48 results about "Nuclear material" patented technology

The invention relates to the technical field of fusion reactor nuclear materials, in particular to a multi-temperature-zone variable-atmosphere material compatibility experimental device and experimental method.The multi-temperature-zone variable-atmosphere material compatibility experimental device comprises an experimental cavity used for containing a to-be-tested material sample, a multi-temperature-zone heating control unit is connected with the experimental cavity, and an atmosphere control unit is communicated with the air inlet end of the experimental cavity; the tail gas emission purification unit is communicated with the gas outlet end of the experiment cavity; through the designed multi-temperature-zone heating control unit and the experiment cavity, a compatibility experiment of multiple temperature points in a single experiment is achieved, accurate temperature control of different temperature zones is achieved through the heat preservation and insulation plug and independent heating control of each temperature zone, the control precision is improved, the experiment efficiency is remarkably improved, and the experiment cost is reduced.

According to the ventilation system of the nuclear material production line and the nuclear material production line workshop, air exhaust equipment and air supply equipment in the system are used for conveying fresh air to a sealed area outside a target area, and the target area is the sealed area with the highest risk level; the air supply equipment is further used for extracting air from the adjacent sealing area with the risk grade lower than that of the target area and conveying the air into the target area so as to form directional airflow flowing from the low-risk area to the high-risk area. And the exhaust equipment is used for exhausting air in each sealing area. The control device is electrically connected with the air supply equipment and the air exhaust equipment and used for adjusting the air supply amount and the air exhaust amount of all the sealing areas so that the negative pressure gradient can be maintained among the multiple sealing areas, and the negative pressure in the target area is higher than the negative pressure in the adjacent sealing areas. The ventilation system can effectively avoid radioactive aerosol leakage in a high-risk area.

The invention discloses a reactor nuclear material cross-scale simulation method coupling microstructure simulation and mechanical-thermal performance calculation. The method comprises the following steps: a, planning a most unit sampling point through a most unit method; b, simulating through a mesoscale phase field model to obtain force thermal parameters; c, the force and heat parameters are input into a finite element model to be simulated to obtain force and heat performance; d, the mechanical and thermal performance is input into the phase field model for simulation to obtain the structure morphology of the target nuclear material; e, obtaining a macroscopic scale-mesoscopic scale coupling model by circulating the step c and the step d; f, obtaining a strong coupling model through physical time axis control; and g, constructing a database of the relationship between the mechanical and thermal parameters and the mechanical and thermal performance through a data-driven neural network algorithm. According to the method, strong coupling is achieved through parameter transmission of the mesoscale phase field model and the macroscale finite element, the macroscale and the mesoscale are connected through the most element method, the parameter transmission precision can be adjusted according to convergence, and novel reactor material mechanical and thermal performance prediction can be achieved.

The application discloses a kind of nuclear material plate uranium uniformity detection devices, it includes the feeding roller way conveying device (1) sequentially arranged from left to right, support (2) and discharging roller way conveying device (3), support (2) is respectively provided with detector assembly (4) and calibration assembly (5) on front and rear ends, detector assembly (4) is set to the front side of feeding roller way conveying device (1), detector assembly (4) includes linear module (6), detector (7) and gamma ray radiation source (8), linear module (6) is set to the top surface of support (2), the slider (22) of linear module (6) is fixed with connecting frame (9), the rear end of connecting frame (9) is provided with L plate (10), and the horizontal plate of L plate (10) is provided with gamma ray radiation source (8).The beneficial effects of the present application are: compact structure, realize the automatic flow detection of the uranium surface density value of nuclear material plate, improve the detection efficiency, simple operation.

