556 results about "Nuclear fuel" patented technology

The invention provides a nuclear fuel grillwork clamping device and a vibration acceleration control method, and relates to the technical field of nuclear fuel grillwork clamping devices.The nuclear fuel grillwork clamping device comprises the steps that a plurality of clamping pieces are slidably installed in a nuclear reaction box body in the first direction and connected with the multiple clamping pieces at the same time through a transmission mechanism, and the clamping pieces are connected with the nuclear reaction box body through the structural characteristics of the transmission mechanism; and the plurality of clamping pieces are enabled to vibrate in two opposite directions of the first direction. Based on the structural design, the mode that the clamping pieces are installed in the nuclear reaction box body is changed into sliding connection from fixed connection, and the multiple clamping pieces are connected at the same time through the transmission mechanism. In the nuclear reaction process, the transmission effect of the transmission mechanism can be utilized to enable the plurality of nuclear fuel grillages to vibrate along opposite directions; therefore, the vibration amplitude of the nuclear fuel grillwork is reduced, and finally the purpose of reducing the failure rate of the nuclear fuel grillwork is achieved.

A system is described that includes a sputter target and a magnetic element array including multiple sets of magnets arranged to have a Hall-Effect region that extends along a length of the sputter target. The elongated sputtering electrode material tube is interposed between the magnetic array and an object to be deposited with a sputtered material from the sputter target. During a direct current high-power impulse magnetron sputtering operation, the system performs a depositing on a surface of the object by generating and controlling an ion and neutral particle flux by: providing a vacuum apparatus containing a sputter target holder electrode; first generating a high-power pulsed plasma magnetron discharge with a high-current negative direct current (DC) pulse to the sputter a target holder electrode; and second generating a configurable positive voltage kick pulse to the sputter target holder electrode after terminating the negative DC pulse.

The invention relates to a nuclear fuel cladding produced with a substrate (14) which is made of pure zirconium or of a zirconium based alloy and a multilayer protective coating (16) which covers a surface (14B) of the substrate (14), the protective coating (16) comprising a main layer (18) made of pure chromium and one or more additional layers (20), each additional layer (20) being made of pure chromium or from a material made of chromium and, additionally, oxygen and / or nitrogen, with the possible presence of unavoidable impurities.

A nuclear fuel cell cladding that includes a zirconium-carbon covetic material. The zirconium-carbon covetic material has a carbon component associated with the surface of zirconium particles. The amount of carbon present in the zirconium-carbon covetic material is in a range of greater than 0.1 wt % to about 25 wt % of the zirconium-carbon covetic material. The carbon component may include carbon nanotubes, carbon nanomaterials, graphene, or graphene nanoplatelets. The carbon component may be uniformly distributed within the zirconium matrix. The zirconium-carbon covetic material may be formed from a zirconium alloy. The zirconium-carbon covetic material may be configured for use in various types of nuclear reactors. The synthesis of the cladding involves a process of plasma-enhanced chemical vapor deposition. The resulting nuclear fuel cell cladding offers improved performance and reliability for nuclear reactor applications.

The invention belongs to the technical field of nuclear fuel element surface protection, and particularly relates to a multi-layer composite (Cr, Al) Nx / Cr protective coating for a nuclear reactor zirconium alloy cladding and a preparation method of the multi-layer composite (Cr, Al) Nx / Cr protective coating. The coating adopts a multi-layer structure design consisting of a buffer layer, a middle layer and an outermost layer in sequence from inside to outside, wherein the buffer layer is a Cr coating; the middle layer is a nitrogen-saturated (Cr, Al) Nx coating, x is larger than or equal to 1, the middle layer is of a columnar crystal structure, and the hardness of the middle layer is 5-10 GPa; the outermost layer is a non-stoichiometric ratio (Cr, Al) Nx coating, x is larger than or equal to 0.6 and smaller than or equal to 0.9, a nanocrystalline structure is shown, and the hardness is 12-25 GPa. Component regulation, hardness gradient and a microstructure are subjected to collaborative design, so that the coating has excellent wear resistance, high oxidation resistance and structural integrity at the same time under a harsh nuclear working condition, the safety margin of the zirconium alloy cladding under normal service and accident working conditions is greatly improved, and the service life of the zirconium alloy cladding under normal service and accident working conditions is greatly prolonged.

