86 results about "Nuclear technology" patented technology

The invention provides a detector signal processing circuit and a detector signal processing method, which are applied to the field of nuclear technology application, and the detector signal processing circuit comprises a detector used for generating a detection signal; the pulse mode circuit is connected with the detector and is used for performing pulse shaping on the detection signal to obtain a pulse signal; the high-order mode circuit comprises a plurality of Campbell modules connected in parallel, the input ends of the Campbell modules are connected with the detector, and the Campbell modules are used for converting detection signals into first-order signals; the current mode circuit is connected with the detector; the current module is used for converting a detection signal into a current signal; the control module is connected with the pulse mode circuit, the current mode circuit and the output end of each Campbell module. According to the embodiment of the invention, the accuracy of a neutron flux detection result can be improved, and the measurement range of the detector is expanded.

The embodiment of the invention discloses a preparation method of a nuclear track membrane with orderly arranged nuclear micropores. The preparation method of the nuclear track membrane with orderly arranged nuclear micropores comprises the following steps: preparing a single-pore membrane; preparing a nuclear track membrane with two holes at a certain interval by utilizing a plurality of single-hole nuclear track membranes; a plurality of two-hole nuclear track membranes are used for preparing a four-hole nuclear track membrane which has a certain hole distance and is arranged in order, and so on, so that the required nuclear track membrane with nanoholes arranged in order is prepared. Nuclear micropores of a nuclear track membrane prepared by a traditional method are uncontrollable in number and disordered in distribution, and the nuclear track membrane plays an important role in multiple fields such as medicine, military industry and nuclear technology, such as production of semi-permeable membranes and wave-absorbing materials. No matter what membrane material is, the production quality of the nuclear track membrane can be improved by the method provided by the invention. A novel probe of a radiation detector can be manufactured by clamping a layer of nuclear track membrane made of an insulating material between two layers of nuclear track membranes made of conductive materials prepared by the method. Or conductive layers are plated on the two surfaces of a film made of an insulating material, and nuclear micropores which are arranged in order are punched on the film by using the method, so that the probe of the radiation detector of the same type can be manufactured. Nuclear track membranes with different pore diameters and pore taper angles are overlaid and arranged in a staggered manner, so that the range of the wavelength of electromagnetic waves absorbed by the wave-absorbing material is expanded.

The invention relates to a resin life prediction method and device, a storage medium and electronic equipment, and relates to the field of nuclear technology, and the method comprises the steps: firstly obtaining the exchange capability data of resin in a purification system and the purification flow of the purification system; measuring the inlet conductivity of the coolant flowing into the resin on line by using a first conductivity measuring instrument, wherein the first conductivity measuring instrument is arranged at the inlet of the coolant flowing into the resin; measuring the outlet conductivity of the coolant outflow resin on line by using a second conductivity measuring instrument, wherein the second conductivity measuring instrument is arranged at the outlet of the coolant outflow resin column; based on the inlet conductivity and the outlet conductivity, the equivalent concentration of impurity ions in the coolant is determined; and finally, predicting the service life of the resin according to the equivalent concentration of the impurity ions, the exchange capacity data of the resin and the purification flow of the purification system. Through the scheme disclosed by the invention, online prediction of the service life of the resin can be realized, and the time hysteresis of prediction of the service life of the resin is effectively reduced.

The present invention relates to the field of nuclear technology. A container for storing, transporting, and disposing of solid radioactive waste comprises a canister made of reaction-sintered silicon carbide with a free silicon content of 3% to 30% by weight, with a fumed silicon carbide layer deposited on the canister's surface. The canister's outer layer is made of metal foam with an open porosity of 60% to 70% and a pore size of 5 mm to 6 mm. The pores are filled with dispersed boron carbide powder of 40 to 50 μm in diameter, thereby protecting the environment from nuclear radiation emitted by high-level radioactive waste. A stainless steel cylinder with a thickness of 1 mm to 1.5 mm, intended to contain radioactive waste, is placed within the silicon carbide canister. A 5 mm gap between the inner surface of the silicon carbide canister and the stainless steel cylinder is filled with boron carbide powder, thereby protecting the environment from nuclear radiation emitted by high-level radioactive waste. The silicon carbide canister is sealed with a lid made of silicon carbide using a reaction welding method. The metal foam used as the metal foam with an open porosity is selected from the group consisting of titanium, aluminum, and copper. The present invention can increase the strength of containers for solid radioactive waste.