The invention belongs to the technical field of nuclear material measurement and analysis, and particularly relates to a neutron detection method for measuring the mass of a nuclear material and a radiation dose field of the nuclear material. According to the embodiment of the invention, the method comprises the steps: carrying out the energy response calculation and calibration of a detector through a parallel beam standard neutron source, and obtaining a calibration factor; performing energy response calculation and calibration on the detector by taking a to-be-measured object as a target object, determining the energy response and fluence-ambient dose equivalent conversion coefficient of the detector in a radiation field of the to-be-measured object, and realizing simultaneous measurement of a neutron counting rate, a fluence rate and an ambient dose equivalent rate; and the requirements of carrying out radiation dose field monitoring while carrying out quantitative analysis on the to-be-detected object in a special scene are met. The technical scheme of the invention can be popularized and applied to fixed-point monitoring of a to-be-detected system with known or measurable nuclear material source item composition and isotope abundance in nuclear facilities, or retention volume analysis and radiation dose measurement before transfer or treatment of some equipment for treating nuclear materials.

According to the invention, there is provided an overpack cask (1) for reversibly receiving, handling and transporting a repository container (100) for spent nuclear material (150), wherein the overpack cask (1) comprises a tubular cask shell (30) defining a receiving cavity (35) for reversibly receiving a repository container (100) therein, a first cask cover (10) providing a sealing closure of the receiving cavity (35) at a first axial end (31) of the tubular cask shell (30) and a second cask cover (20) providing a sealing closure of the receiving cavity (35) at a second axial end (32) of the tubular cask shell (30) opposite to the first axial end (31). At least the first cask cover (10) is removable from the tubular cask shell (30) to uncover a first opening (33) of the tubular cask shell (30) at the first axial end (31) for loading and / or unloading a repository container (100) into and from the receiving cavity (35), respectively. Furthermore, the overpack cask (1) comprises a sliding guide arrangement (40) arranged at an inner circumferential side (38) of the tubular cask shell (30) for guiding and supporting a repository container (100) during loading it into and unloading it from the receiving cavity (35), and when it is received in the receiving cavity (35).

1. The name of the design product: side detection passive nuclear material level meter (1). 2. The use of the design product: for not need radioactive source, use sensor in the crystal side detection container material level. 3. The design points of the design product: in shape. 4. The picture or photo that can best show the design points: perspective drawing.

The utility model relates to the field of level gauges, in particular to a level gauge for powder storage equipment, which comprises the storage equipment, a feeding pipe is fixedly mounted on the upper portion of the left side of the storage equipment, a first infrared sensor is arranged at the joint of the feeding pipe and the storage equipment, and a first electric valve is fixedly mounted on the left portion of the feeding pipe. A discharging pipe is fixedly installed in the middle of the lower end of the storage device, a second infrared sensor is arranged at the joint of the discharging pipe and the storage device, and a second electric valve is fixedly installed on the lower portion of the discharging pipe. When the storage device is used, the storage device is integrally sealed, the radiation quantity of the material stored in the storage device is integrally measured by the external level gage, and the radiation quantity in unit time is calculated, so that the volume of the consumed material is calculated, and the level height of the material is calculated; therefore, the measurement error of a traditional passive nucleon level gauge caused by flying dust generated when dust materials are loaded and unloaded in storage equipment is avoided.

The application discloses a method for improving high-temperature water vapor oxidation resistance of FeCrAl alloy based on pre-oxidation, and the method comprises the following steps: performing surface pretreatment on FeCrAl alloy material; placing the pretreated FeCrAl alloy material in a high-temperature and high-pressure water environment to perform pre-uniform corrosion treatment, so as to form a double-layer oxide film; and placing the FeCrAl alloy material with the double-layer oxide film in a high-temperature water vapor environment to test the oxidation resistance thereof. The pre-uniform corrosion oxide film formed under the normal working condition of a nuclear reactor improves the oxidation resistance of the FeCrAl alloy under the accident high-temperature condition, especially enables the low-Cr-content FeCrAl alloy (Cr 10-12%) to have high-temperature oxidation resistance equivalent to or even better than that of the high-Cr-content alloy, and meanwhile, the irradiation embrittlement problem caused by the high-Cr-content is avoided, thereby providing technical support for development of a new generation of nuclear materials and safe operation of a nuclear power station.