The invention provides a nuclear fuel assembly grabbing positioning and deviation correction method and system based on industrial vision. The method comprises the steps of obtaining multi-view image data of a target nuclear fuel assembly and structured light deformation image data under a corresponding view; extracting an outer contour boundary of the target nuclear fuel assembly in a three-dimensional space based on the multi-view image data and the structured light deformation image data to obtain initial outer contour boundary data; performing pose positioning based on the initial outer contour boundary data in combination with original design data of the target nuclear fuel assembly to obtain a current pose parameter of the target nuclear fuel assembly in a three-dimensional space; performing deviation analysis based on the current pose parameter in combination with a preset standard pose parameter to obtain current deviation data and a corresponding deviation analysis result; and if the deviation analysis result is correctable deviation, performing instruction generation based on the current deviation data to obtain a pose correction instruction. The problem that the correction effect is poor in an existing method is solved.

The application discloses a preparation method of a zirconium alloy rod for a nuclear fuel assembly end plug, and comprises the following steps: first, a zirconium alloy forging rod is subjected to machining and lubrication, heated and kept warm, and then extruded, and after extrusion, the extruded rod blank is obtained by sequentially performing straightening and polishing treatment; second, the extruded rod blank is heated and kept warm again, and then subjected to multi-pass hot rolling, and after hot rolling, the hot-rolled rod blank is obtained by centerless peeling; third, the hot-rolled rod blank is subjected to multi-pass pilger two-roll cold rolling to obtain a cold-rolled rod blank; fourth, the cold-rolled rod blank is subjected to vacuum annealing treatment to obtain an annealed rod blank; and finally, the annealed rod blank is subjected to finishing treatment, which sequentially comprises straightening, centerless grinding and degreasing, so that the target product is obtained; and the key parameters such as temperature, holding time and deformation amount of each step are strictly controlled. The preparation method breaks through the key technology of long rod cold rolling, realizes large plastic deformation of the zirconium alloy rod, significantly reduces deformation passes and intermediate heat treatment, and thus breaks through the problems of non-uniform structure, low production efficiency and insufficient product precision and the like existing in the traditional process.

The embodiment of the invention provides a fault detection device and method for a fuel loading and unloading system of a high-temperature gas cooled reactor, relates to the technical field of nuclear fuel transportation systems, and aims to solve the technical problem that whether a fuel ball blocking fault exists in a pipe section or not cannot be accurately reflected due to the fact that an existing detection method is limited by detection precision and installation positions. The system comprises a detection assembly and a data processing unit, wherein the detection assembly comprises a ray unit and an acquisition unit; the data processing unit comprises an image data processing module and an image data analysis module; wherein the image data processing module is used for receiving an electric signal sent by the acquisition unit and forming a visual digital image; and the image data analysis module is used for identifying and judging the visual digital image and outputting a judgment result. By arranging the detection assembly based on the X rays, the blocking condition in the to-be-detected pipe section can be detected, and the detection precision and accuracy can be further improved through the data processing unit.

The invention belongs to the technical field of nuclear fuel preparation, aims to solve the problems of non-uniform particle distribution and high breakage rate in traditional coated particle fuel preparation, and discloses a coated particle dispersion fuel pellet and a preparation method thereof.The method comprises the steps that firstly, silicon carbide slurry is prepared, a silicon carbide base band is prepared through tape casting equipment, and then the silicon carbide base band is prepared; after tape casting, the base band enters a drying area to be partially dried, TRISO particles are arranged on the surface of the base band at set intervals, the silicon carbide base band is freely curled and formed from one end, degreasing treatment and sintering densification are conducted, and the core block is prepared through the method. Through high-precision tape casting and three-dimensional curling technologies, uniform distribution of TRISO particles in CDF is achieved, and the contact and breakage rate of fuel in the preparation process is reduced.

