coolant for heat dissipation, effectively improving the heat dissipation capacity and cold neutron output efficiency of the cold pack during neutron moderation.

The invention relates to the field of nuclear technology, in particular to a method for separating thorium from a thorium and lanthanum mixture. The method comprises the following steps: in a three-electrode system, carrying out pulse type constant-current potential control electrolysis on a fused salt system containing thorium and lanthanide series mixed ions, and utilizing the deposition rate difference of thorium and lanthanide series ions during pulse type constant-current electrolysis and the electrolytic reaction limitation of the control potential on the lanthanide series ions to obtain the thorium and lanthanide series mixed ions. And efficient separation of thorium in the thorium and lanthanum mixture is realized. Wherein the molten salt system containing the thorium and lanthanide series element mixed ions is obtained by adding a thorium and lanthanum mixture into chloride molten salt, uniformly mixing, introducing chlorine mixed gas to chloridize the mixture and dissolving the mixture into the chloride molten salt. The method is simple in process, high in irradiation resistance, free of additional radioactive waste, low in cost and suitable for industrial application of molten salt reactor spent fuel post-treatment.

The application provides a radioactive material vertical sealing transportation structure and belongs to the nuclear technology field. When radioactive materials are transported in the sealing channel to the vertical channel, the fork unit forks the radioactive materials from the first horizontal channel or the second horizontal channel. After the fork unit forks the radioactive materials, the control unit controls the first driver and the second driver to move upward or downward at the same time. Since there is a traveling wave magnetic field between the first driver and the first secondary plate, the fork unit moves with the first driver under the driving of the traveling wave magnetic field. There is also a traveling wave magnetic field between the second driver and the second secondary plate, which drives the locking part to move away from the vertical rod. When the power is suddenly cut off, the magnetic field between the second driver and the second secondary plate disappears. Under the action of the spring's own elastic force, the locking part moves close to the vertical rod, thereby playing a braking role on the fork unit, avoiding the out-of-control of the fork unit in the falling process and causing the radioactive material leakage.

1. Name of the product in this design: Portable Surface Contamination Measuring Instrument. 2. Application of this design: for monitoring radioactive contamination and the environment in places such as nuclear power plants, nuclear facilities, nuclear technology application and research institutions. 3. The key design feature of this product is its shape. 4. The image or photograph that best illustrates the design's key points: a 3D model.

The invention belongs to a neutron detector technology in the field of nuclear technology and high-energy physics, and particularly relates to a preparation method and application of a high-precision ceramic GEM based on short pulse laser. The method comprises the following steps: firstly, covering a double-side copper-clad substrate with a solder mask layer, then perforating by using short pulse laser, then carrying out acid etching on the copper layer at the perforated part to form a rim circular ring, removing the solder mask layer, then carrying out plasma cleaning to remove carbonized substances on the hole wall, and finally cleaning, cutting and baking to complete preparation. The laser drilling speed is high and can reach 20000 holes / min, the alignment precision reaches the micron level, ceramic GEM with different hole diameters, intervals and thicknesses can be prepared, laser non-contact machining is free of drill bit loss, and the insulation performance is improved through plasma cleaning; when the ceramic GEM prepared by the method is used for a neutron detector, a high-performance neutron detector can be researched and developed, and the technical development of the neutron detector is promoted.

This invention relates to a method for separation and purification using a combination of gel resin and cationic resin. 68 The method described in this invention, belonging to the field of nuclear technology applications, solves the problem that existing separation and purification methods cannot achieve one-step separation and purification, resulting in low separation and purification efficiency. The separation and purification method of this invention is as follows: Step (a): Dissolve and leach the gallium target after accelerator irradiation to obtain a feed solution; Step (b): Adjust the pH of the feed solution to alkaline; Step (c): Pass the alkaline feed solution through a gel resin column; Step (d): Elute the gel resin column; Step (e): Remove the contents containing... 68 The eluent of Ge was passed through a cation exchange resin column and washed to obtain... 68 Ge eluent. This invention achieves the [missing information - likely related to elution]. 68 One-step separation and purification of Ge.

