294 results about "Reactor pressure vessel" patented technology

The invention discloses a reactor pressure vessel fatigue analysis method in a marine environment, which belongs to the technical field of pressure vessel analysis, and comprises the following steps: calculating a transfer function of a structure under the action of a wave load through a spectrum analysis method, and obtaining a stress component power spectrum density function of a node based on a sea condition parameter; converting the frequency domain signal into a time domain signal; calculating a stress time history under the action of the thermal-internal pressure load and carrying out superposition of stress component time histories to obtain a fatigue stress value of the structure; and calculating the fatigue damage value of each stress cycle, obtaining the fatigue damage of each group of time history, and obtaining the total fatigue damage under the working condition. According to the method, the defects of the existing fatigue analysis method can be overcome, and the fatigue damage of the reactor pressure vessel of the nuclear power station under the combined action of multiple loads in heat, internal pressure and marine environment can be accurately evaluated.

The invention discloses a passive nuclear steam supply system which is characterized in that two ends of a hot-section main pipeline are respectively connected with a reactor pressure vessel and a steam generator, a main pump is vertically connected with the steam generator, two ends of a cold-section main pipeline are respectively connected with the pressure vessel and the main pump, the hot-section main pipeline is provided with a first liquid level pressure tap and a second liquid level pressure tap, and the cold-section main pipeline is provided with a second liquid level pressure tap. The first liquid level pressure tap is located at the bottom of the hot section main pipeline, the second liquid level pressure tap is close to the steam generator and located at the top of the hot section main pipeline, the reactor coolant system further comprises a pressure stabilizer and a surge pipe, one end of the surge pipe is connected to one of the hot section main pipelines, and the other end of the surge pipe is connected to the bottom of the pressure stabilizer. The passive nuclear steam supply system can allow the rated power of the reactor core to be remarkably improved.

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.

A reactor body structure and a reactor system. The reactor body structure (100) comprises: a reactor pressure vessel (1), the reactor pressure vessel (1) comprising a bottom closure head (3), and the bottom closure head (3) having an arc-shaped inner wall (32); a reactor internal (2), the reactor internal (2) comprising a reactor core supporting lower plate (22); and a reactor core (21), the reactor core (21) being arranged on the reactor core supporting lower plate (22); wherein the reactor core supporting lower plate (22) is located on the lower side of the center of sphere of the arc-shaped inner wall (32), and the outer edge of the reactor core supporting lower plate (22) is provided with a flow guide corner (221); the reactor core supporting lower plate (22) is provided with through holes (222), the through holes (222) are stepped holes and each comprise a top hole (222a) in the upper side and a bottom hole (222b) in the lower side, and the diameter of the bottom hole (222b) is smaller than that of the top hole (222a).

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 relates to the technical field of reactors, and provides a reactor core outlet heat conduction process equivalent experiment device and method, the experiment device comprises a reactor pressure vessel and a cooling system, the reactor pressure vessel comprises a main body and a thermocouple measurement assembly, and the main body comprises a heat flow cavity, a cold flow cavity and a convection heat exchange cavity; a hot flow outlet is communicated between the hot flow cavity and the convection heat exchange cavity, and a cold flow outlet is communicated between the cold flow cavity and the convection heat exchange cavity; the thermocouple measuring assembly comprises a positioning part and a thermocouple, the positioning part is arranged in the convection heat exchange cavity and corresponds to the heat flow outlet, and the thermocouple can telescopically move in the positioning part in the axial direction so as to adjust the depth of the thermocouple inserted into the heat flow cavity through the heat flow outlet; and the cooling system is circularly connected with the hot flow cavity, the cold flow cavity and the convection heat exchange cavity respectively. According to the invention, the relation between the thermocouple insertion depth and the measurement temperature deviation under different boundary conditions can be researched, and more reliable technical support is provided for safe operation of the nuclear power station.

