352results about "Reactor fuel elements" patented technology

The invention discloses an in-situ global foreign matter checking and grabbing system for reactor internals on the lower portion of a reactor. The in-situ global foreign matter checking and grabbing system for the reactor internals on the lower portion of the reactor comprises a base device, an adjusting device and a foreign matter grabbing device. The foreign matter grabbing device is arranged on the base device through the adjusting device. The foreign matter grabbing device comprises a flexible arm assembly, a flexible arm driving assembly, a quick change assembly, at least one clamping jaw used for grabbing foreign matter and a video tracking and checking assembly used for searching the foreign matter and monitoring the grabbing condition. The flexible arm assembly is arranged on the adjusting device; the at least one clamping jaw is detachably arranged on the flexible arm assembly through the quick-change assembly; the video tracking and checking assembly is arranged on the adjusting device; according to the invention, the flexible arm assembly carries various types of clamping jaws to carry out automatic grabbing operation, so that the in-service nuclear power plant reactor internals can accurately, efficiently and safely grab and clean foreign matters in the in-situ interior of a pressure vessel in a whole region.

The embodiment of the invention provides an anti-blocking device and an anti-blocking method for a fuel loading and unloading system of a high-temperature gas cooled reactor, and relates to the field of high-temperature gas cooled reactors. The anti-blocking device of the high-temperature gas cooled reactor fuel loading and unloading system comprises a driving module, a transmission module, an execution module and a control module, the transmission module is connected with the driving module, the execution module is connected with the transmission module, and the transmission module is driven by the driving module to drive the execution module to do linear reciprocating motion. And the control module is connected with the driving module and the execution module at the same time. A cooperative structure of the driving module, the transmission module and the execution module is arranged, a driving signal is converted into linear reciprocating motion of the execution module, mechanical disturbance is applied to a fuel ball stack in a blocked area, the lap joint balance or a bridge junction structure of the fuel ball stack is damaged, and active bridge breaking and jam releasing are achieved. And the jam releasing efficiency and success rate are obviously improved.

The invention belongs to the technical field of maintenance of upper reactor internals of a reactor, and particularly relates to a process and a device for maintaining a bolt locking cap of a control rod guide cylinder of an upper reactor internals. Comprising a bolt torque reinspection tool and a locking cap automatic bending tool. The bolt torque reinspection tool is matched with a torque wrench to realize reinspection of bolt torque; the automatic bending tool for the locking cap is used for controlling the bulging claw to correspond to a groove of a locking bolt head and driving the bulging claw to press downwards to bend the locking cap, so that looseness prevention of the bolt is achieved. The device has the beneficial effects that a bolt torque reinspection tool and an automatic locking cap bending tool are conveyed to a working area through a ring crane, and the field environment is not changed; the bolt torque reinspection tool and the locking cap automatic bending tool are both of a long-rod-shaped structure as a whole, an operator can perform remote operation on an operation platform, and whether the torque of the guide cylinder bolt meets the locking requirement or not can be inspected in cooperation with a torque spanner.

The invention discloses a reactor lower internals reactor core supporting plate global foreign matter video inspection grabbing device which comprises a rack, a video inspection device and a foreign matter grabbing device, a positioning pin is arranged at the bottom of the rack, and a containing groove used for storing foreign matter is further formed in the rack; the video inspection device is movably installed on the rack in the height direction and comprises a video inspection mechanism used for conducting foreign matter video inspection. The foreign matter grabbing device is movably installed on the rack in the height direction and comprises a foreign matter grabbing mechanism used for grabbing foreign matter. By applying the video inspection and grabbing equipment for the global foreign matters on the reactor core supporting plate of the reactor lower reactor internals, the foreign matters on the reactor core supporting plate can be inspected and grabbed when the reactor internals are in situ, the damage to the components and a pressure vessel body caused by hoisting of the reactor internals can be effectively avoided, the steps of hoisting of the reactor internals and the like are reduced, and the production efficiency is improved. And the overall foreign matter checking and grabbing operation efficiency can be effectively improved.