The utility model discloses a kind of material species correction measurement devices of passive nuclear material level meter, it is related to passive nuclear material level meter technical field, including material level meter body, the material level meter body one end is provided with adjusting mechanism, the adjusting mechanism includes lead screw sliding table assembly, support and bevel gear one, in the utility model, bevel gear two drives bevel gear one rotation, bevel gear one and crossbar synchronous rotation, the inclination angle of adjustable detector assembly, by motor one drive screw rotation, can drive connecting block along guide plate sliding, to change the use position of detector assembly, adjustment detector assembly received effective radiation amount, compensate the absorption difference of different material species, complete material species correction measurement, satisfy the requirement of precision and stability to industrial production, the installation bolt in limit block can limit installation ring, improve the stability when using built-in component, also convenient for installation ring and built-in component installation and disassembly in the installation of positioning ring bottom.

The invention discloses a double-loop water chemical multi-parameter coordinated regulation and control test device for nuclear power material corrosion simulation. The double-loop water chemical multi-parameter coordinated regulation and control test device comprises a low-pressure water chemical pretreatment loop, a high-temperature and high-pressure corrosion simulation loop and a control module, the low-pressure water chemical pretreatment loop is connected with the high-temperature and high-pressure corrosion simulation loop through a pipeline and a valve, so that the test medium regulated and controlled by the low-pressure water chemical pretreatment loop is conveyed to the high-temperature and high-pressure corrosion simulation loop and flows back to the low-pressure water chemical pretreatment loop; the control module is electrically connected with the first monitoring component, the second monitoring component, the dissolved hydrogen / oxygen control unit, the dosing unit and the autoclave heating unit, receives and analyzes the hydrochemical parameter deviation of the low-pressure hydrochemical pretreatment loop and the high-temperature and high-pressure corrosion simulation loop by adopting a PID (Proportion Integration Differentiation) regulation algorithm, and dynamically regulates the working state of the execution unit; therefore, multi-parameter coordinated regulation and control are realized.

The application discloses a method for calculating the upper limit of impurity content of nuclear materials of a fusion device, and is used for analyzing the impurity control requirement of materials serving in the nuclear environment of the fusion device. The method is characterized in that: under the service environment conditions and examination requirements of the specified nuclear materials, the main components and the possible impurity elements contained in the materials are simulated, and the control requirement of the impurity elements is determined according to the nuclear response calculation of various normalization indexes. The method can realize the rapid calculation of the upper limit of impurity content of the nuclear materials of the fusion device under the service conditions, and can be applied to the design analysis of the nuclear materials of the fusion device and the fusion device.

The invention relates to the technical field of nuclear material analysis, in particular to a plutonium solution specific activity determination method and application, and the plutonium solution specific activity determination method comprises the following steps: taking a plutonium solution with known specific activity A1 as a standard solution, determining the liquid flash counting rate of the standard solution, and recording the liquid flash counting rate as C1; obtaining a second-order conductance spectrum of the standard solution by using an ultraviolet-visible absorption spectrum method, determining the peak area of a characteristic absorption peak of tetravalent plutonium in the second-order conductance spectrum of the standard solution, and recording the peak area as S1; measuring the liquid flash counting rate of the to-be-measured solution, and recording the liquid flash counting rate as C2; obtaining a second-order conductance spectrum of the to-be-detected solution by using an ultraviolet-visible absorption spectrometry, determining the peak area of a characteristic absorption peak of tetravalent plutonium in the second-order conductance spectrum of the to-be-detected solution, and recording the peak area as S2; and calculating the specific activity A2 of the solution to be detected according to the following formula. According to the method, liquid scintillation counting and ultraviolet-visible spectrophotometry are combined, so that combination of activity response and concentration response is realized, and acquisition of isotope composition of a plutonium solution to be detected and complex sample pretreatment are not needed.