The invention provides a nuclide activity measurement method based on a neural network, and relates to the technical field of nuclide activity measurement, and the method comprises the steps: constructing a coincidence measurement system, obtaining a plurality of groups of counting parameters of nuclides under different beta detection efficiencies through the system, and carrying out the correction processing of the counting parameters, and inputting the corrected counting parameter into a one-dimensional convolutional neural network model, and obtaining the nuclide real activity based on a condition that model output extrapolation is performed until beta detection efficiency is 100%. According to the method, the neural network is adopted to replace a traditional extrapolation fitting method, the nonlinear relation between the nuclide counting rate and the activity is automatically learned, the accuracy and robustness of activity measurement are remarkably improved, the method does not depend on an explicit function form, abnormal data can be effectively handled, the accuracy of beta, gamma and coincidence counting rates is ensured, and the method is suitable for popularization and application. The interference of environmental background and cosmic rays is effectively inhibited, the signal-to-noise ratio is increased, the measurement efficiency is improved, the operation is simpler and more convenient, and the method is suitable for various nuclear technology application scenes.

The application belongs to the technical field of lunar soil measurement, and particularly relates to an analysis method for non-destructive measurement of stable metal isotope ratio in lunar soil, which comprises the following steps: step S1, placing a sample (2) to be measured into a reactor (1) to perform irradiation operation, wherein the sample (2) comprises a lunar soil sample and an earth sample, and the earth sample comprises standard substance and neutron fluence rate parameter monitoring material; step S2, after the irradiation operation is completed, cooling is performed, and then a HPGe detector (4) is combined with a digital multi-channel spectrometer (7) to perform radioisotope activity measurement on the activated lunar soil sample; step S3, after the radioisotope activity measurement is completed, the ratio of stable isotope A and stable isotope B in the lunar soil sample is calculated; and step S4, the difference between the isotope ratio of the lunar soil sample and the isotope ratio of the earth sample is calculated. The non-destructive isotope ratio measurement method provided by the application is complementary to the mass spectrometry method such as ICPMS, and more nuclear technology scientific data is provided for non-destructive analysis of lunar soil.

The invention relates to the field of nuclear technology, in particular to a complex system multi-layer progressive monitoring method based on fuzzy C-means clustering and a Gaussian mixture model, which comprises the following steps of: acquiring multi-working condition data of a complex system by adopting an integrated simulator, and calculating parameter dynamic correlation by combining a Spearman correlation coefficient method; calculating the dynamic correlation of the parameters, and calculating the Euclidean distance between the real-time parameter sequence and the first clustering center based on a fuzzy C mean value; carrying out differential processing on the data, combining a static distance and a dynamic distance at the same moment, taking a dynamic characteristic distance as a monitoring reference, calculating a Mahalanobis distance from each data point to Gaussian distribution, training a model by an initial section, establishing a monitoring task piece through cyclic operation, and judging whether true abnormity exists or not during abnormity monitoring. According to the method, the problems that a traditional complex system monitoring method is difficult to accurately capture multi-working-condition dynamic characteristics, incapable of effectively processing non-stationary time-varying data and low in anomaly judgment accuracy can be solved.

The invention relates to a double-loop anti-radiation phase-locked loop circuit and a control method. The double-loop anti-radiation phase-locked loop circuit comprises a digital regulation branch, an analog regulation branch, a multiband voltage-controlled oscillator and a frequency divider, the digital regulation branch, the multiband voltage-controlled oscillator and the frequency divider jointly form a coarse tuning loop, and the digital regulation branch is used for generating a frequency band control word according to a frequency difference between a reference clock and an output clock of the frequency divider to coarsely tune the frequency of the multiband voltage-controlled oscillator until the coarse tuning loop is locked; the analog regulation branch, the multiband voltage-controlled oscillator and the frequency divider jointly form a fine regulation loop, and the analog regulation branch is used for generating a control voltage signal according to the phase difference between the reference clock and the output clock of the frequency divider after the coarse regulation loop is locked, and performing fine regulation on the gain of the multiband voltage-controlled oscillator. The frequency range of the multi-band voltage-controlled oscillator can be expanded, the gain can be reduced, and radiation interference can be reduced. The method can be widely applied to the fields of microelectronic chip design and nuclear technology.

