72 results about "Reactor pressure vessel" patented technology

The invention relates to the technical field of safety monitoring of a nuclear reactor, and particularly discloses a multi-region pressure overrun leakage monitoring method for a reactor pressure vessel. The method comprises the steps of multi-region pressure data acquisition, real-time pressure field model construction, dynamic pressure field baseline modeling, pressure overrun anomaly detection, leakage source accurate positioning and evaluation based on spatial gradient analysis, topological structure mapping and multi-sensor data fusion, and graded early warning and response. By the adoption of the technical scheme, real-time and fine monitoring of pressure distribution in the reactor pressure vessel can be achieved, the rapid and accurate judgment and positioning capacity of leakage events is improved, the operation safety and reliability of nuclear power facilities are enhanced, and the reliability of the nuclear power facilities is improved. The defects of independent monitoring, local accurate positioning and complex working condition adaptability of multi-region pressure over-limit leakage in the prior art are overcome.

This application discloses a method and system for identifying internal defects in a reactor pressure vessel, a computer device, and a computer-readable storage medium. The method includes the steps of: collecting image data of the pressure vessel surface; constructing a light irradiation angle attention mechanism to create a feature map using the images in the dataset and generating a spatial attention map based on the spatial relationships of the feature maps; and improving a YOLOv5 model using the light irradiation angle attention mechanism and identifying surface defects in the pressure vessel using the improved YOLOv5. It features high defect identification efficiency and a low detection miss rate. It effectively addresses common problems when performing defect detection under various light irradiation conditions. This ability to accurately identify and distinguish defects is crucial for ensuring the safe operation of the pressure vessel, especially during routine maintenance and inspections.

The present application relates to a reactor pressure vessel protection method and system, the method comprising the following steps: when the unit is in a downward state, judging whether the residual heat removal system is in an effective state, if the result is yes, generating a cold-state overpressure protection signal for activating a cold-state overpressure protection means; if the result is no, further judging whether the primary loop average temperature is less than a first conversion temperature, if the result is yes, generating the cold-state overpressure protection signal; when the unit is in an upward state, judging whether the primary loop average temperature is greater than a second conversion temperature, if the result is yes, generating a hot-state overpressure protection signal for activating a hot-state overpressure protection means; the present application can control the unit to switch between the cold-state overpressure protection state and the hot-state overpressure protection state according to the primary loop average temperature, thereby executing the corresponding overpressure protection means, and effectively preventing the reactor pressure vessel from breaking.

An underwater crawling device for a nuclear reactor comprises: a drive control mechanism; the crawling mechanism is connected with the driving control mechanism, the crawling mechanism comprises a machine body and a plurality of crawling legs, each crawling leg comprises a translation component and a lifting component, the translation component is connected with the machine body, the lifting component is connected with the translation component, and the machine body is connected with the machine body. The driving control mechanism controls the translation part and the lifting part to move in sequence according to a set sequence; the plurality of crawling legs are matched with one another to drive the machine body to crawl and move to a set position; wherein the lifting component is provided with a positioning guide pin corresponding to a flow hole of a lower grid plate of a reactor core in a reactor pressure vessel, and the lifting component is inserted into the flow hole through the positioning guide pin to realize crawling movement; a machine body guide pin matched with the flow hole is arranged on the bottom face of the machine body, and the machine body guide pin is inserted into the flow hole to achieve positioning and bear horizontal component force.

This invention belongs to the field of reactor structural mechanics technology, specifically relating to a method for analyzing and evaluating high-temperature bolted connection sealing structures. The invention includes the following steps: selecting the location of the reactor pressure vessel flange and main bolt connection as the object of sealing structure analysis; analyzing and summarizing the influence of gasket data on the deformation and contact surface pressure of various components of the flange bolted connection structure; researching current standards and analyzing their limitations; using the finite element software ANSYS to perform finite element modeling of the reactor pressure vessel flange and main bolt structure; applying loads and constraints to the finite element model of the reactor pressure vessel flange and main bolt structure; calculating the deformation and stress of the structure using different creep and relaxation models; analyzing the relaxation behavior of the high-temperature flange bolted connection structure using experimental methods; and comparing and analyzing the results of the finite element method, the standard method, and the experimental method. This invention enables the analysis and evaluation of the sealing performance of flange bolted connection structures under high-temperature environments.