The present invention relates to a high-voltage power supply device for an electrostatic deflection plate, comprising a high-voltage generator and a high-voltage connection device. The high-voltage generator is used to convert and transmit incoming alternating current (AC) electricity through high-frequency inversion and soft-switching technology. The high-voltage connection device is connected to the generator at one end and to the electrostatic deflection plate at the other end, transmitting the stepped high-voltage electrical energy output by the generator to the plate while ensuring the voltage resistance of the connection. The present invention can be widely applied in the field of nuclear 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 embodiment of the present application provides a method for estimating the age of a spontaneous fission source, which relates to the field of nuclear technology and is used to solve the problem that the estimation process of the age of a spontaneous fission source is relatively difficult. The method for estimating the age of a spontaneous fission source provided in the embodiment of the present application includes: obtaining the radioactivity A of the sum of the first nuclides in the spontaneous fission source; 11 ; According to the radioactivity A of the sum of the first nuclides produced by spontaneous fission of at least one nuclide in the spontaneous fission source 21 and the first estimated age t1 of the spontaneous fission source, and let A 21 =A 11 , calculate t1; wherein the at least one nuclide includes the main nuclide of the spontaneous fission source. The embodiment of the present application provides a method for estimating the age of a spontaneous fission source for estimating the age of the spontaneous fission source.

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 present invention discloses a control rod drive mechanism for a small modular pressurized water reactor, which belongs to the field of nuclear technology and includes a protective shell with a radial dimension not exceeding 70 mm; a control rod drive body is installed inside the protective shell, and the control rod drive body includes a flange located at the left end of the protective shell for connecting with a reactor cover plate, a precision drive mechanism is arranged on the right side of the flange, a gear box is arranged on the right side of the precision drive mechanism, a position sensor unit is arranged on the right side of the gear box, and a sealed electrical input terminal is arranged on the right side of the position sensor unit. The present invention adopts the control rod drive mechanism of the small modular pressurized water reactor mentioned above, which has a compact size, a stroke of more than 1.5 meters, and an increment of less than 5 mm each time, ensuring precise control of the operation of the reactor core, and can stably operate under high temperature, high pressure, ionizing radiation and corrosive conditions.

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 invention discloses a neutron source cold package with a simplified structure, a reactor-based cold neutron source and a reactor, and belongs to the technical field of nuclear. A neutron source cold package with a simplified structure comprises a moderator input pipe, a moderator output pipe, a cold package cavity and a coolant circulating pipeline. The moderator input pipe comprises an inflow section parallel to the moderator output pipe and a bent backflow section, the cold package cavity is connected with the backflow section, and the maximum cross section area of the cold package cavity covers the direction interval of neutron guide pipe arrangement. The coolant circulating pipeline comprises an input interlayer, an output interlayer and a cold bag cavity interlayer, and a vacuum interval is formed between the cold bag cavity interlayer and the input interlayer.

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 discloses a small reactor control rod driving mechanism with a built-in optical position sensor, which belongs to the technical field of nuclear and comprises a protective shell, a control rod driving main body is mounted in the protective shell, and a rack is movably arranged in the middle of the control rod driving main body. The control rod drive body comprises a flange located at the left end of the protective shell and used for being connected with a reactor cover plate, and a precise drive mechanism is arranged on the right side of the flange. A sensor mounting groove is formed in the protective shell on the right side of the gearbox, an optical position sensor is arranged in the sensor mounting groove, and a reflecting plate matched with the optical position sensor is arranged at the end of the inner side of the rack; the right end of the protective housing is provided with a sealed electrical input end. According to the invention, the problems of precision, reliability, cost and complexity of the existing control rod driving mechanism are solved, and powerful support is provided for precise control and safe operation of a small modular pressurized water reactor.