The embodiment of the invention provides a high-temperature gas cooled reactor fuel ball conveying pipeline cutting and jam releasing device, and relates to the technical field of high-temperature gas cooled reactors. The high-temperature gas cooled reactor fuel ball conveying pipeline cutting jam releasing device comprises a cutting function block, the cutting function block is provided with a first end and a second end which are opposite, a first mounting hole penetrating through the first end and the second end is formed in the cutting function block, and the first end is used for abutting against a blocked fuel ball; the cutting assembly is arranged in the first mounting hole and connected with the cutting function block, and the cutting assembly is used for cutting the fuel ball abutting against the first end; and the driving assembly is detachably arranged at the second end, and the driving assembly is used for driving the cutting function block to move. The problem that the blocked fuel balls in the conveying pipeline are inconvenient to release at present is solved, and continuous and stable operation of the high-temperature gas cooled reactor is guaranteed.

The embodiment of the invention provides a ball path dredging device for a high-temperature gas cooled reactor fuel loading and unloading system, and relates to the technical field of high-temperature gas cooled reactor fuel dredging. The ball path dredging device for the high-temperature gas cooled reactor fuel loading and unloading system comprises a box body, a coiling machine, a gas cylinder set, a tool head set, a filtering piece and a negative pressure piece, the box body is provided with a first installation cavity, a second installation cavity, a third installation cavity and a fourth installation cavity, the coiling machine is installed in the first installation cavity, and the gas cylinder set is installed in the second installation cavity; the tool head set is installed in the second installation cavity, the filtering piece is installed in the third installation cavity, and the negative pressure piece is installed in the fourth installation cavity. Key components such as a coiling machine, a gas cylinder group, a tool head group, a filtering piece and a negative pressure piece are intensively arranged in different mounting cavities in the same box body to form an integrated equipment framework. All the functional units are rapidly connected through preset gas paths, circuits and signal paths, and the operation process is remarkably simplified.

The embodiment of the invention provides an impact jam releasing device for a fuel ball conveying pipeline of a high-temperature gas cooled reactor, and relates to the field of high-temperature gas cooled reactors. The impact jam releasing device comprises an impact block, wherein the impact block is provided with a first end and a second end which are opposite in the axial direction; the moving part is connected with the first end of the impact block and used for driving the impact block to move along the ball path pipeline; the impact head is movably arranged in the impact block; and the impact driving piece is arranged in the impact block, is connected with the impact head and is used for driving the impact head to move in the axial direction of the impact block so as to drive the impact head to extend out of the second end or retract into the impact block. The impact jam releasing device provided by the invention can solve the problem that the blocked fuel balls in the ball path pipeline are inconvenient to dredge at present, and ensures the continuous and stable operation of the high-temperature gas cooled reactor.

The invention relates to an intelligent composite cladding tube with real-time monitoring and early warning regulation and control functions, which comprises a compact metal layer and a SiCf / SiC composite material layer which are used as an inner layer or an outer layer of the composite cladding tube, and a flexible transition layer positioned between the compact metal layer and the SiCf / SiC composite material layer, the sensing element is a silicon dioxide optical fiber or a molybdenum-rhenium alloy thermocouple and is used for monitoring the temperature change of the reactor core, so that the flexible transition layer has a temperature sensing function. According to the structure, the phenomena of local overheating and stress concentration can be effectively inhibited, so that the operation safety and the long-term service stability of a reactor system are remarkably improved.

The invention relates to the field of nuclear power operation, in particular to a control method for quantitative ammonia potassium saturation of cation resin of a primary loop purification system of a nuclear power plant. The method comprises the following steps: before starting ammonia-potassium saturation, calculating the total amount of ammonia absorbed by resin, the total amount of potassium absorbed by resin and the ideal concentration of a potassium hydroxide solution by utilizing the total exchange amount of nuclear-grade cation resins in a series I and a series II and the ammonia and potassium selectivity coefficients of the nuclear-grade cation resins; the flow of the first series and the flow of the second series are adjusted, the theoretical ammonia adding amount of the first loop under different flows of the second series is calculated, and continuous ammonia adding is conducted on the first loop with the theoretical ammonia adding amount as the target; when the series of diamino potassium is saturated, sampling and analyzing the total alkali metal content and the ammonia content from the outlets of the primary loop and the series II, and calculating and adjusting the adding amount and the frequency of the potassium hydroxide solution to ensure that the total alkali metal concentration of the main loop is in the area A and does not deviate; after the potassium ammonia saturation of the series II is finished, stopping the operation of the series I, calculating the required ammonia adding amount under the normal working condition, timely adjusting the continuous ammonia adding amount of the loop I, quitting the series I, and keeping the operation of the series II. According to the invention, smooth implementation of saturation of series of diamino potassium and safety of primary loop water quality are ensured.