The invention relates to a liquid level measurement and correction method for a reactor pressure vessel. The method comprises the following steps: dividing the pressure vessel into a plurality of measurement intervals along a pressure difference direction; acquiring first measurement data of the pressure vessel, calculating the initial measurement liquid level height of the pressure vessel according to the first measurement data, assigning the initial measurement liquid level height to the variable liquid level height to be corrected, judging the measurement interval and correcting; according to the corrected liquid level height and the to-be-corrected liquid level height before correction, the measurement intervals where the corrected liquid level height and the to-be-corrected liquid level height are located are compared, when the corrected liquid level height and the to-be-corrected liquid level height belong to different measurement intervals, the to-be-corrected liquid level height is replaced with the current corrected liquid level height, and the latest to-be-corrected liquid level height continues to be corrected; when the latest corrected liquid level height and the latest to-be-corrected liquid level height are located in the same measurement interval, the latest corrected liquid level height is used as the liquid level height of the coolant in the pressure container; on-line correction of the liquid level of the coolant in the pressure container can be achieved, and the liquid level characterization precision is improved.

The invention provides a nuclear power plant reactor pressure vessel connecting pipe inner fillet detection device which comprises a connecting pipe, a driving part, a universal adjusting probe frame and a connecting fork, the front end of the connecting pipe and the connecting fork are connected to form a Y-shaped symmetrical structure, the driving part is installed on the outer side of the connecting fork, and the universal adjusting probe frame is installed on the driving part. The driving piece is connected with the universal adjusting probe frame to drive the universal adjusting probe frame to be attached to the inner fillet area of the connecting pipe. The device for detecting the inner fillet of the connecting pipe of the reactor pressure vessel of the nuclear power plant, provided by the invention, is better suitable for ultrasonic nondestructive detection of the inner fillet of the connecting pipe of the pressure vessel, and the accuracy and the reliability of detection are improved.

The invention discloses reactor pressure vessel inspection equipment and a reactor pressure vessel inspection method. The reactor pressure vessel inspection equipment comprises a supporting frame body, a lifting mechanism installed on the supporting frame body, a rotating mechanism installed on the lower portion, facing a barrel, of the lifting mechanism, a connecting pipe scanning arm installed below the rotating mechanism and two multi-degree-of-freedom mechanical arm probe frames. The lifting mechanism can move relative to the supporting frame body and drives the rotating mechanism and the connecting pipe scanning arm and the multi-degree-of-freedom mechanical arm probe frame on the rotating mechanism to move up and down along the central axis of the cylinder body. And the rotating mechanism drives the connecting pipe scanning arm and the multi-degree-of-freedom mechanical arm probe frame on the rotating mechanism to rotate. According to the invention, the two multi-degree-of-freedom mechanical arm probe frames are matched with the connecting pipe scanning arm to scan the reactor pressure vessel, and the multi-degree-of-freedom mechanical arm probe frames and the connecting pipe scanning arm can scan synchronously in parallel, so that the scanning time is shortened, the working efficiency is improved, and the time for overhauling a key path of a nuclear power plant is greatly shortened.

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.

The invention discloses a direct injection safety system for a reactor pressure vessel. The direct injection safety system comprises a first water injection pipeline and / or a second water injection pipeline, the first water injection pipeline penetrates through an upper seal head of the reactor pressure vessel, the input end of the first water injection pipeline is located outside the reactor pressure vessel and connected to a coolant source, and the output end of the first water injection pipeline is located in the reactor pressure vessel and penetrates through the reactor core upper supporting plate to extend to a reactor core area; the output end of the first water injection pipeline can be directly positioned to a reactor core area to output a coolant, so that the reactor core cooling efficiency is improved; the second water injection pipeline penetrates through the reactor pressure vessel, the input end of the second water injection pipeline is located outside the reactor pressure vessel and connected to the coolant source, the output end of the second water injection pipeline is located in the reactor pressure vessel and at least extends to the descending section of the reactor pressure vessel, and the output end of the second water injection pipeline can directly output coolant from the descending section. And more coolants are promoted to flow to the reactor core area from the reactor core lower support plate, so that the function of cooling the reactor core is achieved.

To disassemble a reactor pressure vessel in a state covered with a heat shield wall.SOLUTION: A method for disassembling a reactor pressure vessel, capable of disassembling a reactor pressure vessel including a nozzle connected to a connection pipe in a state of covering a periphery with a heat shield wall comprises of: a step of cutting a connection pipe along an outer periphery of the heat shield wall; a step of forming a cut piece without having an uncut part of the connection pipe having a size capable of conveying the reactor pressure vessel and a cut piece having the uncut part of the connection pipe; a first removal step of removing the cut piece without having the uncut part of the connection pipe from a reactor container; a step of pulling the uncut part from the heat shield wall by moving the cut piece having the uncut part of the connection pipe in a space direction after removing the cut piece having no uncut part; and a second removal step of removing the cut piece having the uncut part of the connection pipe from the reactor container.SELECTED DRAWING: Figure 4