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 field of nuclear technologies, in particular to an oxide-based uranium-plutonium nuclear fuel for an atomic power station with thermal neutron reactors. The technical result of the invention is the development of a REMIX-type fuel composition formulation making it possible, in the nuclear fuel cycle, to involve fissile nuclear materials regenerated from spent nuclear fuel with the simultaneous use of enriched natural uranium, and also moreover providing a solution to the problems of separately using previously isolated plutonium (including weapons-grade plutonium), of reducing the quantity of spent nuclear fuel by recycling fissile materials from spent nuclear fuel during the lifespan of a nuclear reactor (up to 5 recyclings), of economizing on natural uranium for manufacturing fuel and, as a result, of making the nuclear fuel cycle self-sustainable. The invention makes it possible to use REMIX fuel for solving an entire range of problems which a closed nuclear fuel cycle may face, taking into account the number of reactors, availability of natural uranium and the amount of stored spent nuclear fuel.

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 belongs to the technical field of nuclear power foreign matter prevention training examination, and particularly relates to a device for foreign matter digital museum training. Comprising a VR handle, VR glasses, a motion capture glove, a large display screen, a digital foreign matter museum display training platform, a workstation, a desktop computer and a server. The foreign matter digital museum training device has the beneficial effects that the educational training effect of foreign matter management in the nuclear power industry is effectively improved by establishing the foreign matter digital museum training device. The device integrates a digital display platform, a VR training platform and a training examination background management system, and realizes all-around foreign matter prevention education and practical operation training. Through a vivid three-dimensional scene and immersive experience, employees can intuitively feel the importance and the operation process of foreign matter management, and the safety awareness and the emergency processing capability are enhanced.

The embodiment of the present application provides a method for estimating the age of a spontaneous fission source, which relates to the field of nuclear technology and is used to improve the accuracy of estimating the age of a spontaneous fission source. The method for estimating the age of a spontaneous fission source provided in the embodiment of the present application includes: obtaining the radioactivity A of the sum of the first nuclides in the spontaneous fission source; 11 ; According to the sum of the radioactivity A of the first nuclide produced by spontaneous fission of multiple nuclides in the spontaneous fission source 21 and the age t of the spontaneous fission source, and let A 21 =A 11 , calculate t; wherein the multiple nuclides include one or more of the decay products of the main nuclides of the spontaneous fission source and the decay products of the impurity nuclides of the spontaneous fission source and the main nuclides. The method for estimating the age of a spontaneous fission source provided in an embodiment of the present application is used to estimate the age of a spontaneous fission source.

The present application relates to nuclear technology and nuclear medical technology field, and relates to a beam window system, comprising: a beam window and a beam scraping mechanism, comprising a beam window assembly, a beam scraping assembly, an air-filled bellows assembly and a shielding body, the beam window assembly, the beam scraping assembly and the air-filled bellows assembly are connected to form a module, and the module is connected with the shielding body to form an integral assembly; a guiding installation and support mechanism, comprising a three-dimensional adjustable support assembly, an installation positioning box, a guide sleeve, a hydraulic bellows assembly, a vacuum pipeline, a quick-release chain assembly, a front-end bellows assembly and a pipeline support assembly, the three-dimensional adjustable support assembly is arranged on the ground, the installation positioning box is arranged on the three-dimensional adjustable support assembly, the guide sleeve is connected with the installation positioning box, the installation positioning box is welded with the hydraulic bellows assembly along the downstream of the beam line, and the vacuum pipeline is welded with the front-end bellows assembly along the upstream of the beam line, the vacuum pipeline is connected and sealed with the front-end bellows assembly through the quick-release chain assembly, and the quick-release chain assembly is supported and arranged on the pipeline support assembly.

