156 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.

The invention provides a reactor pressure vessel flange sealing surface defect measuring method and a reactor pressure vessel flange sealing surface defect repairing method. The measuring method comprises the following steps: S1, measuring and calculating the minimum diameter and the maximum diameter of an annular band formed in a sealing ring indentation range; and S2, by taking the center of the container flange as a circle center, inspecting the defect, falling into the range of the annular belt between the minimum diameter and the maximum diameter, on the container flange, and measuring and recording the size and the orientation of the defect. According to the method, corresponding recommended repair methods are provided according to repair criteria for defects of different scales, the method can be used for guiding implementation of nuclear power plant RPV sealing surface maintenance, the repair workload is effectively reduced, the radiation dose of personnel is reduced, the operation and maintenance efficiency is improved, safe and stable operation of a nuclear power unit is ensured, and the economic benefit of power station operation is improved.

The invention provides a floating nuclear power platform and a radiation shielding system thereof. The radiation shielding system comprises a primary shielding structure, a combined shielding structure and a containment shielding structure, the primary shielding structure covers the outer side of a reactor pressure vessel to perform primary shielding on the reactor pressure vessel, the combined shielding structure sleeves a loop pipeline and accessory equipment thereof, and the containment shielding structure is arranged in the containment structure; the primary shielding structure comprises a first heat preservation layer wrapping the outer side of the pressure vessel shell, a shielding water tank covering the first heat preservation layer and a shielding material layer covering the outer side of the shielding water tank, and the shielding water tank is hollow to contain water. The combined shielding structure comprises a high-density alloy layer and a second thermal insulation layer; the high-density alloy layer is sleeved outside the loop pipeline and accessory equipment thereof; the second thermal insulation layer is sleeved outside the high-density alloy layer; the containment shielding structure comprises a lead plate, a polyethylene plate layer, a boron-containing polyethylene plate layer and an outer steel plate layer which are arranged in the containment structure. According to the radiation shielding system provided by the invention, the total radiation shielding weight can be reduced by 20-40% while the radiation shielding effect is guaranteed.

To suppress horizontal vibration of a nuclear pressure vessel during an earthquake, and make a dynamic vibration absorber compact to suppress an installation space of the dynamic vibration absorber, in a nuclear power plant.SOLUTION: An earthquake resistance support device for a nuclear power plant is arranged in a reactor containment vessel, and suppresses horizontal vibration of a reactor pressure vessel supported by a pedestal when an earthquake occurs. The earthquake resistance support device comprises a horizontal beam that is arranged between the reactor containment vessel and the reactor pressure vessel, and the pedestal, and extends in the radial direction of the reactor pressure vessel and the pedestal; and a dynamic vibration absorber installed on the horizontal beam. The dynamic vibration absorber comprises a hollow weight extrapolated to the horizontal beam and movable in an axial direction of the horizontal beam, and a coil spring device that elastically supports the hollow weight in the axial direction of the horizontal beam, to allow the hollow weight to vibrate in the axial direction of the horizontal beam.SELECTED DRAWING: Figure 3

The invention discloses a reactor pressure vessel top cover penetration piece inspection device which comprises a supporting mechanism, a main rotating mechanism, a translation mechanism, a connecting mechanism, a space adjusting mechanism, a detection executing mechanism, a water circulation mechanism and a calibration mechanism. According to the reactor pressure vessel top cover penetration piece inspection device, by arranging the main rotating mechanism, the translation mechanism, the connecting mechanism and the space adjusting mechanism, the space position of the detection executing mechanism is adjusted, so that the detection executing mechanism can inspect reactor pressure vessel top cover penetration pieces at different positions. And meanwhile, a water circulation mechanism is arranged to supply and recycle coupling water utilized by the detection execution mechanism, and a calibration mechanism is arranged to calibrate the detection execution mechanism, so that the working accuracy and stability of the detection execution mechanism are ensured. The reactor pressure vessel top cover penetration piece inspection device is complete in module and has important significance for guaranteeing all-directional inspection work of the reactor pressure vessel top cover penetration piece.