The present disclosure relates to nuclear chemical, particularly radiochemical, technologies at different stages of the nuclear fuel cycle, such as the production of purified nuclear materials (uranium, zirconium) or the reprocessing of spent nuclear fuel from nuclear power stations, in which extraction processes and operations for purifying nuclear materials are used. An example method, which includes the partial decomposition of nitric acid during continuous evaporation while a solution containing a reducing agent is fed into the bottom part of an evaporator having a circulating bottoms solution, consists in that the process is carried out such that the solution is kept in the bottom part of the evaporator for more than 2 hours under the addition of an aqueous solution of formaldehyde and formic acid (hereinafter “the mixture”) or a solution of formic acid after the process has been started using the mixture.

The application discloses a kind of algorithm for reducing the relative standard deviation of neutron multiplicity measurement, it is related to nuclear material measurement technical field, the method includes the following steps: S1, when neutron pulse reaches detection system, neutron pulse is pre-delayed;S2, after pre-delaying, neutron pulse enters true coincidence pulse window, the pulse number in window is recorded, and true multiple coincidence neutron pulse count R+A that is superimposed with accidental multiple coincidence pulse is obtained;S3, neutron pulse passes through long delay queue, enters accidental coincidence pulse window, and the pulse number in window is recorded, and accidental multiple coincidence neutron pulse count A is obtained;S4, to neutron pulse, the R+A and A value corresponding to each pulse are carried out multiplicity statistics in steps S1-S3, and the multichannel statistics diagram of R+A and A is obtained;S5, R+A and A are subtracted after being counted, and true multiple coincidence neutron pulse count R distribution is obtained.The method provided by the application can effectively reduce the volatility of neutron multiplicity measurement result, improve the measurement accuracy of equipment.

The invention belongs to the field of nuclear materials, and particularly relates to a nuclear fuel and a preparation method thereof and a reactor, the nuclear fuel comprises a nitride matrix and an MAX phase material distributed in the nitride matrix; the chemical formula of the MAX phase material is Mn + 1AXn, M is selected from at least one of Ti and Cr, A is selected from at least one of Si and Al, X is selected from C, and n is 1 or 2. According to the nuclear fuel disclosed by the invention, the MAX phase material is introduced into the nitride matrix, and a corrosion-resistant oxide layer is generated in situ by utilizing the MAX phase material to block water permeation, so that the nuclear fuel has the properties of high thermal conductivity, relatively strong water corrosion resistance, steam oxidation resistance and radiation resistance, and can be applied to a light water reactor; the application risk of the existing uranium nitride fuel in the light-water reactor is reduced, and the safety and economy of the reactor are improved.

The application relates to the field of non-destructive measurement and verification and authentication of nuclear material properties, and particularly relates to a fast neutron multiplicity measuring device with active and passive measuring functions. The fast neutron multiplicity measuring device with active and passive measuring functions comprises a fast neutron detector, a sample container, a neutron generator and a neutron source shielding cone. The fast neutron detector is in a ring array structure. The sample container is arranged in a central measuring area. The neutron generator is arranged at the end of the ring array. The neutron generator has an alpha timing function. The neutron source shielding cone is arranged between the neutron generator and the fast neutron detector. The fast neutron multiplicity measuring device can realize the functions of active and passive fast neutron multiplicity measurement without changing the existing components of the device.

The invention relates to a nuclear material warehouse abnormity early warning system and method, and the system comprises a laser radar personnel tracking system, a nuclear radiation detection network system, and an ultra wide band wireless positioning system, each system employs a layered architecture, and is composed of a sensing layer, a network layer, a processing layer, and an application layer, in the sensing layer, the laser radar personnel tracking system is provided with a plurality of laser radars, the nuclear radiation detection network system is provided with radiation detectors distributed in a place, and the ultra-wideband wireless positioning system is provided with an ultra-wideband tag and an ultra-wideband base station antenna which are arranged on a nuclear material container; the network layer is provided with equipment responsible for data transmission and communication of each system sensor; the processing layer is provided with a processor for carrying out processing, feature extraction and information fusion on data of each system sensor; and the application layer realizes anomaly detection, early warning and visual display. According to the invention, identification of illegal behaviors of personnel and early warning and positioning of illegal transfer of nuclear materials based on multi-sensing information fusion are realized, and the intelligent detection and identification capability of nuclear security is significantly improved.