The invention provides an electrochemical selective extraction process and device, and relates to the technical field of material extraction in the nuclear fuel circulation process, and the electrochemical selective extraction process comprises the following steps: respectively carrying out corrosion potential tests on an uncontaminated stainless steel sample and a stainless steel sample contaminated with nuclide; obtaining the corrosion potential a of the uncontaminated sample relative to the cathode and the corrosion potential b of the contaminated nuclide sample relative to the cathode; putting a to-be-treated metal workpiece of which the surface is adhered with a pollutant material into an anode basket of an electrochemical extraction device, and connecting the anode basket to an anode of a direct-current power supply or a potentiostat; mixing nitric acid, sodium nitrate, a corrosion inhibitor and a slow release agent to prepare an electrolyte; a cathode basket of the extraction device is connected with a negative electrode of a direct-current power supply or a potentiostat, an anode basket is nested in the cathode basket, and the anode basket and the cathode basket are jointly immersed in electrolyte for extraction. According to the method, pollutants adhering to the irregular metal surface are effectively removed and recycled, the extraction rate of contaminated metal nuclides reaches 99% or above, and the introduction amount of matrix elements is lower than 3500 ppm / g matrix.

The invention belongs to the field of manufacturing of nuclear fuel elements for nuclear power stations, and particularly relates to a method for adjusting the rod position of a nuclear fuel assembly. Comprising the following conditions: 1, the condition that an upper end plug of a fuel rod penetrates through the eleventh layer of grillwork of the fuel assembly; 2, the lower end plug of the fuel rod does not penetrate through the first layer of grillwork of the fuel assembly; 3, the lower end plug of the fuel rod is close to the end face of the lower tube seat and cannot be adjusted by a pull claw. The nuclear fuel assembly rod position adjusting tool and method have the beneficial effects that the nuclear fuel assembly rod position adjusting tool and method are applied to manufacturing of pressurized water reactor nuclear fuel assemblies. According to the method, the position of the fuel rod in the rod pulling process can be effectively adjusted, and the technical requirements of nuclear fuel assembly products are met. The adjusted rod position of the nuclear fuel assembly meets the expectation.

The embodiment of the invention relates to the field of processes specially used for manufacturing a reactor or parts thereof, in particular to an assembling and charging testing method of a reactor using bulk nuclear fuel and the reactor comprising the bulk nuclear fuel. According to the method, loading test is carried out on the reactor by using the test facility after assembly, so that the backup reactivity of the reactor core can be determined, safe assembly and safe criticality of the reactor are realized, and the safety of experimenters is ensured; by gradually adding the partitioned side reflecting layers outside the reactor core formed by the bulk nuclear fuel, the reactor gradually reaches the critical value, the operation risk is reduced, and meanwhile, the critical point can be accurately obtained; after the critical loading of the reactor core is determined, the partitioned side reflecting layers are replaced by the complete side reflecting layers with the same value, so that the influence of gaps existing between the partitioned side reflecting layers can be avoided, and the reaction efficiency of nuclear fuel is improved. The reactor assembled according to the method has high neutron economy and low critical risk.

The application discloses a device and method for simulating the quick strain failure behavior of a nuclear fuel cladding, which comprises a support structure composed of a rack moving base and a hoisting device, a load loading system composed of a fixing frame, a powerful electromagnet, a limiting device, a counterweight and a guide column, and a test and measurement system composed of a punch, a plunger, a guide sleeve, a piston cylinder, a cylinder body, a hydraulic oil chamber, a pressure sensor, a driving pipe mounting flange, a locking nut, a driving pipe, a cladding pipe and a data acquisition device; the application provides a test method matched with the device; the application can realize the simulation of the quick strain failure behavior of the nuclear fuel cladding under the plane strain stress condition in the millisecond scale, can accurately record the internal pressure and strain response relationship of the cladding under the quick strain condition, and thus realizes the evaluation of the quick strain failure performance of the cladding.