The invention discloses a passive reactor core and containment cooling and heat extraction system which comprises a steam generator, an air cooling heat exchanger, an exhaust unit and a water cooling heat exchanger, the primary side of the steam generator is communicated with the reactor pressure vessel; a secondary side outlet of the steam generator communicates with a heat source channel inlet of the air-cooling heat exchanger through a first pipeline; a heat source channel outlet of the air-cooling heat exchanger communicates with a heat source channel inlet of the water-cooling heat exchanger; a heat source channel of the air-cooling heat exchanger communicates with an inlet of the exhaust unit; a heat source channel outlet of the water-cooling heat exchanger is communicated with a secondary side inlet of the steam generator through a second pipeline; the interior of the containment is communicated with the first pipeline through a third pipeline; and the interior of the containment is communicated with the second pipeline through a fourth pipeline. The invention also discloses a method for cooling and heat removal of the reactor core and the containment. The system has the beneficial effects that the reactor core heat removal and the containment cooling are integrated, so that the number of system equipment is greatly reduced, and the occupation of total resources is reduced.

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.

To adjust a core flow rate higher than that in a stationary state when a power change operation of a natural circulation reactor is conducted by a control rod, thus improving a power change speed while maintaining a stable condition.SOLUTION: A nuclear power plant includes a large-diameter space chimney at an upper part of a reactor core within a shroud of a reactor pressure vessel. The nuclear power plant includes: a coolant flowmeter configured to measure a coolant flow flowing into the reactor core; a water level gauge configured to measure a water level of a coolant within the reactor pressure vessel; a water-supply flowmeter connected to the reactor pressure vessel and configured to measure a water-supply flow rate; a monitoring device configured to monitor the coolant flow, the water level and the water-supply flow rate; and a controlling device configured to control a control rod and the water-supply flow rate. The controlling device is configured to: calculate, in advance, a fluctuation margin of a core flow rate by a time lag of the water-supply flow rate; change a target value of the water level to be higher than a fluctuation margin of the water level corresponding to the fluctuation margin of the core flow rate so as to change the water-supply flow rate and then change the water level; and thereafter operate and control a position change of the control rod.SELECTED DRAWING: Figure 4

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 relates to a device for detecting penetration pieces of a reactor pressure vessel, which comprises the reactor pressure vessel, a top cover is arranged at the top of the reactor pressure vessel, a plurality of penetration pieces are arranged on the top cover, a rigid chain execution device is arranged in the reactor pressure vessel, and the rigid chain execution device comprises a rigid chain driving box. A planetary reducer is fixedly installed on the outer wall of the rigid chain driving box through bolts, a servo motor is installed on the planetary reducer, the output end of the servo motor is fixedly connected with the input end of the planetary reducer, and a plurality of rigid chains are fixedly installed at the top end of the planetary reducer. The multi-section rigid chain is composed of four sections of hollow circular pipes which are sleeved from inside to outside and arranged in a sliding mode. The rigid chain execution device is arranged in the device for detecting the penetration piece of the reactor pressure vessel, so that the detection of the penetration piece of the top cover of the reactor pressure vessel is efficiently and stably completed, and the safe operation of the reactor pressure vessel is ensured.

The utility model discloses a fastener for a top cover of a reactor pressure vessel of a nuclear power station, which comprises a chuck and a penetrating rod, a fixing shaft is arranged at the connecting position of the penetrating rod and the chuck, a threaded rod is arranged at the upper side position of the penetrating rod, a fixing nut is arranged at the external position of the threaded rod, and the fixing nut is connected with the chuck. Rubber pads are arranged on the outer sides of the penetrating rods, the clamping plates can rotate in the inner grooves, the rotating plates can rotate on the clamping plates through the movable shafts, the connecting holes and the clamping holes are used for being matched with connecting bolts to conduct fixing and other operations, and the overall height of the chuck can be increased according to needs. Therefore, nuclear power plant reactor pressure vessel top covers (hatch covers) with different thicknesses can be effectively clamped, the application range of the fastener is greatly widened, fasteners with different specifications do not need to be frequently replaced due to the thickness difference of the hatch covers, and the use cost and the reserve pressure of spare parts are reduced.