The invention belongs to the technical field of distributed control, and particularly relates to a distributed reactor pressure vessel inspection control system. Comprising a motion controller, a main control computer, sub-control computers and a lower computer, the main control computer is respectively connected with the plurality of sub-control computers to form an upper computer system, the sub-control computers communicate through the main control computer to obtain information and control authority, and the main control computer uniformly communicates with the lower computer to send information. The distributed control system has the beneficial effects that the independence of each manipulator and an inspection tool is ensured by the distributed control system architecture, and meanwhile, the management of control authority and the reading of data of all controllers in the system are realized through the architecture of the master control computer and the sub-control computer.

The invention discloses a PVT synthesis-based ultrasonic inspection control method for a welding seam of a connecting pipe of a reactor pressure vessel. The method comprises the following steps: S1, setting parameters; s2, obtaining a spatial trajectory discrete point sequence; s3, the shaft speed is constant; s4, any two continuous spatial trajectory discrete points in the spatial trajectory discrete point sequence are selected; according to the movement time t between the two points, the axis and the component velocity of the axis are obtained through calculation; s5, based on the three-axis component, PVT interpolation operation between two points is carried out, and a smooth track is generated; s6, repeatedly executing the steps S4-S5; and S7, performing incremental motion according to the stepping interval, and repeating the steps S2-S6. Aiming at the special geometrical characteristics of the saddle surface shape of the connecting pipe, a PVT synthesis algorithm is designed by virtue of a linear rotating device, a scanning motion track is obtained, the fitting state of the ultrasonic probe and the detected surface is accurately controlled, and automatic nondestructive detection of the welding seam in the inner fillet connecting pipe area is realized.

The invention provides a position monitoring system for a nuclear instrument of a nuclear power plant and an arrangement method. The position monitoring system comprises a guide cylinder, a first position monitoring unit, a second position monitoring unit and an upper computer. The guide cylinder is pre-buried at the periphery of a reactor pressure vessel, and a mounting channel for nuclear instruments to pass through is arranged in the guide cylinder; the first position monitoring unit is used for monitoring the current position of the nuclear instrument in real time, is arranged on the side wall of the guide cylinder and corresponds to the upper side of the working position of the nuclear instrument; the second position monitoring unit is used for monitoring the current position of the nuclear instrument in real time, is arranged on the side wall of the guide cylinder and corresponds to the lower side of the working position of the nuclear instrument; and the upper computer is in communication connection with the first position monitoring unit and the second position monitoring unit and is used for acquiring the current position of the nuclear instrument monitored by the first position monitoring unit and the second position monitoring unit in real time. The nuclear instrument installation precision requirement can be met, and the arrangement efficiency is improved.

The invention particularly relates to a defect grinding device for a sealing surface of a reactor pressure vessel, which belongs to the field of nuclear power equipment maintenance and comprises a fixing seat, a grinding head and a grinding head, the three-coordinate driving assembly is mounted on the fixed seat and is used for driving the grinding assembly to perform positioning in an XY plane and feed movement in a Z direction; the grinding assembly is used for grinding defects of the sealing surface; the waste chip collecting assembly is integrated on the grinding assembly and is used for collecting waste chips and cooling the grinding assembly while grinding; and the control system is used for controlling the three-coordinate driving assembly to work, calculating the real-time grinding depth and stopping the grinding assembly from working after the preset depth is reached. According to the device, online efficient collection of grinding scraps, automatic and accurate control over the grinding depth and rapid and stable installation and self-adaptive leveling of the device are achieved, and therefore the safety, the automation degree and the machining quality of repairing operation are improved.

