CN109243641B - Reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor - Google Patents

Abstract

The invention discloses a reactor pressure vessel experiment simulator for loss of coolant accident of a pressurized water reactor, which comprises a reactor pressure vessel experiment simulator body and a reactor core detection device, wherein the reactor pressure vessel experiment simulator body comprises a pressure vessel simulator body and a hanging flange simulator body positioned in the inner cavity of the pressure vessel simulator body, a reactor core assembly simulator body is arranged in the hanging flange simulator body, the pressure vessel simulator body comprises an upper end enclosure, a middle cylindrical barrel and a lower end enclosure, the reactor core assembly simulator body comprises a cladding tube and an electric heating element, the top of the cladding tube is a closed end, the bottom of the cladding tube is an open end, the bottom of the cladding tube extends downwards and penetrates out of the lower end enclosure, and the data acquisition end of the reactor core detection device is inserted into the cladding tube from the open end of the cladding tube. The invention designs a reactor pressure vessel experiment simulator for loss of coolant accident of a pressurized water reactor, and a detection device for the reactor core component simulator can not influence the sealing performance between a reactor pressure vessel cylinder and a reactor core due to an acquisition path.

Description

Reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor

Technical Field

The invention relates to the field of nuclear power, in particular to a reactor pressure vessel experiment simulator for a loss of coolant accident of a pressurized water reactor.

Background

The loss of coolant accident refers to an accident that a pressure boundary of a reactor main loop is cracked or broken, and a part or most of coolant leaks. The loss-of-coolant accident is a critical research content in nuclear reactor safety, is one of design benchmark accidents of a nuclear power plant, has high accident occurrence probability, and relates to complex thermal hydraulic phenomena, such as a breach critical flow and a reduction of reactor core cooling effect when the reactor core liquid level collapses; even the core is exposed and is in an insufficient cooling process, so that the temperature of the fuel cladding is increased, and the thermal safety of the core is influenced. After a loss of coolant accident occurs, the safety system of the nuclear reactor starts to inject water into the reactor core to submerge the reactor core again, and the nuclear reaction fission gas in the cladding is prevented from being released and even the reactor core is burnt out due to the fact that the fuel cladding is broken after the temperature of the fuel cladding further rises to reach the melting point.

The research and development aim of the advanced nuclear reactor is to reduce the reactor core melting probability, continuously improve the safety performance of the reactor and realize the safe use of the nuclear energy by designing and optimizing a safety system. Therefore, mastering the reactor core safety characteristics obtained by carrying out the nuclear reactor thermal hydraulic experiment under the condition of the loss of coolant accident and verifying the design scheme of the advanced nuclear reactor safety system are important for the research and development of the advanced nuclear reactor system.

In order to measure the change rule of parameters such as temperature, pressure, liquid level and the like of a coolant in a reactor pressure vessel and the safety characteristics of a reactor fuel element when a nuclear reactor has a loss of coolant accident, a reactor pressure vessel experiment simulator is designed. The reactor pressure vessel experiment simulator comprises a pressure vessel simulator for simulating a prototype reactor pressure vessel and a hanging flange simulator positioned in an inner cavity of the pressure vessel simulator, wherein a reactor core assembly simulator is arranged in the hanging flange simulator and comprises an upper end socket, a middle cylindrical barrel and a lower end socket which are welded in sequence, the reactor core assembly simulator comprises a reactor core fuel element simulator and a positioning grid for fixing the reactor core fuel element simulators, and the reactor core fuel element simulator comprises an electric heating element and a cladding pipe. The electric heating element mainly simulates the nuclear fission reaction heat of a fuel element of the nuclear reactor, and different nuclear powers are simulated by controlling the heating power of the electric heating element. The outer diameter of the electric heating element is consistent with the outer diameter of the uranium dioxide pellet of the nuclear reactor fuel element, and the heating length and the heating position are also the same as those of the prototype. The electric heating element is arranged in a cladding tube which passes through the basket analogue. The hanging basket simulator has two main functions: firstly, fixing and supporting a reactor core; and secondly, isolation and flow guide effects. The coolant entering the reactor pressure vessel simulator from the primary loop flows downwards along the outer wall of the hanging basket simulator, and the reactor core is turned by 180 degrees and then is carried out from the lower part of the reactor core, so that the prototype reactor is realized. The reactor core detection device and the pressure vessel detection device are arranged in the reactor pressure vessel experiment simulator, so that corresponding pressure measuring components, temperature measuring components, liquid level measuring components and the like are arranged at relevant positions, the change rule of the parameters such as temperature, pressure and liquid level inside the reactor pressure vessel simulator and the safety characteristic parameters of the reactor fuel elements are collected, and an experimenter can accurately control relevant parameter information inside the reactor pressure vessel simulator.