The invention relates to a high-voltage power supply device applied to an electrostatic deflection plate. The high-voltage power supply device comprises a high-voltage generating device and a high-voltage connecting device, the high-voltage generating device is used for performing electric energy conversion and transmission on accessed alternating current through a high-frequency inversion and soft switching technology; one end of the high-voltage connecting device is connected with the high-voltage generating device, the other end of the high-voltage connecting device is connected with the electrostatic deflection plate, and the high-voltage connecting device is used for outputting the stepped high-voltage electric energy output by the high-voltage generating device to the electrostatic deflection plate and ensuring the voltage resistance of the connecting part. The method can be widely applied to the nuclear technology field.

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 invention discloses an exit window neutron energy spectrum determination method based on energy spectrum superposition and a machine learning algorithm, and relates to the crossing field of nuclear technology and artificial intelligence. The method comprises the following steps: when target neutron energy spectrum data corresponding to a proton energy point on an incident direction point is not indexed, predicting target neutron energy spectrum data generated after targeting; multiplying the proton fluence ratio by the target neutron energy spectrum data corresponding to the protons, and adding the multiplication results of all the energy points and the protons corresponding to the incident direction points to obtain a target neutron energy spectrum; when the energy spectrum data corresponding to the neutron energy point on the direction point is not indexed, predicting the energy spectrum data; and substituting the corresponding energy spectrum data of the neutron energy point on the direction point indexed in the BSA database and the predicted energy spectrum data of the neutron energy point on the direction point into an exit window energy spectrum formula for calculation to obtain an exit window neutron energy spectrum. The method improves the determination precision of the neutron energy spectrum of the exit window.

The invention belongs to the field of nuclear technology application, and discloses a treatment method for improving the anti-fatigue performance of a metal material, the metal material and application, and the treatment method comprises the following steps: irradiating the metal material to be treated through neutrons generated by a reactor; the irradiation dose rate is 900 [mu] Sv / h to 1200 [mu] Sv / h, and the irradiation time is 7 days to 15 days. The metal material is irradiated by the neutrons generated by the reactor, and the microstructure of the metal material is optimized by regulating and controlling the irradiation dose rate and the irradiation time, so that the anti-fatigue performance is improved, and the service life of the metal material is prolonged.

The present invention relates to the field of nuclear technology and discloses a substrate structure, a target system, and a method for preparing the same. The target system comprises a substrate having a honeycomb-like upper surface. Several flow guide holes are provided at the junctions between adjacent through-holes in the honeycomb structure. Each through-hole is sequentially arranged, from bottom to top, within a hydrogen diffusion layer, a target material layer, and an oxidation mitigation layer. By optimizing the substrate structure of the target system, the present invention accelerates the diffusion rate of hydrogen released during nuclear reactions in the target material layer, thereby significantly improving the service life of the target system.

A thermodynamic process and system for material modification including software procedures and methodologies comprising specialized thermodynamics processes. The system applies thermal bonding and division through thermal contrasting in nuclear technology, medicine, computing, and military operations. The process uses heat to bond fibers while employing contrasting temperatures to maintain different thermal zones in confined areas for specific results. The system emphasizes absolute containment to prevent contamination and organizes disordered cells, atoms, and mechanisms. Both automated algorithms and operator technicians provide oversight to verify data within the system database using current and real-time information.

The application provides a kind of detector signal processing circuit and detector signal processing method, applied in nuclear technology application field, detector signal processing circuit, comprising: detector is used to generate detection signal;Pulse mode circuit is connected with detector, pulse mode circuit is used to pulse shaping to detection signal, obtains pulse signal;High-order mode circuit includes multiple parallelly connected Campbell modules, the input end of Campbell module is connected with detector, and Campbell module is used to convert detection signal into first-order signal;Current mode circuit is connected with detector, and current mode circuit is used to convert detection signal into current signal;Control module is connected with pulse mode circuit, current mode circuit, the output end of each Campbell module respectively.This embodiment can improve the accuracy of neutron flux detection result, and expand the measurement range of detector.

1. Name of the product in this design: Personal Dosimeter. 2. Purpose of this design: For monitoring the personal dose (neutron, X-ray and gamma rays) of radiation workers in nuclear technology applications, to ensure the life, health and safety of personnel. 3. The key design features of this product are the combination of shape and pattern. 4. The image or photograph that best illustrates the design's key points: a 3D model.