The invention discloses a reactor pressure vessel kingbolt ultrasonic inspection device, which comprises a positioning mechanism, a detection mechanism, a circumferential movement mechanism and an axial movement mechanism, and is characterized in that the positioning mechanism is used for positioning and clamping a kingbolt, the detection mechanism is used for detecting the kingbolt, and the axial movement mechanism is connected to the detection mechanism; the circumferential movement mechanism is connected to the positioning mechanism and the axial movement mechanism at the same time, and the circumferential movement mechanism is used for driving the axial movement mechanism to rotate in the circumferential direction. And the detection mechanism can spirally move relative to the main bolt under the combined action of the circumferential movement mechanism and the axial movement mechanism. The ultrasonic inspection device for the kingbolt of the reactor pressure vessel can perform spiral scanning on the kingbolt, and can perform comprehensive scanning on a thread area and a polish rod area of the kingbolt, so that defects such as cracks, stress corrosion, collision traces and scratches on the surface of the kingbolt can be found and displayed in time, and the inspection efficiency is greatly improved.

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 production process of a C-shaped sealing ring of a reactor pressure vessel, which comprises the following steps: preparing a silver plate and a nickel-based alloy plate for cutting off, cleaning the cut-off plate for use, finely cutting off the plate again after the plate is cleaned, re-welding the cut-off plate, and finishing a welding seam of the re-welded plate. After finishing is completed, all the plates are subjected to integrity inspection and prefabricated forming, after plate treatment is completed, a spring is wound and subjected to heat treatment, the spring is cut, a notch is finished, and the finished spring is welded; according to the invention, the line tightness and the sealing width are increased, a high-temperature test bench is developed, comprehensive performance test under 35 MPa and 600 DEG C can be completed, technological equipment required by production of the large-diameter sealing ring is developed, the sealing ring does not have any leakage and abnormal phenomena in the experiment process of the sealing ring, and after the test, the sealing ring is not damaged. And the sealing silver layer has no crack, folding, shedding and peeling phenomena.

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.

The invention discloses a reactor chamber water injection cooling system and method.The outer wall face of a reactor pressure vessel is provided with an outer-layer heat preservation vessel, a sealed chamber is formed between the reactor pressure vessel and the outer-layer heat preservation vessel, and the reactor chamber water injection cooling system comprises two rows of circulation loops; each of the two columns of circulation loops comprises a high-position cooling water tank, a water injection valve group and a first water collecting device, an outlet of the first water collecting device is communicated with an inlet of the high-position cooling water tank, and a water source outside the containment is communicated with the high-position cooling water tank; a bottom outlet of the high-level cooling water tank is communicated with an inlet of a cavity between the reactor pressure vessel and the outer-layer heat preservation vessel, and an outlet of the cavity between the reactor pressure vessel and the outer-layer heat preservation vessel is communicated with a sprayer in the high-level cooling water tank; and the outlet of the second water collecting device is communicated with the inlet of the reactor cabin through the water tank in the containment vessel. The system and the method can meet the long-term cooling in a design basis exceeding accident.

The invention discloses an ultrasonic phased array reference test block and a use method thereof. The ultrasonic phased array reference test block comprises a test block body, a surfacing layer, a transverse hole group, a first notch groove group and a second notch groove group, the first side surface of the test block body is provided with a convex part for simulating a boss at the orifice of a connecting pipe, the upper surface of the test block body is provided with a slope, and the lower surface of the test block body is provided with a concave inner bend angle; the surfacing layer is arranged on the upper surface and the first side surface of the test block body, the transverse hole group is arranged on the first surface of the test block body, the first notch groove group is arranged at the joint of the first side surface and the second surface, and the second notch groove group is arranged on the inner corner. According to the ultrasonic phased array reference test block, the surfacing layer, the slope, the inner bend and other structures are arranged according to the characteristics of a reactor pressure vessel barrel and a connecting pipe welding seam, the real structural characteristics of the reactor pressure vessel barrel and the connecting pipe welding seam are effectively simulated, and the influence of a complex structure on a phased array ultrasonic detection process is simulated.

