751 results about "Nuclear fuel" patented technology

The invention relates to the technical field of deformation monitoring, in particular to a nuclear fuel cladding tube biaxial creep deformation monitoring system which comprises an acquisition monitoring module, a deformation recognition module, a sample aggregation module, a deformation training module and a risk early warning module. According to the method, the deformation abrupt change section is screened out on the basis of multiple parameters, the same-direction change trend of each physical parameter in the same section is further extracted, the collaborative change fragment is constructed, and the physical inducement of the cladding tube structure abrupt change can be accurately locked. On the basis, quantitative analysis is carried out on the time difference between response inflection point time and deformation rate peak time of different parameters, a difference value mode reflecting system response lag is formed, and the time dynamic sensing capacity of deformation behaviors is enhanced. According to the method, samples with different lag characteristics are extracted, a linkage relation point set among parameters is reconstructed, and a multi-dimensional response interval is constructed in combination with Euclidean distance, so that clustering classification of deformation density and change trend is realized.

The invention discloses a welding seam detection system and method for a lower end plug of a nuclear fuel rod, and relates to the technical field of nuclear power equipment manufacturing automation, the welding seam detection system comprises a data acquisition module and a data analysis module, and the data acquisition module comprises a camera, a movement mechanism and an endoscopic detection probe installed on the movement mechanism. A data analysis module is used for positioning a welding seam detection position on an environment image of a target to-be-detected lower end plug shot by a camera after a nuclear fuel assembly hovers, sending a motion control instruction to control a motion mechanism to bear an endoscopic detection probe to reach the welding seam detection position, and controlling the endoscopic detection probe to reach the welding seam detection position; the endoscope video is collected, and the small-size endoscopic detection probe goes deep into the weld joint detection position, so that the limitation of a tiny space is overcome, and the full-coverage collection of the weld joint image of the lower end plug to be detected is realized. Furthermore, the data analysis module is used for achieving welding seam anomaly detection based on endoscope videos, and the accuracy of anomaly detection results is improved.

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.

The invention provides natural graphite powder for a high-temperature gas cooled reactor nuclear fuel element and a preparation method of the natural graphite powder. The preparation method comprises the following steps: performing screening, crushing, shaping and grading treatment on graphite fine powder to obtain powder with the particle size of 25-35 microns; carrying out primary purification and secondary purification on the powder by using a mixed acid solution, heating the obtained chemically purified graphite powder in vacuum, carrying out high-temperature purification, introducing a mixed gas of argon and carbon tetrafluoride when the temperature reaches 1800 DEG C, and starting a circular reaction; then continuously carrying out gradient heating to 2000 DEG C, 2200 DEG C and 2400 DEG C, and respectively carrying out circular reaction at each temperature; then cooling to 300-400 DEG C, introducing mixed gas of oxygen and argon, and carrying out oxidation treatment; and finally, scattering and homogenizing to obtain a natural graphite powder finished product. The natural graphite powder provided by the invention realizes lower total impurity element content and total boron equivalent and higher specific surface area.

The present invention relates to a nuclear fuel assembly suitable for a small modular reactor (SMR), the nuclear fuel assembly comprising: a spacer grid (110) having 17x17 grid cells; a plurality of fuel rods (120) arranged in the grid cells of the spacer grid (110); guide tubes (130) arranged and fixed in the grid cells of the spacer grid (110); and an upper end fitting (140) and a lower end fitting (150) fixed to the upper end and the lower end of the guide tubes (130), respectively, wherein the guide tubes (130) are characterized in that eight guide tubes are arranged in the main axis directions (Ixx)(Iyy) with respect to the central cell of the spacer grid (110), four guide tubes are arranged in the diagonal direction (Ixy), and sixteen guide tubes are arranged in the non-diagonal directions (Ixxy)(Ixyy).

The invention provides a remote communication method, device and equipment for a mixing and clarifying tank stirring device, belongs to the technical field of nuclear fuel post-processing, and solves the problem of control signal fluctuation caused by electromagnetic interference, cable aging and environment abrupt change in a complex industrial scene. The method comprises the following steps: acquiring signal quality data; preprocessing the signal quality data to obtain preprocessed signal quality data; determining predicted signal attenuation data according to the preprocessed signal quality data; comparing and analyzing the predicted signal attenuation data with a preset threshold value to obtain a control instruction; and controlling the mixing and clarifying tank to communicate according to the control instruction. According to the scheme, the anti-interference performance and the prediction precision can be improved, and the communication failure rate and the energy consumption are reduced.