The invention belongs to the technical field of equipment maintenance, and discloses a repairing device and a repairing method.The repairing device comprises a fixed connecting piece which comprises at least two fixed sub-connecting pieces, the at least two fixed sub-connecting pieces are sequentially connected end to end in the circumferential direction of a first axis and detachably connected, and a fixed hole is defined by the at least two fixed sub-connecting pieces; the axis of the fixing hole is parallel to a first axis, and the first axis extends in the first direction; the rotary connecting piece is movably connected to the fixed connecting piece and can move around a first axis; and the cutter is connected to the rotary connecting piece. The technical problem that the reactor pressure vessel has a leakage risk after the conical sealing surface on the thermocouple column is damaged can be solved.

The invention relates to the technical field of reactor safety analysis, and provides a pressurized water reactor pressure vessel lower chamber flow distribution analysis method. The method comprises the following steps: establishing a fluid domain three-dimensional geometric model; performing unstructured grid division on the fluid domain three-dimensional geometric model to generate a three-dimensional calculation grid; selecting a turbulence model; setting inlet boundary conditions and physical property parameters of the coolant; the reactor core area is equivalent to a porous medium area, and flow resistance parameters of the porous medium area are set; based on the three-dimensional calculation grid, adopting a selected turbulence model, loading flow resistance parameters of a porous medium area and inlet boundary conditions and physical property parameters of a coolant, and carrying out three-dimensional flow simulation to obtain coolant flow distribution at the section of a reactor core inlet; and evaluating the flow distribution uniformity of the reactor core inlet according to a preset flow distribution evaluation criterion. The method is used for simulating three-dimensional flow of a coolant in a lower cavity of a reactor pressure vessel, acquiring a flow field structure and flow distribution data of the coolant, and performing quantitative evaluation.

The invention belongs to the technical field of intelligent monitoring of reactors, and particularly relates to a method, a system, equipment and a medium for monitoring a reactor pressure vessel. Comprising the following steps: acquiring container monitoring data and judging whether an initial monitoring condition is met or not, and if so, acquiring the container monitoring data again after a preset monitoring period for judgment; if not, acquiring a risk monitoring threshold value corresponding to the container monitoring data according to the monitoring model, and judging whether the container monitoring data has a risk in combination with a risk monitoring strategy, and if the risk monitoring result is that the risk exists, generating corresponding monitoring prompt information; if the risk monitoring result is that the risk does not exist, the container monitoring data are obtained again after a preset monitoring period for judgment. The method has the advantages that the risk monitoring threshold values are flexibly adjusted according to different pressure vessel devices for risk judgment, and the accuracy and reliability of pressure vessel monitoring are improved by obtaining the risk monitoring threshold values corresponding to the devices in a targeted mode.

Disclosed is a reactor pressure vessel direct injection safety system, which comprises a first water injection pipe and / or a second water injection pipe. The first water injection pipe passes through an upper head of a reactor pressure vessel, an input end of the first water injection pipe is located outside the reactor pressure vessel and connected to a coolant source, and an output end of the first water injection pipe is located inside the reactor pressure vessel and extends through an upper core support plate to a core region. The output end of the first water injection pipe can be directly positioned to the core region to output coolant, thus enhancing core cooling efficiency. The second water injection pipe passes through the reactor pressure vessel, an input end of the second water injection pipe is located outside the reactor pressure vessel and connected to the coolant source, and an output end of the second water injection pipe is located inside the reactor pressure vessel and at least extends to a downcomer section of the reactor pressure vessel. The output end of the second water injection pipe can output the coolant directly from the downcomer section, thus facilitating flow of more coolant from a lower core support plate to the core region to serve the function of core cooling.