The application discloses a nuclear power plant control rod guide tube, which comprises a protective cover and a full-range guide assembly installed in the protective cover; the full-range guide assembly comprises at least one full-range continuous guide element, and a full-range continuous guide channel is arranged in the full-range continuous guide element; the full-range continuous guide channel continuously guides at least one control rod in a single control rod assembly in a full range of the guide tube stroke. Compared with the prior art, the nuclear power plant control rod guide tube changes the non-continuous guide segment structure of the interval arranged guide panel into the full-range continuous guide segment formed by the long double-pipe structure, optimizes the control rod stroke protection function, reduces the risk of the control rod bending deformation caused by the transverse impact of the upper chamber fluid on the control rod, and guarantees the integrity of the control rod structure; meanwhile, the wear resistance of the specific position of the control rod guide tube is improved, so that the service life of the control rod guide tube is greatly prolonged, and the later operation cost is reduced.

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.

Systems and methods use inserted dividers with control rod drives (CRDs) to flush CRDs from inside a reactor. Dividers separate flush zones from the reactor. Pumps, injectors, gravity, or any other drive, potentially in combination with draining pumps, vacuums, or other suctioning device, flow the fluid through the divider into the zone and back out of the same for flushing out radiation and debris or bringing about any other desired condition. A handling pole may extend several meters to allow operators to reach any CRD and place the divider and flushing components about the same. Flow paths may extend with the pole to deliver and remove the flush fluid to divider ports. Additional cameras, locators, detectors, batteries, etc. may be used to inspect and clean. The systems may clean CRDs whenever accessible, such as after a control blade or rod has been removed during a maintenance outage.

A reactor containment vessel vent system continuously releases steam generated in a reactor containment vessel to the atmosphere even when a power supply is lost. In the reactor containment vessel vent system, the noble gas filter that allows steam to pass through but does not allow radioactive noble gases to pass through among vent gas discharged from the reactor containment vessel is provided at a most downstream portion of the vent line. An immediate upstream portion of the noble gas filter and the reactor containment vessel are connected to each other by the return pipe via the intermediate vessel. When the radioactive noble gases having pressure equal to or higher than predetermined pressure stay in the immediate upstream portion of the noble gas filter, the staying radioactive noble gases flows into the intermediate vessel by the relief valve. Thus, steam permeability is maintained and the system continuously releases the steam.

Described herein is a process for recovering uranium tetrafluoride (UF4) from hexafluoride uranium (UF6), the process comprising dissolving UF6 directly in an ionic liquid and recovering UF4, which can be processed to obtain UO2(s) or metallic uranium.

The present application relates to the technical fields of research reactor material irradiation, and specifically relates to a kind of with autonomous rotating function's pressure hole irradiation device, including remote terminal and sequentially connected driving section, protection tube and test section, driving section is set in the pressure hole mouth outside, including transmission shaft, power transmission component, servo motor and controller, transmission shaft is connected with protection tube, servo motor is installed in the side direction of transmission shaft, controller is electrically connected with servo motor, remote terminal is connected with controller, power transmission component includes transmission shaft gear and the motor gear meshed with transmission shaft gear, transmission shaft gear is coaxially keyed with transmission shaft, motor gear is coaxially keyed with servo motor output shaft;Transmission shaft and protection tube are provided with the gas outlet pipe and several gas inlet pipes communicated with test section.In the present application, the mode of remote control of servo motor is used, the irradiation device of the test piece can be realized autonomous rotation and test piece temperature control function, and the accuracy and reliability of test result are improved.

The invention discloses a fuel assembly assembling method and belongs to the field of nuclear power. The method comprises the following steps: immersing the zirconium alloy pipe fitting into a shellac resin coating agent dissolved by ethanol, then taking out and drying to form a protective paint film on the surface of the zirconium alloy pipe fitting; and after the fuel assembly is assembled by pulling the rod, the fuel assembly is immersed in a release agent to remove the protective paint film. The method can effectively reduce the risk of damage to the surface coating of the zirconium alloy pipe fitting in the assembling process, and improves the service life and reliability of the fuel assembly.