The utility model discloses a plate type nuclear fuel element tube base machining device, relates to the technical field of nuclear fuel element manufacturing, and is used for solving the research and machining problems of an existing plate type fuel element tube base. The device comprises a tailstock, an indexing head and an inner support of a pipe seat, wherein the tailstock and the indexing head are fixedly arranged on a machine tool workbench; the tailstock is provided with a center capable of freely stretching out and drawing back, and the center of the left end face of the inner support is tightly jacked through the center. A clamping jaw for clamping and fixing the right end of the inner support is arranged on the dividing head; the inner support can support and fix the pipe base with the inner hole as the reference, the left side of the inner support presses the check block through the nut to fix and press the pipe base, a workpiece has the rotary machining function, and after one face is machined, the workpiece is rotated to sequentially machine a four-face structure. The machining device is designed and manufactured according to the structural characteristics of the tube socket, and all the structures of the tube socket are machined through one-time clamping while the machining precision is guaranteed.

The nuclear fuel assembly comprises grids distributed along the length of nuclear fuel rods (4), the grids including spacer grids (14), at least one reinforcing grid (16), a lower end grid (18) and an upper end grid (20). The spacer grids (14) define, between them and together with the lower end grid (18) and the upper end grid (20), a plurality of intervals (15). The grids further include one or more intermediate mixing grids (22) and, optionally, an anti-debris grid (24) in which the lower ends of the nuclear fuel rods (24) engage.

The invention belongs to the field of nuclear fuel, and particularly discloses a metal material for fuel rod gap filling and a preparation method of the metal material. The metal material comprises the following components in parts by mass: 0.01 to 44.5 parts of lead, 0.01 to 56.5 parts of bismuth, 0.01 to 61.9 parts of tin, 0.01 to 5 parts of magnesium and 0.01 to 5 parts of zirconium; and the melting point of the metal material is less than 300 DEG C. The metal material disclosed by the invention has a lower melting point and a higher heat conductivity coefficient, is applied to the in-pile service process of the fuel rod of the advanced reactor, is molten at a loop temperature to be in a liquid state, fills a gap, immerses and seals a fuel core block section, and serves as an excellent heat-conducting medium, so that heat transfer between the fuel rod core block and a cladding is realized, and the heat transfer efficiency is improved. The core block and the cladding are combined to reduce the temperature of the nuclear fuel, meanwhile, the gasification phenomenon does not exist, the phenomenon of irradiation swelling-fission gas release does not occur, the contact force between the core block and the cladding is reduced, and the service life of the fuel rod is prolonged.

The invention belongs to the technical field of nuclear fuel assembly detection, and particularly relates to a fuel assembly operation robot positioning method and system based on adaptive contour feature matching. Comprising the steps that a global surface optimization tool converts acquired contour data into optimized ideal contour surface information, a fitting contour surface evaluation optimization tool feeds back a fitting contour surface overall deviation evaluation result, and a task data integration tool is an algorithm tool for relevant data configuration in the operation execution process of an execution assembly. And the task data integration tool constructs a data warehouse, and obtains trajectory data corresponding to the contour from the contour surface optimization module. The method has the advantages that one-key self-adaptive execution of assembly operation work can be achieved, accurate operation positioning is achieved through high-precision contour modeling, the method adapts to various motion control cards through a configuration type mode, dependence on the operation technology of operators is reduced, and the operation safety of equipment is improved.