The application discloses a reactor pressure vessel top cover nondestructive testing equipment and a positioning method and a testing method. The testing equipment comprises a testing driving part, a positioning part and the like. The testing driving part comprises a flexible chain mounting plate, an adjustable angle flexible chain mounted on the flexible chain mounting plate, and a probe assembly mounted on the flexible chain. The flexible chain moves in a first direction on a second plane parallel to a first plane. The probe assembly moves in a second direction perpendicular to the first direction on the second plane. The positioning part comprises a positioning mounting plate, a laser emitter and a camera mounted on the positioning mounting plate. The laser emitted by the laser emitter is perpendicular to the first direction. Through the perpendicular arrangement of the laser emitter and the movement direction of the flexible chain, the position calibration of the probe in the first direction and the second direction can be quickly realized, and the implementation is simple and effective. The accurate positioning of the testing equipment under the complex curved surface environment of the top cover surface can be realized.

The application discloses a marine bottom shielding structure combining a cabin bottom shielding structure and a stack bottom shielding structure, which comprises the cabin bottom shielding structure and the stack bottom shielding structure located on the lower plate of the double bottom of a ship body, and a stack cabin shielding structure and a primary shielding structure located on the upper plate of the double bottom of the ship body, wherein the primary shielding structure is wrapped outside a reactor pressure vessel, the stack cabin shielding structure is located outside the primary shielding structure, the cabin bottom shielding structure is located outside the stack bottom shielding structure, and the stack bottom shielding structure is located below the primary shielding structure. The marine bottom shielding structure combining the cabin bottom shielding structure and the stack bottom shielding structure has simple structure, can form a complete radiation shielding system together with the stack cabin shielding structure, the primary shielding structure and seawater below the ship body, and can weaken the radiation leakage of the stack cabin to the surrounding area to the maximum extent, so as to ensure the radiation safety of personnel and equipment in the surrounding area.

The application relates to a method for constructing and controlling turbulence morphology in a reactor pressure vessel, which includes the following steps: S1, obtaining a vortex relevant data, obtaining a temporal distribution characteristic, a spatial distribution characteristic, a power spectral density of the turbulence, and turbulence energy spectrum at the inlet region of each fuel assembly; S2, obtaining the water gap variation of the fuel assembly, and determining whether it is needed to modify the three-dimensional flow field model of the fuel assembly based on the water gap variation of the fuel assembly until the water gap of the fuel assembly does not change; S3, obtaining a core nuclear power fluctuation amplitude among the fuel assemblies; S4, determining whether the core nuclear power fluctuation amplitude falls within a target range, and if not, modifying the internal three-dimensional structure of reactor pressure vessel until the core nuclear power fluctuation amplitude falls within the target range. The application can eliminate or improve the core nuclear power fluctuation, reduce the wear of the reactor core grid, and ensure the safety of reactor operation.

A nozzle expansion tool includes a frame with a drive system on the frame. A rotary mandrel is drivingly connected to the drive system and is engageable with an expansion roller device. A plurality of vacuum cups are mounted to the frame and each include a vacuum fitting adapted to be connected to a vacuum source. A depth adjustment mechanism is connected to the expansion roller device and is configured to adjust a distance that the expansion roller device extends from the frame.

This application provides an online measurement system for irradiated stress corrosion crack propagation in reactor materials, comprising: a reactor pressure vessel with a test section inside; a test unit disposed inside the test section, containing a sample; and a load adjustment unit connected to the test unit for applying tension to the sample. The beneficial effect of this invention is that by configuring the reactor pressure vessel, test unit, load adjustment unit, and measurement control unit, the sample is installed inside the test unit, and the measurement control unit controls the load adjustment unit to apply a load to the sample, thereby achieving in-reactor stress loading. By connecting multiple measuring elements to the sample, the potential change generated by crack propagation in the sample can be monitored in real time, achieving real-time monitoring of crack propagation.