The present application relates to a kind of modular molten salt loading and storage system, including vertical support structure rack, normal temperature electrical equipment group and high temperature molten salt equipment group are physically separated by isolation fence and integrated in it;High temperature molten salt equipment group includes molten salt storage tank and molten salt pipeline connected with molten salt storage tank;Molten salt pipeline includes molten salt loading pipe, molten salt loading pipe has upward inclination relative to horizontal plane, to realize the self-weight backflow of molten salt;The middle section of molten salt loading pipe is connected with molten salt isolation valve, and molten salt isolation valve and normal temperature electrical equipment group are cooperated to realize the on-off control of molten salt loading pipe.The present application effectively avoids the thermal influence of high temperature environment on normal temperature electrical equipment by isolation fence, and the integrated design of overall modularization integrates the core equipment of molten salt loading and storage into one, discards the traditional dispersed equipment arrangement and installation mode, greatly simplifies the field construction process, reduces installation period, also makes the debugging, maintenance and overhaul of system more convenient.

The application discloses a kind of test methods for the operation of high temperature reactor fuel handling system choke, including outlet ball disturbance test and inlet ball disturbance test, by building test bench, build choke in test bench, gas is supplied to choke outlet pipeline, choke executes ball sending action, place multiple graphite ball elements at the inlet of choke, start choke and transport multiple graphite ball elements to downstream pipeline of choke, graphite ball element rolls during transport, graphite ball element rolls smoothly to downstream pipeline and emits sound, if each graphite ball element emits sound when rolling to downstream pipeline, then judge multiple graphite ball elements all roll smoothly to downstream pipeline.The application can verify whether choke can run smoothly, set multiple graphite ball elements to verify test, can improve verification structure accuracy, ensure that the choke verified can run continuously and stably, to guide choke factory acceptance work, so as to improve the availability of choke.

The reactor core fuel assembly transfer device comprises the reactor core fuel assembly transfer trolley, the reactor core fuel assembly transfer trolley comprises a trolley body, walking wheels and guide wheels, the walking wheels and the guide wheels are arranged on the trolley body, and the trolley body comprises a bottom plate, a rear side wall, a front side wall, a left side wall and a right side wall. The vehicle body is connected with a driving mechanism, and one end, far away from the driving mechanism, of the bottom plate is provided with a hole for a loader to turn over and pass through; the travelling wheels and the guide wheels are respectively mounted on the left side wall and the right side wall; the reactor core fuel assembly transfer trolley realizes cross-rail travelling along the transfer rail through the travelling wheels and the guide wheels; the left side wall and the right side wall of the vehicle body are concentrically provided with hinge holes used for being connected with a loader respectively. No power alternation, no clamping stagnation and no power failure risk exist in the whole working process, and the transfer efficiency and safety are improved.

The application discloses a reactor safety system, a shell member of which comprises a first shell and a second shell; a first injection pipeline is arranged in the first shell and the second shell and is communicated with a reactor core of a reactor; a first injection subsystem of a passive injection system comprises a first injection pool arranged outside the second shell, the first injection pool is higher than the reactor, and the first injection pool is communicated with the first injection pipeline; the first injection pool is configured to make the stored coolant in the first injection pool under the action of gravity to be injected into the reactor core through the first injection pipeline under a first working condition; a passive heat removal system comprises a first heat removal subsystem, a heat exchange mechanism of the first heat removal subsystem comprises an evaporator and a condenser communicated with the evaporator, the evaporator is arranged in the inside of the first shell, the condenser is arranged outside the second shell, and the evaporator is provided with a cooling medium. The first injection subsystem and the first heat removal subsystem are matched, so that the accident can be effectively controlled.