The invention discloses a heat preservation structure of a reactor pressure vessel and a reactor pit water injection system. The heat preservation structure of the reactor pressure vessel comprises a heat preservation assembly which is provided with a first mounting part and a second mounting part; the first supporting piece is suitable for being connected with a supporting ring arranged on the periphery of the reactor pit, and the first supporting piece is configured to bear the first mounting part when the heat preservation assembly stretches into the reactor pit in the preset direction; and the second supporting piece is suitable for being arranged in the reactor pit, and the second supporting piece is configured to center the heat preservation assembly when the heat preservation assembly stretches into the reactor pit in the preset direction and limit the second mounting part in the circumferential direction of the heat preservation assembly. The heat preservation assembly, the first mounting part and the second mounting part are preassembled into a whole; hoisting and positioning installation work is integrated in one step, so that side wall welding and adjusting work possibly needed in a traditional method is omitted; the construction difficulty can be reduced, the construction time is shortened, and the overall construction efficiency is improved; therefore, the installation and adjustment period in a pile pit is shortened.

The invention provides an integrated reactor pressure vessel and a refueling method, the pressure vessel comprises a cylinder and a bottom seal head arranged at the bottom of the cylinder, and the bottom seal head and the cylinder are detachably connected through a flange; at least one supporting boss is arranged in the bottom sealing head, the supporting boss is used for supporting a reactor core supporting plate and a reactor core shroud mounted on the reactor core supporting plate, and the fuel assembly is arranged in the reactor core shroud; in the refueling and refueling period, the bottom sealing head, the reactor core supporting plate, the reactor core shroud and the fuel assembly are transferred to a refueling pool together for fuel disassembly operation by disassembling the bottom sealing head. According to the pressure vessel, the bottom sealing head is detachably arranged at the bottom of the barrel, refueling is carried out from the bottom of the barrel, and the bottom sealing head, the reactor core supporting plate and the fuel assembly are transferred to the refueling pool together for centralized refueling during refueling, so that the refueling period is shortened, and due to refueling at the bottom of the barrel, the refueling efficiency is improved. Therefore, the problems that a traditional reactor top penetrating piece is difficult and tedious to disassemble and assemble, and long-distance deep reactor core reloading is difficult are solved.

The invention provides a protection method, device, equipment and medium for a nuclear power plant reactor, and the method comprises the following steps: obtaining water level data in a reactor pressure vessel in real time, and sending a low water level signal when the water level data is lower than a preset low water level; a protection signal is sent out according to the low water level signal; based on the protection signal, a safety injection system supplements coolant into the reactor pressure vessel and initiates a blowdown isolation function of the chemical and volume control system. By means of the protection method, device and equipment for the nuclear power plant reactor and the medium, water can be supplemented to the reactor core under the condition that a reactor protection system fails.

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.

The invention provides a reactor pressure vessel top cover welding seam visual inspection device, which comprises a camera device, a telescopic adjusting device and a mobile trolley platform, and is characterized in that the camera device comprises a detection camera device and an auxiliary camera device; the detection camera device is arranged at the top of the telescopic adjusting device and is used for detecting a welding seam; the auxiliary camera device is arranged on a trolley body of the mobile trolley platform and is used for observing the visual field of the mobile trolley platform in the moving process; the telescopic adjusting device is connected to the top of the moving trolley platform and can adjust the height and the horizontal position of the detection camera device relative to the moving trolley platform. The visual inspection device for the welding seam of the top cover of the reactor pressure vessel can flexibly move in a shielding ring of the top cover, can reach a specified area, realizes full-coverage inspection, can realize full-coverage visual inspection on the interior of the top cover, and has the advantages of high flexibility, high inspection efficiency and the like.

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

The invention discloses a method and a system for evaluating the service life of a reactor pressure vessel. Relates to the technical field of reactor pressure vessel management. The method is improved on the basis of a traditional reactor pressure vessel service life estimation technology, and an irradiation reference sample and an auxiliary irradiation original sample which belong to the same reactor pressure vessel to be evaluated are constructed while an irradiation original sample is constructed; the service life of the reactor pressure vessel to be evaluated is evaluated on the basis of the irradiation reference sample, the irradiation original sample and the auxiliary irradiation original sample, and the service life of the reactor pressure vessel to be evaluated is evaluated from multiple dimensions; the angle between the original sample and the reference sample in the reactor pressure vessel to be evaluated and the angle between the original sample in the reactor pressure vessel to be evaluated and the reference samples in other reactor pressure vessels can improve the accuracy of the service life evaluation result of the reactor pressure vessel.

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.