The invention relates to the field of nuclear fuel element manufacturing, in particular to a molybdenum boat for nuclear fuel pellet boat pushing sintering. The molybdenum boat comprises a molybdenum boat frame, molybdenum pins and a molybdenum bottom plate. The molybdenum boat frame is integrally formed, the bottom face of the molybdenum boat frame is pressed into a corrugated shape, and the molybdenum boat frame is used for horizontally loading and stacking green pellets. The side wall of the molybdenum boat frame is partially provided with holes for atmosphere circulation in the hearth, and the molybdenum boat frame is fixed on the molybdenum bottom plate through molybdenum pins; the molybdenum pin comprises a stop part and a molybdenum pin part; a through hole is formed in the bottom surface of the backstop part, and the molybdenum pin part penetrates into the through hole and is fully welded with the backstop part; the four corners of the molybdenum bottom plate are respectively provided with a square groove, a positioning hole is formed in the center of each groove, and molybdenum pin parts of the molybdenum pins are inserted into the positioning holes. The device is simple in structure and not prone to deformation, the pins are firmly fixed, and the nuclear fuel pellets are horizontally and neatly loaded, so that the nuclear fuel pellets can be efficiently loaded into a furnace for sintering in a high-load mode.

The invention discloses a method and a system for predicting the breakage rate of a pressurized water reactor nuclear fuel rod, which are particularly suitable for a pressurized water reactor AFA3G nuclear fuel rod and are used for predicting the breakage rate of the pressurized water reactor AFA3G nuclear fuel rod by establishing a data-driven nuclear fuel rod fretting wear model and combining random uncertainty analysis. The method specifically comprises the following steps: firstly, constructing a flow-induced vibration-fretting wear coupling analysis finite element model of the nuclear fuel rod, and calculating the wear depth of the nuclear fuel rod along with time change; fitting a nuclear fuel rod wear depth curve according to the operation condition of the nuclear fuel rod by combining the geometric dimension and the material attribute of the nuclear fuel rod, and establishing a data-driven nuclear fuel rod wear mathematical model; further, establishing an uncertainty analysis model of the nuclear fuel rod, and calculating the failure probability of the nuclear fuel rod by using Monte Carlo simulation; according to the method, finite element analysis and uncertainty analysis are combined, so that the damage rate of the pressurized water reactor AFA3G nuclear fuel rod is predicted.

The application discloses an accident-tolerant nuclear fuel cladding coating and a preparation method and application thereof, and belongs to the technical field of nuclear fuel cladding coating. The coating is a Cr-based binary alloy coating deposited on the surface of a zirconium alloy base, and is selected from Cr-Fe, Cr-V or Cr-Mo alloy coatings, wherein the content of Fe is 3-10 wt%, the content of V is 3-20 wt%, and the content of Mo is 1-5 wt%. The preparation method adopts a magnetron sputtering process, and is realized by controlling parameters such as vacuum degree, temperature, gas pressure, bias voltage and power. Through alloying, the application effectively inhibits the element interdiffusion between the coating and the zirconium alloy base at high temperature, reduces the formation of brittle intermetallic compounds and Kirkendall holes, significantly improves the thermal stability and service life of the coating under accident conditions, meanwhile, the process is mature, the coating quality is high, and the coating is suitable for surface protection of zirconium alloy fuel cladding of nuclear power stations.

Nuclear fuel pellet incorporating thermally conductive inserts in the form of rods distributed at different azimuths and altitudes, fuel rod and associated nuclear fuel assembly. The invention consists of a nuclear fuel pellet (6), comprising: - a straight cylinder (60) of fissile material with a central axis (X) whose length and diameter respectively define the length (H) and diameter (Ø) of the pellet; - a plurality of inserts (8; 80, 81, 82) of thermally conductive material, the inserts being solid rods, distributed at different azimuth and altitude positions within the straight cylinder, and spaced apart from each other. Figure for the abstract: Fig. 4