The invention provides a quality measurement method, a model training method, electronic equipment, a storage medium and a program product. The method comprises: obtaining neutron counting data of a nuclear material to be measured within a first measurement duration, the first measurement duration being less than a second measurement duration, and the second measurement duration being a duration for performing neutron multiplex measurement on the nuclear material to be measured based on a preset mathematical model; based on the neutron counting data, a neural network model is used for conducting mass measurement on the to-be-measured nuclear material, a measurement result output by the neural network model is obtained, and the measurement result represents the mass of the to-be-measured nuclear material. By means of the scheme, the accuracy of nuclear material quality measurement can be improved under the condition that the measurement time of the to-be-measured nuclear material is shortened.

Described herein, are methods for providing a protective coating to a storage container for storing nuclear material, the method comprising depositing nickel particles on at least one surface of the substrate to produce the protective coating, wherein the nickel particles are deposited by cold spraying a composition comprising nickel particles and a carrier gas comprising nitrogen. In one aspect, the carrier gas consists essentially or consists only of nitrogen. The methods do not require pretreatment or modification of the nickel particles prior to cold spraying, which makes the methods described herein economically practical. The coatings produced by the methods described herein possess several advantageous properties including, but not limited to, high adhesion strength to the storage system and low porosity. The coatings produced by the methods described herein are effective against chemical attack such as, for example, CISCC.

The invention provides an electronic seal assembly, an electronic seal automatic sealing management system and an automatic sealing management method. The method comprises the following steps: setting an electronic seal to record traceability information and an on-off state of a nuclear material container; an on-off locking assembly is arranged to be connected to an installation position of the electronic seal, the electronic seal is connected in the state that the nuclear material container is locked, and the electronic seal is disconnected in the state that the nuclear material container is loosened, so that the storage information and the storage state of the nuclear material can be automatically supervised, the harm of the nuclear material to workers is reduced, and the working efficiency is improved. And the sealing state of the nuclear material container is monitored in real time, stability and reliability are achieved, manual stocktaking is not needed, and the management efficiency is improved.

The application relates to an anti-radiation high-entropy MAX phase ceramic and a preparation method thereof, and belongs to the technical field of nuclear reactor materials. a Zr b V c Nb d Ta e )2AlC; wherein a+b+c+d+e=1. The high-entropy MAX phase ceramic provided by the application has radiation resistance and high-temperature resistance, has a small neutron absorption cross section, can inhibit the growth of radiation defects and helium bubbles, and improves the safety of nuclear materials.

The present invention relates to a nuclear reactor loop capable of containing molten salt containing nuclear fuel in channels arranged substantially vertically and providing up and down flow paths. The loop can be used to build a modular reactor from removable individual molten salt nuclear loops, a portion of which ("channels") have been placed in a critical configuration in which the channels contain non-critical amounts of nuclear material, but together produce a critical region of a reactor core. The present invention further relates to a method of operating a modular nuclear reactor loop and nuclear reactor.

The invention relates to the technical field of alloy materials, and provides a FeCrV-series low-activation high-strength high-entropy alloy and a preparation method and application thereof. The general formula of the alloy is represented by FeCrVTaxWyTiz, wherein 0 lt; x is less than or equal to 0.2, 0lt; y < = 0.2, 0 < = z < = 0.4. The room-temperature mechanical property of the high-entropy alloy is emphatically optimized, the yield strength of the designed alloy at the room temperature is higher than 1370 MPa, and the plastic deformation exceeds 3.3%. Compared with an existing low-activation high-entropy alloy, the low-activation high-entropy alloy has the advantages of being excellent in comprehensive performance and low in cost. The material can be used as a nuclear material of a fusion reactor and a new generation fission reactor, solves the problems of high cost, insufficient mechanical properties and the like of a traditional low-activation nuclear material, and further promotes the application and development of nuclear energy in the fields of aviation, national defense and the like.