The invention relates to the technical field of digital twinning, and discloses a reactor pressure vessel life prediction method and system based on digital twinning, and the method comprises the steps: carrying out the temperature change rate and stress distribution extraction of real-time monitoring data, and obtaining dynamic characteristic parameters; performing consistency verification on the reference output in the initial model based on the dynamic characteristic parameters to obtain a parameter suitability evaluation result of the reactor pressure vessel; adjusting numerical values of physical parameters in the initial model according to a parameter suitability evaluation result, and reconstructing the initial model based on the adjusted physical parameters to obtain a digital twin model; simulating a thermal-mechanical behavior of the reactor pressure vessel under the current operation condition to obtain a damage accumulation record of the reactor pressure vessel; performing damage integration on the reactor pressure vessel to obtain a total damage index; mapping the total damage index into the residual life of the reactor pressure vessel; the method can improve the accuracy of life prediction of the reactor pressure vessel.

The invention discloses a laser ultrasonic inspection device for an end socket bolt of a reactor pressure vessel. The laser ultrasonic inspection device comprises a rack, a guide transmission mechanism, a pair of detection rotating mechanisms, a clamping mechanism, a horizontal turnover mechanism and a detection action mechanism. According to the laser ultrasonic inspection device for the end socket bolt of the reactor pressure vessel, clamping and posture adjustment of the end socket bolt and excitation and reception of ultrasonic signals are realized through cooperation of the rack, the guide transmission mechanism, the detection rotating mechanism, the clamping mechanism, the horizontal turnover mechanism and the detection action mechanism. Compared with a traditional detection device, the device is high in integration level and capable of achieving comprehensive detection of the end socket bolt, a laser ultrasonic detection mode is adopted, direct contact with the surface of the end socket bolt is not needed, and damage to the surface of the end socket bolt is avoided. The pair of detection rotating mechanisms is provided with an ultrasonic detection probe and a pulse laser respectively, the relative position and angle of the ultrasonic detection probe and the pulse laser can be flexibly adjusted, and the accuracy and flexibility of detection are ensured.

The invention discloses a reactor pressure vessel connecting pipe welding seam ultrasonic inspection control method based on vector synthesis. The method comprises the steps that S1, parameters are set; s2, obtaining a spatial trajectory discrete point sequence; s3, acquiring the radial movement speed and the circumferential movement speed of the ultrasonic probe during movement; S4, selecting two continuous discrete points in the spatial trajectory discrete point sequence; s5, according to the movement time between the two points, the speed and the vector resultant speed of the three axes are obtained through calculation; s6, performing vector interpolation calculation according to the motion distance and the vector resultant speed of the axis, and generating a section of motion control track between the two points; and S7, repeating the steps S4-S5 until the vector synthesis motion of all discrete points of the spatial trajectory discrete point sequence is completed. Aiming at the special geometrical characteristics of the saddle surface shape of the connecting pipe, a vector motion synthesis algorithm is designed by virtue of a linear rotating device, the fitting state of the ultrasonic probe and the detected surface is accurately controlled, and the automatic nondestructive detection of the welding seam in the inner fillet connecting pipe area is realized.

To provide a support structure of a nuclear reactor pressure vessel capable of restricting displacement in a vertical direction and easily adjusting a gap between shear lag members, and a nuclear power plant.SOLUTION: A support structure includes a female shear lug 101 divided into two members in a vertical direction 111B and a male shear lug 102 supported by the female shear lug 101. The female shear lug 101 includes a first member 101A and a second member 101B disposed below the first member 101A in the vertical direction 111B with the male shear lug 102 interposed therebetween. The first member 101A and the second member 101B are disposed with a gap 112 with the male shear lug 102 in the vertical direction 111B.SELECTED DRAWING: Figure 3

The invention relates to the technical field of safety detection of nuclear reactor equipment, in particular to an irradiation damage nondestructive detection method for a reactor pressure vessel and related equipment. The method comprises the following steps: carrying out material magnetic non-uniformity reference analysis before a reactor pressure vessel is put into operation to obtain corresponding original material magnetic distribution characteristics and a magnetic reference database, carrying out detection region division, and determining a plurality of vessel detection regions; performing local magnetization reference retest on each container detection area to calibrate the magnetic reference database; in the process that the reactor pressure vessel is put into operation, each vessel detection area is magnetized again, and corresponding target area magnetic field signals are collected; and according to the magnetic reference database and the target area magnetic field signal, performing damage evaluation on each container detection area to obtain a damage evaluation result. Therefore, the nondestructive testing and evaluation of the material irradiation damage of the reactor pressure vessel can be carried out, and the method has better accuracy performance.