The application relates to the technical field of nuclear fuel pellet temporary storage, in particular to a special tray for nuclear fuel pellets and a manufacturing process thereof, which comprises a loadable tray body and a reinforcing body arranged at the bottom of the loadable tray body to prevent the loadable tray body from being deformed, a plurality of positioning grooves extending along the Y-axis direction are arranged at the loadable tray body along the X-axis direction at intervals, the bottom of the positioning groove is in contact with a part of the reinforcing body, the reinforcing body has a protruding part protruding towards the positioning groove, the protruding part is staggered along the Y-axis direction and is provided with upper and lower protruding parts extending along the X-axis direction, the upper protruding part is provided with an airflow flowing position along the X-axis, the relative two outer sides of the loadable tray body are provided with reinforcing parts, the reinforcing body is provided with a fixing part around the protruding part, the fixing part is connected with the loadable tray body, and the reinforcing parts and the fixing parts are provided with positioning parts correspondingly. The application has the advantages of simple structure, stackability, high hardness, light weight, corrosion resistance, high temperature resistance, long service life of the product, small deformation coefficient, convenient carrying and storage.

The application relates to the technical field of nuclear fuel assembly manufacturing, and particularly discloses a nuclear fuel assembly casting type upper tube seat machining clamp and a machining method, wherein the tube seat blank is placed on a bottom plate, a positioning top block is pushed, four positioning pins below a positioning push rod are centered with positioning holes of the tube seat blank; then a positioning rod B is screwed to shorten the distance with a positioning rod A; then locking screws and fastening bolts on the edge frame of a positioning frame plate are locked to fix the tube seat blank on the bottom plate, positioning is realized; finally, the positioning rod B is completely screwed out of the positioning rod A, the bottom plate, the positioning frame plate and the tube seat blank are integrally taken out and fixed on a machining equipment. The machining clamp and the machining method satisfy the positioning clamping and machining requirements of the casting type upper tube seat, the upper tube seat after machining completely satisfies the drawing and technical condition requirements, and the machining scheme satisfies the engineering production.

The invention relates to the technical field of power generation devices, and provides a nuclear fuel built-in loop free piston thermo-acoustic generator which comprises a thermo-acoustic engine, a control drum, a neutron reflecting layer and a piston type generator, the thermo-acoustic engine comprises a low-temperature heat exchanger, a heat regenerator and a reactor core which are arranged in sequence, and the reactor core is provided with a channel for working medium gas to circulate; the control drum is arranged on the peripheral side of the thermoacoustic engine; the thermoacoustic engine and the control drum are both arranged in the neutron reflecting layer, the piston type generator is provided with a main piston air cylinder, the main piston air cylinder is in acoustic communication with the outlet end of the thermoacoustic engine through an expansion cavity side resonance tube, and the piston type generator is used for converting part of acoustic power into electric energy and adjusting the acoustic impedance phase angle. The reactor core is directly used as a heat source to be arranged in the thermoacoustic engine, so that the reactor core directly replaces a hot end heat exchanger in the traditional technology, the arrangement of a middle heat transfer structure is avoided, the structure is simple, and the problem of low heat transfer efficiency caused by the heat transfer structure is avoided.

The application relates to the technical field of fuel assembly transportation devices and discloses a nuclear fuel assembly transportation container hook locking device which comprises connecting pieces fixed on a fuel assembly transportation container and arranged adjacently on the same side, a limiting piece, and a lock hook piece pivotally connected to the connecting pieces and detachably connected with the limiting piece; the limiting piece is provided with a limiting groove and a locking mechanism, the locking mechanism comprises a lock tongue; the lock hook piece is provided with a hooking part which can be inserted into the limiting groove and hooked to a movable door assembly; when the lock tongue is subjected to a preset size of thrust of the hooking part, the lock tongue exits the limiting groove; when the hooking part is inserted into a set position of the limiting groove, the lock tongue extends into the limiting groove and stops the hooking part. The application has the limiting effect on the circumferential freedom of the movable door assembly, solves the constraint problem of the movable door assembly during fuel assembly loading and unloading, improves the stability of the fuel assembly transportation container, and has the advantages of simple structure, convenient operation and easy maintenance.