The manufacturing method comprises obtaining a strip of nickel-based alloy having a composition, in percentages by weight: trace amounts ≤ C ≤ 0.08%; trace amounts ≤ Mn ≤ 0.35%; trace amounts ≤ Si ≤ 0.35%; trace amounts ≤ P ≤ 0.015%; trace amounts ≤ S ≤ 0.015%; 17.0% ≤ Cr ≤ 21.0%; trace amounts ≤ Co ≤ 1.0%; 2.80% ≤ Mo ≤ 3.30%; 4.75% ≤ Nb + Ta ≤ 5.50%; 0.65% ≤ Ti ≤ 1.15%; 0.20% ≤ Al ≤ 0.80%; trace amounts ≤ Cu ≤ 0.30%; 50.0% ≤ Ni ≤ 55.0%; trace amounts ≤ B ≤ 0.006%; the balance being Fe and impurities resulting from the production, and performing desensitization annealing in a static furnace at an annealing temperature equal to or higher than 900°C and / or equal to or lower than 1050°C for an annealing time equal to or greater than 1 min and / or equal to or less than 180 min.

The manufacturing method comprises obtaining a nickel-based alloy strip having the following composition in percentage by weight: trace < = C < = 0.08%; trace < = Mn < = 0.35%; trace < = Si < = 0.35%; trace amount < = P < = 0.015%; trace < = S < = 0.015%; 17.0% < = Cr < = 21.0%; trace < = Co < = 1.0%; 2.80% < = Mo < = 3.30%; 4.75% < = Nb + Ta < = 5.50%; 0.65% < = Ti < = 1.15%; 0.20% < = Al < = 0.80%; trace < = Cu < = 0.30%; 50.0% < = Ni < = 55.0%; trace < = B < = 0.006%; the remainder being Fe and impurities resulting from processing, and desensitizing annealing in a static furnace at an annealing temperature equal to or higher than 900 DEG C and / or equal to or lower than 1050 DEG C for an annealing time equal to or greater than 1 minute and / or equal to or less than 180 minutes.

The invention belongs to the technical field of nuclear power maintenance, and particularly relates to a bionic mechanical arm for cleaning foreign matters at the bottom of a nuclear reactor. Comprising an underwater electrical cabin body, an adapter seat, an underwater linear driving rod, a supporting rod, a pulley block, a mechanical arm mounting plate, a mechanical arm module and an underwater suction pump, the underwater linear driving rod is fixed to the underwater electrical cabin body, and the underwater electrical cabin body supplies power and provides a control signal; the two ends of the supporting rod are fixedly connected with the underwater electrical cabin and the mechanical arm mounting plate respectively to connect the underwater electrical cabin and the mechanical arm mounting plate into a whole, the pulley block is fixed to the mechanical arm mounting plate, the mechanical arm modules are rotatably connected end to end to form the bionic mechanical arm, one end of the bionic mechanical arm is fixed to the mechanical arm mounting plate, and the other end of the bionic mechanical arm is connected with the underwater suction pump. The device has the beneficial effect that the problem that small foreign matters at the bottom of a nuclear reactor are difficult to clean can be well solved. The small foreign matters at the bottom of the nuclear reactor can be sucked under the condition that reactor internals at the lower part of the nuclear reactor are not detached.

The invention discloses a spraying type gaseous efficient deoxidizing device and method for lead-based liquid metal. The device and method are applied to gaseous efficient deoxidizing of the lead-based liquid metal. Comprising an inlet connecting pipe, an inlet cavity, a reaction cylinder, an exhaust cylinder, a jet pipe column and a nozzle, the upper end of the reaction cylinder is provided with a top plate, the lower end of the reaction cylinder is provided with a bottom plate, and the bottom of the reaction cylinder is provided with an inlet cavity; a jet pipe column is arranged in the reaction cylinder, the bottom end of the jet pipe column is welded to the bottom plate, the bottom of the jet pipe column is communicated with the inlet cavity, lead-based liquid metal flowing in from the inlet connecting pipe is divided in the inlet cavity and then enters the jet pipe column, and nozzles in the circumferential direction are arranged at the top of the jet pipe column. And the lead-based liquid metal sprayed from the nozzle can enter the reaction cylinder or the exhaust cylinder. The multiple jet flow pipe columns are arranged in the reaction cylinder, lead-based liquid metal is sprayed out from the nozzles after being accelerated in the jet flow pipe columns, and the gas-liquid reaction area is increased; an air inlet pipe is arranged beside the jet pipe column, and the initial turbulent flow and mixing effect is enhanced.