The embodiment of the invention provides a waste heat removal system of a high-temperature gas cooled reactor, and relates to the technical field of nuclear reactors. The high-temperature gas cooled reactor waste heat removal system comprises a reactor cabin, a reactor pressure vessel, a heat exchanger, a cooler, a water inlet pipe, a water outlet pipe, a water replenishing isolation valve, a water return isolation valve and a waste heat auxiliary removal device, the water inlet pipe and the water outlet pipe are communicated with the heat exchanger and the cooler at the same time; the auxiliary waste heat discharging device is simultaneously communicated with the water inlet pipe and the water outlet pipe, the water supplementing isolating valve is connected between the auxiliary waste heat discharging device and the water outlet pipe, and the water returning isolating valve is connected between the auxiliary waste heat discharging device and the water inlet pipe; under the condition of the first waste heat discharging mode, the water supplementing isolating valve and the water returning isolating valve are both in a closed state; and under the condition of the second residual heat removal mode, the water supplementing isolation valve and the water returning isolation valve are both in an open state. The device can be suitable for different operation conditions, the operation flexibility is high, and the waste heat removal efficiency is improved.

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.

This invention discloses a reactor pressure vessel inspection device and method. The reactor pressure vessel inspection device includes a support frame, a lifting mechanism mounted on the support frame, a rotating mechanism mounted on the lower part of the lifting mechanism facing the vessel body, a nozzle scanning arm mounted below the rotating mechanism, and two multi-degree-of-freedom robotic arm probe frames. The lifting mechanism is movable relative to the support frame, driving the rotating mechanism and its nozzle scanning arm and multi-degree-of-freedom robotic arm probe frames to move up and down along the central axis of the vessel body. The rotating mechanism drives the nozzle scanning arm and multi-degree-of-freedom robotic arm probe frames to rotate. This invention uses two multi-degree-of-freedom robotic arm probe frames to cooperate with the nozzle scanning arm to perform inspection work on the reactor pressure vessel. The multi-degree-of-freedom robotic arm probe frames and the nozzle scanning arm can scan synchronously and in parallel, reducing inspection time, improving work efficiency, and greatly reducing the critical path time for nuclear power plant overhauls.

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 present invention discloses a primary circuit hot test system and method for a high-temperature gas-cooled reactor nuclear power plant. The system and method include graphite and carbon brick internals, a reactor pebble bed core, a reactor pressure vessel, a steam generator, a primary circuit concrete cabin, a first stop valve, a cabin ventilation device, a second stop valve, a vacuum pump, a third stop valve, a fourth stop valve, a fifth stop valve, an electric heater, a helium-helium cooler, a cryogenic adsorption bed, a helium compressor, a sixth stop valve, a helium-water cooler, a gas-water separator, and a seventh stop valve. The system and method can accelerate the helium heating rate during the primary circuit hot test, improve the heating and dehumidification efficiency, shorten the hot test time, and improve the hot test efficiency.

The embodiment of the present application discloses a method for disassembling and assembling a hot gas duct for maintenance of a high-temperature gas-cooled reactor nuclear power plant. During disassembly and assembly, first, the connection between the fourth hot gas duct section and the third hot gas duct section, as well as the connection between the steam generator and the hot gas duct housing, is released, and the steam generator is moved away from the reactor pressure vessel to provide space for the disassembly of the hot gas duct. Secondly, the fourth hot gas duct section is translated with the steam generator, the third hot gas duct section is disassembled and assembled within the operation space vacated after the steam generator is displaced, the hot gas duct housing and the second hot gas duct section are lifted and lowered synchronously, and the first hot gas duct section is disassembled and assembled within the space vacated by the overall downward movement of the hot gas duct housing and the second hot gas duct section. According to the length characteristics of each structure of the hot gas duct, the disassembly and assembly steps are reasonably arranged to create conditions for the replacement of the hot gas duct sealing ring, the inspection of the connection bolts of each section, and other maintenance work.

The present invention discloses a device for inspecting penetration pieces of a reactor pressure vessel top cover, which includes a supporting mechanism, a main rotating mechanism, a translation mechanism, a connecting mechanism, a space adjustment mechanism, a detection actuator, a water circulation mechanism, and a calibration mechanism. The device for inspecting penetration pieces of a reactor pressure vessel top cover is configured to adjust the spatial position of the detection actuator by providing a main rotating mechanism, a translation mechanism, a connecting mechanism, and a space adjustment mechanism, so that the detection actuator can inspect penetration pieces of the reactor pressure vessel top cover at different positions. At the same time, a water circulation mechanism is provided to realize the supply and recovery of coupling water used by the detection actuator, and a calibration mechanism is provided to calibrate the detection actuator to ensure the accuracy and stability of the detection actuator. The device for inspecting penetration pieces of a reactor pressure vessel top cover is complete in modules, which is of great significance for ensuring the all-round inspection of the penetration pieces of the reactor pressure vessel top cover.