The invention provides a device and method for continuously measuring the diameter of a fuel rod, and belongs to the technical field of post-irradiation inspection of nuclear fuel circulation, and the method comprises the steps: enabling a support to move relative to a sliding rail assembly so as to adjust the detection point position of a diameter measuring instrument in the axial direction of the fuel rod, and enabling the diameter measuring instrument to move relative to the support so as to adjust the detection height of the diameter measuring instrument, the measurement of different axial positions of the fuel rod is realized; the chuck can clamp the fuel rod placed on the supporting piece, and the chuck drives the fuel rod to rotate so as to realize continuous measurement of different radial positions of the fuel rod, so that the diameter value and the ovality of a detection point are obtained, the detection continuity of the diameter and the ovality of the fuel rod is improved, and the detection efficiency of the diameter and the ovality of the fuel rod is improved; meanwhile, the driving device drives the chuck to rotate so as to drive the fuel rod to rotate, it can be ensured that the fuel rod rotates in place, the rotating accuracy of the fuel rod is improved, and the accuracy of data collection and the corresponding accuracy can be improved.

A covering for reinforcing a base layer of a nuclear fuel cladding is provided. The covering comprises a first layer configured to cover a first portion of the outer surface of the base layer of the nuclear fuel cladding, a second layer surrounding the first layer and the base layer and a third layer surrounding the second layer. The first layer comprises a fiber based material, the second layer comprises an interfacing material and the third layer comprises Chromium. A reinforced cladding for nuclear fuel and a method for producing a reinforced nuclear fuel cladding are also provided.

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 relates to the technical field of nuclear fuel accident simulation, and discloses a nuclear fuel behavior out-of-pile simulation experiment device under a reactive introduction accident, the nuclear fuel behavior out-of-pile simulation experiment device comprises a high-temperature steam chamber, a cladding, an insulating tube, at least one core block, an electrifying assembly and an induction heating assembly, the cladding is arranged in the high-temperature steam chamber, the insulating tube is arranged in the cladding, and the core block is arranged in the high-temperature steam chamber; at least one core block is arranged in the insulating tube, and an electrifying assembly and an induction heating assembly are further arranged in the high-temperature steam chamber; the power-on assembly is connected with the two ends of the core block, the core block can be powered on, the core block is heated, the conductivity of the core block can be changed, induction heating can be conducted on the core block through the induction heating assembly, short-time extremely-high heating power can be provided through superposition of the core block and the induction heating assembly, and rapid high-power density change under a reactive introduction accident is simulated. The technical problems of low heating power, non-self-heating and insufficient heat penetration depth in the traditional method are effectively solved.

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 utility model relates to the technical field of nuclear fuel transportation, in particular to a mechanism convenient to mount and dismount and a nuclear fuel transverse transportation fixing device. Comprising a storage unit which comprises a storage grillwork, a storage head bin and a connecting cylinder; the disassembly locking unit comprises a clamping assembly, a pushing assembly, a gear, a fixing shaft and a pawl. The containing unit comprises a supporting bottom plate and a containing plate. The nuclear fuel container loading and unloading device has the beneficial effects that an operator can more easily complete the loading and unloading process, so that the nuclear fuel container can be quickly positioned and locked or unlocked, the loading and unloading time is shortened, and the dependence on professional tools is reduced; through the design of the pressing unit, the pressing unit can transversely press and fix the nuclear fuel assembly in the transportation container, so that in the transportation process of the nuclear fuel assembly, the stability of the transportation container can be improved, and the safety of the nuclear fuel assembly in the transportation process is ensured.

The invention discloses a powder double-layer suction pipe matched with an upper-opening storage barrel and a method, and relates to the technical field of uranium conversion related equipment. The device is used in cooperation with a storage barrel with an upper opening. The device comprises a double-layer pipe and a material gathering plate, the double-layer pipe comprises an outer pipe and an inner pipe coaxially arranged in the center of an inner hole of the outer pipe, an annular column area is formed between the inner pipe and the outer pipe, and a plurality of feeding holes are formed in the lower end of the inner pipe; the plurality of material gathering plates are annularly and uniformly distributed and fixedly connected to the outer circular surface of the lower end of the outer pipe; a feeding hole is formed in the excircle surface of the lower end of the outer pipe. A powder suction method matched with a storage barrel with an upper opening is used for being matched with a negative pressure system to suck powder materials in the storage barrel with the upper opening, and by being matched with an original eccentric rotation suction method, powder material residues are avoided, and the suction efficiency is improved. The invention provides a powder material suction structure and a powder material suction method which are adaptive to an upper-opening storage barrel based on a technical route of conveying powder materials through a gas path, and the powder material suction structure and the powder material suction method are particularly suitable for efficient and residue-free transfer of UO2 powder in the nuclear fuel preparation process.