The invention discloses a method for catalytically reducing U (VI) under mild conditions, and belongs to the technical field of heavy metal pollution treatment and nuclear environment engineering. The spherical cobalt metal loaded nitrogen-doped carbon material (Co (at) MNCS) prepared by pyrolysis is used as a catalytic material, HCOOH is used as a reducing agent and an acidity regulator, strong acid does not need to be additionally added, and U (VI) can be efficiently reduced into U (IV) under the mild conditions of 25-60 DEG C and normal pressure by utilizing high catalytic activity and structural stability of the catalyst. Within the initial concentration range of 10-30 mg / L, the reduction efficiency exceeds 99%; almost complete reduction can be realized within 9 minutes at 25 DEG C; compared with a traditional hydrazine-containing and heavy metal-containing reducing agent or precious metal catalytic system, the method has the advantages that the environmental negative influence and the catalytic cost are remarkably reduced, the reduction efficiency of the catalyst still reaches 87% or above after 5 times of circulation, and the method conforms to the industrial development direction of nuclear fuel circulation greening and uranium-containing wastewater efficient treatment and has a wide practical application prospect.

The invention relates to a shock absorber of a nuclear fuel assembly transportation container, in which a shock absorption wood block is arranged at the end part of the nuclear fuel assembly transportation container and is used for shock absorption when the nuclear fuel assembly transportation container falls off, the shock absorption wood block is in a round cake shape, and the round cross section of the round cake shape is in contact connection with the nuclear fuel assembly transportation container; the petal-shaped structures are arranged in the circumferential direction of the round-cake-shaped cylindrical surface of the damping wood block, the multiple petal-shaped structures are arranged in the circumferential direction of the cylindrical surface of the damping wood block, the damping wood blocks are arranged at the two-thirds height positions of the petal-shaped structures, and the petal-shaped structures are filled with EVA foam. According to the shock absorber, the purposes that energy is absorbed when the shock absorber is subjected to low-speed impact through the petal-shaped structures protruding out of the shock absorption wood blocks, and energy is absorbed when the shock absorber is subjected to high-speed impact through the petal-shaped structures and EVA foam filled in the petal-shaped structures are achieved, and therefore the technical effect that the shock absorber gives consideration to energy absorption and shock absorption under the low-speed impact and the high-speed impact is achieved.

The invention discloses a nuclear rod power generation device, the nuclear rod power generation device comprises a nuclear fuel device, a neutron source device, a control assembly and a water pipe device, the nuclear fuel device is arranged in the water pipe device, and the water pipe device comprises a water inlet and a water outlet; the nuclear fuel device comprises an inner sleeve and an outer sleeve, the inner sleeve is arranged in the outer sleeve, the two ends of the inner sleeve and the two ends of the outer sleeve are connected in a sealed mode, a vacuum cavity is formed between the inner sleeve and the outer sleeve, and nuclear fuel coatings are arranged on the outer side of the inner sleeve and the inner side of the outer sleeve. The neutron source device is arranged on the inner side of the inner sleeve through a guide pipe, and the control assembly adjusts the position of the neutron source device in the guide pipe. The invention provides a nuclear rod power generation device which adopts a safe and miniaturized nuclear reaction technology and can replace a huge nuclear reactor.

The present application provides a kind of nuclear fuel conduit expansion joint device, comprising: the outer conduit and inner conduit are arranged between multiple expansion joint petals, the inner side wall of expansion joint petal has protruding part for abutting with outer conduit.The expansion joint ring is arranged outside multiple expansion joint petals.The inner diameter of expansion joint ring is larger near one end of connecting pipe, and the inner diameter of the other end is smaller.The thickness of expansion joint petal is smaller near one end of connecting pipe, and larger away from one end of connecting pipe.When transmission assembly drives expansion joint ring to move along the outside of multiple expansion joint petals away from connecting pipe, expansion joint ring abuts with the outer side wall of expansion joint petal, so that multiple expansion joint petals are close to each other away from the end of connecting pipe, and then the abutting force of protruding part and outer conduit increases, and plastic deformation occurs in outer conduit and inner conduit.Thereby, outer conduit and inner conduit are expanded and fixed.This way of applying pressure from the outside to produce plastic deformation, outer and inner conduit are not easy to crack, and the yield after expansion is higher.

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.