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 field of nuclear power plant detection, and discloses a support platform and a nuclear power plant reactor pressure vessel detection system. The support platform comprises a support, and the support comprises a main bearing column, a supporting leg installed on the outer side of the main bearing column, and an installation assembly and an auxiliary mechanism which are arranged on the outer sides of the main bearing column and the supporting leg; the auxiliary mechanism comprises a weight reduction cavity formed in the main bearing column and a reference block assembly embedded in the main bearing column, and the weight reduction cavity comprises a cavity, a drainage hole and an exhaust hole; the main bearing column body can be installed and fixed in the nuclear reactor pressure vessel by arranging the supporting feet, the weight and buoyancy borne by the main bearing column body can be reduced by arranging the weight reduction cavity, and therefore supporting force needing to be provided by a hydraulic rod is reduced, pressure borne by the inner surface of the pressure vessel and a surfacing layer is reduced, and the service life of the pressure vessel is prolonged. Therefore, the purpose of protecting the inner surface and the surfacing layer is achieved.

The utility model discloses an inflation protection device for welding a thermocouple penetration piece of a high-temperature gas cooled reactor pressure vessel, which comprises a protection gas tank, a protection gas main pipeline and at least two protection gas branch pipelines, one end of the protection gas main pipeline is connected with the protection gas tank, and the protection gas branch pipelines are respectively communicated with the protection gas main pipeline. The protective gas branch pipeline is provided with a switch valve, one end, far away from the protective gas main pipeline, of the protective gas branch pipeline enters a nozzle of the high-temperature gas cooled reactor pressure vessel, the thermocouple penetration piece and the nozzle are sleeved and welded together, and the switch valve is located outside the high-temperature gas cooled reactor pressure vessel. All shielding gas branch pipelines required by a certain stage can be arranged in the pipe nozzle of the high-temperature gas cooled reactor pressure vessel in advance, and the on-off of the shielding gas is controlled outside the high-temperature gas cooled reactor pressure vessel, so that the suffocation risk caused by personnel entering the high-temperature gas cooled reactor pressure vessel with the shielding gas can be avoided, the personal safety is ensured, and the working efficiency is improved. Operation is convenient, and construction efficiency is improved.

The invention discloses a mechanical fuel ball lifting device for a high-temperature gas cooled reactor, which comprises a lifting pipe provided with a lifting pipe feed port and a lifting pipe discharge port, the lifting pipe feed port is communicated with the discharge port of a reactor pressure vessel, and the lifting pipe discharge port is communicated with the feed port of the reactor pressure vessel; the clamp holder is provided with an open first position and a closed second position; the clamping device driving mechanism is connected with the clamping device; when the clamping device driving mechanism drives the clamping device to move to the position of the feeding port of the lifting pipe, the clamping device is located at the first position; after the fuel ball enters the clamp holder, the clamp holder is converted from the first position to the second position; and when the clamping device driving mechanism drives the clamping device to move to the position of the discharging port of the lifting pipe, the clamping device is converted into the first position from the second position. The fuel balls are directly grabbed through the mechanical clamping device and stably lifted along the axis of the lifting pipe, pneumatic friction collision between the fuel balls and the pipeline is avoided, the generation rate of chippings is greatly reduced, and therefore the blocking risk is reduced.

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 present invention relates to a method and system for positioning a horn cover of a reactor pressure vessel top cover penetration piece, comprising the steps of: obtaining a target priori frame of a pixel position of a target horn cover; obtaining an initial position of the target horn cover; detecting the current position of the target horn cover using video target detection technology to obtain the current pixel position of the target horn cover in an image; determining whether the current position of the target horn cover meets pixel accuracy requirements based on the current pixel position and the target priori frame; if the pixel accuracy requirements are met, determining a target position of the target horn cover; if the pixel accuracy requirements are not met, fine-tuning mechanical equipment based on fine-tuning parameters and determining motion adjustment parameters based on the current pixel position of the target horn cover, the pixel position of the target priori frame, the initial position, and fine-tuning data; and accurately positioning the target horn cover based on the motion adjustment parameters to obtain the target position of the target horn cover. The present invention can accurately position the horn cover, providing conditions for non-destructive testing.