The present invention provides a bottom closure of a reactor pressure vessel and a preparation method therefor. The bottom closure of a reactor pressure vessel comprises a bottom closure body and a po

The utility model discloses a leveling and stabilizing device for replacing a top cover of a reactor pressure vessel, which is used for replacing a top cover component of the reactor pressure vessel, and the top cover component of the reactor pressure vessel comprises a reactor top structure, a control rod driving mechanism and the top cover of the reactor pressure vessel, the reactor top structure comprises a plurality of suspenders connected with the top cover of the reactor pressure vessel, and is characterized in that the leveling and stabilizing device for replacing the top cover of the reactor pressure vessel comprises a stabilizing part in a closed ring shape; and each adjusting assembly comprises a limiting hole used for allowing the hanging rod to penetrate through, and the multiple adjusting assemblies are all connected to the stabilizing piece. The leveling and stabilizing device for replacing the top cover of the reactor pressure vessel can effectively prevent the suspender from inclining and twisting.

This invention belongs to the technical field of nuclear power plant reactor pressure vessel maintenance, specifically relating to a device and method for disassembling and assembling the main nut of a reactor pressure vessel. The device includes: a support sleeve, a locking ring, inserts, and lifting hooks. Several lifting hooks are evenly connected along the lower outer periphery of the support sleeve, and these hooks can open and close relative to the support sleeve. The bottom of each lifting hook has an inwardly facing protrusion for hooking the main nut or a main nut washer. Several inserts are evenly connected to the bottom of the support sleeve, the shape of which matches the U-shaped groove on the main nut, for applying torque to the main nut during rotation. The locking ring is sleeved around the outer periphery of the lifting hooks for fixing them. This invention has a simple structure and is easy to operate, effectively improving work efficiency while ensuring that the lifting hooks will not open or close accidentally during operation, greatly reducing operational risks.

The invention belongs to the technical field of nuclear power station primary loop sealing, and particularly relates to a nuclear power station VVER unit reactor pressure vessel main seal test method, which comprises the following steps: S1, performing multi-material staged extrusion to simulate the deformation quantity of a sealing structure, and performing a helium leakage rate test after each simulation, carrying out a hydraulic pressure test under the condition that the deformation quantity of the V-shaped groove of the cylinder body is maximum, so as to obtain the maximum critical defect; the sealing structure is composed of a pressure container, a pressure container top cover and a pure nickel sealing ring. And S2, comprehensively calibrating the deformation quantity of the V-shaped groove of the barrel and the sealing plane of the top cover by adopting the protruding quantity of the steel balls. According to the application, the deformation quantity of the sealing structure is simulated through multi-material staged extrusion, and the deformation quantity is comprehensively calibrated by adopting the protruding quantity of the steel balls, so that the technical problems of simulation and measurement of the plastic deformation of the main sealing structure of the reactor pressure vessel of the VVER of the nuclear power station are solved.

The invention provides an abnormity determination method and device for a reactor pressure vessel material and electronic equipment, and relates to the technical field of nuclear power station safety management, and the method comprises the steps: obtaining the appearance scanning data of the reactor pressure vessel material, and determining the microscopic feature data according to the appearance scanning data; target working conditions are determined, and under the target working conditions, the brittle transition temperature initial value of the reactor pressure vessel material is determined based on the microscopic characteristic data; predicting the expected variation of the brittle transition temperature of the material according to the initial value of the brittle transition temperature; determining an abnormal result of the brittle transition temperature of the material according to the expected variable quantity; the technical problem that key parts are not monitored in time in the prior art is solved, and safe operation of the reactor pressure vessel is ensured.