To provide a lower component attaching / detaching device capable of easily attaching a spool piece to a main body of a control rod drive mechanism.SOLUTION: And a push-up mechanism 14 that pushes up and holds the spool piece 32a to a height at which the spool piece 32a can be attached to the drive mechanism body 31 of the control rod drive mechanism 3 in a case where the spool piece 32a that can be included in the control rod drive mechanism 3 that drives the control rod cannot be held by the lifting and lowering mechanism 11 due to the expansion and contraction of the plurality of fluid pressure cylinders 11a.SELECTED DRAWING: Figure 10

The present application belongs to the technical field of nuclear fuel management of nuclear power plant, and particularly relates to a loading method for 177 core refueling overhaul of nuclear power unit. The method comprises the following steps: step 1: loading of fuel assemblies in steps 1-2; step 2: loading of fuel assemblies in steps 3-159a; step 3: loading of irradiated assemblies close to the two sides of the coaming; step 4: loading of fuel assemblies in steps 159a-171; and step 5: loading of fuel assemblies in steps 172-177. The present application has the following beneficial effects: (1) for loading critical supervision, the number of times of suspension of loading work caused by alarm value adjustment of the nuclear instrument system is reduced; (2) in the early and middle stages of loading, the "serpentine" loading scheme, the "offset method" loading operation mode and the use of "small boot" are adopted, so that the fuel operation risk is reduced and the loading efficiency is improved.

A container for storing high assay low enriched uranium hexafluoride (UF6) has a cylindrical lateral wall extending between first and second heads defining a container interior for receiving the radioactive material. A plurality of cylindrical poison rods extend in parallel through the container interior. Each of the poison rods has a neutron absorber. A plurality of access nozzles are situated within the first head and / or second head. Each of the access nozzles has a lateral wall defining a throughway that extends through the first head between an interior end region and an exterior end region. The interior end region captures and supports a respective end of a respective poison rod. The exterior end region is opened and closed via respectively detaching and attaching an end cap to thereby respectively expose and contain the respective poison rod.

The application discloses a high-temperature gas cooled reactor primary loop dehumidification column switching time optimization method and device, and belongs to the technical field of high-temperature gas cooled reactor (HTR-PM) debugging optimization. By acquiring the operation parameters of the accident dehumidification column and the normal purification column and the hot-state water concentration target of the primary loop, the dehumidification process analysis basis is established; based on the rate characteristics of the two types of dehumidification columns under different humidities, the corresponding primary loop water concentration when the dehumidification rates are equal is calculated, and the primary loop water concentration is determined as the optimal switching humidity; under the hot-state working condition, the accident dehumidification column is started to dehumidify, and the primary loop water concentration is monitored in real time; when the water concentration decreases to the optimal switching humidity, the switching operation from the accident dehumidification column to the normal purification column is performed; then, continuous dehumidification is performed until the water concentration reaches the final target value. The accurate judgment and dynamic optimization of the switching time are realized, and the overall efficiency of the dehumidification process and the system operation stability are effectively improved.

The embodiment of the invention relates to the field of iron-based alloys, in particular to a material suitable for a fast reactor fuel assembly and a method for preparing a tube blank suitable for the fast reactor fuel assembly. According to the material, the content of Ni and Mn is reduced, the diffusion speed of M23C6 is reduced, and the coarsening speed of M23C6 is inhibited; n is added, a VN phase is separated out, pinning dislocation is achieved, subcrystal formation is delayed, and the tissue coarsening speed is slowed down; the content of B is reduced, and the content of Cr is increased, so that the generated M23C6 and VN can jointly pinning the lath boundary; nb, Zr, Ta and V are introduced, MX phase precipitation is promoted, MX and M23C6 strengthening and M23C6 and VN pinning lath boundaries are achieved, the structure stability is improved, the high-temperature durability of the material is improved, and the safety requirements for high temperature, high irradiation dose and long-time use of a fast reactor core assembly are met. According to the method, the pipe blank with high-temperature durability and irradiation resistance can be prepared, it is ensured that the cladding obtained through subsequent machining can keep a high-temperature-resistant and irradiation-resistant structure, and the economical efficiency and safety of a reactor are improved.