The invention relates to the technical field of nuclear technology application. The invention discloses a method and system for testing the ray shielding performance of radiation protection clothing in a real scene, and the method comprises the steps: vertically placing the radiation protection clothing, and placing the radiation protection clothing in a radiation field; measuring external radiation dose data and internal radiation dose data of the to-be-measured part of the radiation protection suit through a radiation dose measuring device; and according to the external radiation dose data and the internal radiation dose data of the to-be-detected part of the radiation protection suit, calculating to obtain a ray shielding rate and a linear attenuation coefficient of the radiation protection suit. According to the invention, the radiation protection suit is vertically placed, and the external radiation dose data and the internal radiation dose data of the to-be-tested part of the radiation protection suit are measured by the radiation dose measuring device, so that the ray shielding rate and the linear attenuation coefficient of the radiation protection suit are obtained. The gamma-ray shielding performance test of the protective clothing in a real wearing scene can be simulated, and the test accuracy is improved.

The invention provides an extraction method and an extraction device for protactinium elements in thorium-based proliferative salt, and relates to the field of nuclear technology, and the extraction method for the protactinium elements in the thorium-based proliferative salt comprises the following steps: mixing thorium-containing alloy melt with thorium-based proliferative salt melt to reduce protactinium ions in the thorium-based proliferative salt melt into metal protactinium, dissolving metal protactinium into the alloy melt to obtain a protactinium-containing alloy melt; standing the molten liquid after the reduction reaction, layering the protactinium-containing alloy molten liquid and the thorium-based proliferation salt molten liquid after the reduction reaction, and taking out the layered protactinium-containing alloy molten liquid; and distilling the protactinium-containing alloy melt to separate out the protactinium element.

This invention discloses a thorium polymer spinning precursor, thorium oxide nanofibers, their preparation methods, and applications, belonging to the field of nuclear materials technology. The invention uses thorium chloride as the thorium source, acetylacetone as the ligand, and triethylamine as the dechlorination agent to prepare a thorium polymer spinning precursor. A spinning solution is prepared using the thorium polymer as the spinning precursor, methanol as the solvent, and polyethylene oxide (PEO) as the spinning aid. The thorium polymer spinning precursor fibers are obtained by electrospinning, and the precursor fibers are then heat-treated to obtain thorium oxide nanofibers. The thorium polymer spinning precursor prepared by this invention has a high thorium content and stable properties. Its spinning solution exhibits good filamentation properties, and the prepared thorium oxide fibers have uniform diameter, dense structure, and excellent flexibility. This thorium oxide nanofiber material can be used as a neutron shielding material, a breeder fuel for nuclear technology, and also as a high-temperature resistant material for preparing special ceramics.

The invention relates to a reactor core refueling method and device, a storage medium and electronic equipment, and relates to the technical field of nuclear, and the method comprises the following steps: firstly, obtaining a reactor core picture book and a refueling program text of a reactor; the refueling program text is subjected to refueling text analysis, a named entity in the refueling program text is obtained, and the named entity comprises a refueling operation object; according to the named entity and a refueling rule base, generating a refueling operation instruction corresponding to the refueling program text; and finally, based on the reactor core picture book, executing a refueling simulation process of the refueling operation object according to the refueling operation instruction, and obtaining a reactor core loading picture after the refueling operation instruction is executed. According to the scheme, on the basis of the reactor core picture book of the reactor, the refueling simulation process of the refueling operation object in the named entity is overlaid according to the refueling operation instruction, manual participation in the refueling simulation process is not needed, the time of single-time refueling simulation operation is shortened, and the refueling simulation efficiency is improved.