Uranium conversion material multi-round continuous feeding system and feeding method, which relate to the technical field of uranium conversion related equipment. The uranium conversion material multi-round continuous feeding system includes a plant, a storage tank, a transfer tank, a material merging assembly, and a material feeding assembly. The material merging assembly includes a transfer tank rotating device, a storage tank translation trolley, a storage tank outer cover disassembly and assembly device, a storage tank inner cover opening and closing device, a transfer tank upper cover disassembly and assembly device, and a lower pipeline docking device. The material feeding assembly includes a transfer tank translation trolley, an upper pipeline docking device, and a transfer tank inner cover opening and closing device. The uranium conversion material multi-round continuous feeding method is applied to the uranium conversion material multi-round continuous feeding system, and realizes the automatic feeding of UO2 powder material before the start of the uranium conversion process in the nuclear fuel preparation process. On the basis of the traditional feeding process, the present invention optimizes the feeding process, realizes the multi-round continuous feeding operation of UO2 powder material, and meets the needs of large-scale, fast-paced, and high-efficiency uranium conversion production.

The invention relates to the technical field of nuclear fuel, in particular to a nuclear fuel cladding tube biaxial creep process monitoring system, which comprises a temperature stress data capture module, a regional state judgment module, a creep rate adjustment module, a thermal control stress regulation and control module and an effect feedback correction module. According to the method, temperature stress data of key parts are collected and subjected to structural conversion, data standardization and usability are improved, space-time information is fused to construct a coupling data table, accurate recognition of thermal stress is supported, the temperature stress trend of a boundary area is jointly analyzed, abnormal fluctuation sections are recognized, and the creep distribution recognition accuracy is improved; the method comprises the steps of analyzing continuous response increments and mutation points, forming a fine feature point set, enhancing strain mutation capture capability, extracting a key interface adjustment value, analyzing response differences before and after adjustment, realizing dynamic monitoring and verification of a control unit, improving regulation and control targeting, identifying an unreached target area through feedback comparison, constructing a return difference compensation mechanism, and realizing dynamic monitoring and verification of the control unit. And the regulation efficiency and the closed-loop regulation and control capability are enhanced.

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 present invention provides a high-performance, high-temperature-resistant aluminum alloy and its preparation method. This heat-resistant, high-strength aluminum alloy is based on an Al-Mg-Mn-Cu-Sr alloy with appropriate microalloying elements. It exhibits excellent mechanical properties at room temperature, 250°C, and 300°C, and the resulting aluminum alloy products are free of pores and microcracks. This alloy can meet the requirements for high-temperature-resistant aluminum alloys in a variety of fields, including aerospace, transportation, oil exploration, high-voltage power transmission, and nuclear fuel storage.

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

The 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 invention provides a zirconium alloy surface nanocrystallization and self low-temperature diffusion bonding method, relates to the technical field of nuclear fuel element manufacturing, and solves the technical problems that existing zirconium low-temperature diffusion bonding can only realize connection of a certain small-size sample and a sample with a regular and flat surface and cannot be applied in an engineering manner. The method comprises the following steps that the surface roughness of the to-be-welded surface of the zirconium alloy is machined to be smaller than or equal to 1.6 micrometers; performing surface nanocrystallization treatment on the to-be-connected surface of the processed to-be-welded zirconium alloy test piece by adopting ultrasonic peening; the to-be-welded zirconium alloy test piece subjected to nanocrystallization treatment is accurately assembled and fixed through spot welding; and the assembled zirconium alloy test piece to be welded is subjected to diffusion bonding through a vacuum diffusion bonding technology. The machining difficulty of the to-be-welded test piece is reduced, and batch manufacturing can be achieved in the field of nuclear fuel element manufacturing; the method is suitable for processing complex structural members; and a product with high joint connection strength, uniform interface structure components and good corrosion resistance is obtained.

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 present invention discloses a biaxial fatigue performance test system for axially combined internal pressure loading of nuclear fuel cladding tubes, which includes a medium tank for storing test media; a high-temperature test furnace in which a clamping fixture and a measurement module are installed; the clamping fixture is used to clamp and fix the cladding tube to be tested, and the measurement module is used to measure the change value of the outer diameter of the cladding tube; a liquid pumping module connected to the medium tank and the internal pressurization module for pumping the test media in the medium tank to the internal pressurization module; an internal pressurization module connected to the cladding tube to be tested; and a pressurization power module connected to the internal pressurization module and the external tension-compression module. The present invention solves the simulation conditions of the synchronous test of the internal pressure fatigue and the external tension-compression fatigue of the cladding tube, obtains the real-time research data of the synchronous test, fully realizes the fatigue strength test of the cladding tube of nuclear power materials, and further can optimize the design, production and other links of the cladding tube to improve the safe usability of nuclear power applications.