The invention discloses a radiation detection system based on neutron energy spectrum modulation, and relates to the field of neutron radiation measurement, and the system comprises a fast neutron induced fission assembly, a liquid-based neutron energy modulator and a neutron detector array. The fast neutron induced fission assembly is a 144keV-magnitude fast neutron source and is used for exciting a nuclear material to generate fission neutrons containing components and geometric information of the nuclear material; the liquid-based neutron energy modulator is composed of a plastic container, heavy water and boron-containing polyethylene small blocks, is located between the nuclear material and the detector array, and modulates fission neutron energy and paths through non-uniform equivalent density and stirring / standing circulation operation to realize information encryption; the neutron detector array comprises 5 * 5 plastic flash block units of 5 * 5cm < 2 >, and the neutron energy deposition distribution after modulation is recorded. In combination with a template comparison method, the consistency of the nuclear material can be accurately judged by comparing difference measurement values of reference data and monitoring data, and confidentiality and detection efficiency are both considered.

The present application relates to a kind of spent fuel insoluble residue processing and analysis method, comprising extracting residue, drying and weighing, strengthening dissolution, analyzing residue heavy uranium plutonium content and the like steps;Among them, through blank filter membrane contrast and eluent double baseline control, the interference influence of the filtration process in residue analysis is significantly reduced, so that nuclear material weighing result is more quantitative and reliable, conservative and reliable.Through in small sample amount, high radiation, complex multiphase matrix condition, using optimized mixed acid system proportioning and reaction time / temperature window, the full dissolution of trace target nuclide uranium / plutonium is realized, the purpose of quantitative analysis of target nuclide in high radioactivity residue is achieved, to ensure the accurate determination and safe and efficient processing of target nuclide in residue.The present application will soluble nuclide be transferred into liquid phase as far as possible, the source item intensity and volume of high-level solid residue are reduced to the greatest extent, the burden of final radioactive waste disposal is reduced, and it is suitable for insoluble residue component analysis in nuclear fuel reprocessing plant.

The application discloses a kind of nuclear material plate cladding thickness detection devices, it includes from left to right sequentially arranged feed roller conveyor (1), support (2) and discharge roller conveyor (3), the front and rear ends of support (2) are provided with calibration assembly (4) and detection assembly (5), calibration assembly (4) includes driving motor (6) and two left and right opposite vertical plate (7) are set, two vertical plate (7) are all fixed on the top surface of support (2) and are set in the front side of feed roller conveyor (1), driving motor (6) is vertically arranged and is set in the front side of vertical plate (7), driving motor (6) is installed on the output shaft of rotary disc (8), and a plurality of standard zirconium piece clamping mechanisms (9) are provided on the edge of rotary disc (8) and along its circumferential direction.The beneficial effects of the present application are: compact structure, improve zirconium piece thickness detection efficiency, reduce detection cost, simple operation.

The embodiment of the invention relates to the field of data processing methods particularly suitable for matrix or vector calculation of specific functions, in particular to a method for determining a subcritical limit value of a target nuclear fuel cycle process scene. According to the method for determining the subcritical limit value of the target nuclear fuel cycle process scene provided by the embodiment of the invention, the reference experiment is screened according to the target nuclear fuel cycle process scene, so that the selected reference experiment data meeting the predetermined requirements are obtained, and the quality of the reference experiment data for verification is improved; correcting kernel data adopted by theoretical calculation of the selected reference experiment, so that the theoretical calculation is closer to a real physical scene; according to the corrected nuclear data, the offset and uncertainty of the calculation tool are determined, the uncertainty level caused by the calculation tool is reduced, unnecessary conservative assumptions are reduced, under the same safety level, the limiting conditions of the nuclear fuel circulation process are reduced, and the processing capacity of nuclear materials in nuclear fuel circulation process equipment is improved.