In the designed reactor pressure vessel experiment simulator, in the adopted temperature measurement, pressure measurement and liquid level detection structures, the signal acquisition ends of the detection devices for the reactor core component simulator all need to penetrate through the reactor pressure vessel cylinder, the hanging basket simulator and the reactor core fuel element simulator to measure the temperature of the coolant in the reactor core. In order to prevent fluid in the reactor pressure vessel barrel from leaking out of the pressure vessel or flowing into the reactor core in series, high-temperature and high-pressure sealing needs to be carried out on through hole parts which are convenient for a detection device to pass through and are arranged on the reactor pressure vessel barrel and the hanging basket simulator. However, since the nacelle simulator is located inside the reactor pressure vessel cylinder, in terms of overall assembly, the collection portion of the collection device sequentially passes through the reactor pressure vessel cylinder, the nacelle simulator and the reactor core fuel element simulator, and then the corresponding portion is sealed, and due to structural limitation, the sealing device at the corresponding portion is mounted and unchanged, and the sealing cost is high.

Disclosure of Invention

The invention aims to provide a reactor pressure vessel experiment simulator for a pressurized water reactor loss of coolant accident, which solves the problems that high-temperature and high-pressure sealing needs to be carried out on through hole parts which are convenient for a detection device to pass through on a reactor pressure vessel cylinder body and a hanging basket simulator, but due to structural limitation, sealing devices on corresponding parts are mounted and unchanged, and the sealing cost is high. Therefore, the reactor pressure vessel experiment simulator for the pressurized water reactor loss of coolant accident is designed, the loss of coolant accident of a nuclear reactor can be accurately measured, the change rule of parameters such as temperature, pressure and liquid level of a coolant in the reactor pressure vessel and the safety characteristic of a reactor fuel element are acquired, meanwhile, the detection device for the reactor core assembly simulator does not influence the sealing performance between the reactor pressure vessel cylinder and the reactor core due to the acquisition path, a sealing device does not need to be installed on the detection device for the reactor core assembly simulator, and the sealing cost is reduced.

The invention is realized by the following technical scheme:

the reactor pressure vessel experiment simulator for loss of coolant accident of the pressurized water reactor comprises a reactor pressure vessel experiment simulator body and a reactor core detection device, wherein the reactor pressure vessel experiment simulator body comprises a pressure vessel simulator body and a hanging flange simulator body positioned in an inner cavity of the pressure vessel simulator body, the reactor core assembly simulator body is arranged in the hanging flange simulator body and comprises an upper end enclosure, a middle cylindrical barrel and a lower end enclosure which are welded in sequence, the reactor core assembly simulator body comprises a cladding tube, an electric heating element arranged in the cladding tube and a positioning grid frame used for fixing each cladding tube, the reactor core detection device is used for collecting parameter information inside the reactor core assembly simulator during experiment, the top of the cladding tube is a closed end, the bottom of the cladding tube is an open end, the bottom of the cladding tube extends downwards and penetrates out of the lower end enclosure, and the cladding tube is hermetically welded with a through hole which is convenient for the bottom of the cladding tube to penetrate out of the lower end enclosure tube, the data acquisition end of the core testing device is inserted into the cladding tube from the open end of the cladding tube.

The reactor pressure vessel cylinder body is composed of an upper end enclosure, a middle cylindrical cylinder body and a lower end enclosure which are welded in sequence.

By adopting the traditional reactor core detection device, the temperature and the pressure of the coolant in the reactor core and the liquid level of the coolant can be measured only by passing through the reactor pressure vessel cylinder, the hanging basket simulator and the reactor core fuel element simulator. In order to prevent the fluid in the reactor pressure vessel cylinder from leaking out of the pressure vessel or flowing into the reactor core in series, high-temperature and high-pressure sealing needs to be performed on two parts, which are penetrated by the reactor core detection device, of the reactor pressure vessel cylinder and the nacelle simulator. The reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor designed by the invention sets the bottom of the cladding pipe of the simulator of the reactor core fuel element in the reactor core as an open end, and extends downwards to penetrate out of the lower end enclosure, and simultaneously, the cladding tube is hermetically welded with a through hole which is convenient for the bottom of the cladding tube to penetrate out of the lower end enclosure, the temperature, pressure, coolant level and other parameter acquisition ends of the reactor core detection device are inserted into the cladding tube from the opening end to acquire corresponding parameters inside the cladding tube, thereby the reactor core detection device does not need to penetrate through the reactor pressure vessel cylinder and the hanging basket simulator, thereby avoiding the occurrence of adverse conditions such as sealing failure of the penetrating part and the like, but the reactor pressure vessel cylinder and the hanging basket analog body, and the hanging basket analog body and the cladding tube are all independent sealed chambers, thereby preventing fluid in the reactor pressure vessel barrel from leaking out of the pressure vessel or streaming into the core.

In conclusion, the reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor is designed, so that the loss of coolant accident of the nuclear reactor can be accurately measured, the change rule of parameters such as temperature, pressure and liquid level of a coolant in the reactor pressure vessel and the safety characteristic of a reactor fuel element are acquired, meanwhile, the detection device for the reactor core assembly simulator does not influence the sealing performance between the reactor pressure vessel cylinder and the reactor core due to the acquisition path, a sealing device does not need to be installed on the detection device for the reactor core assembly simulator, and the sealing cost is reduced.

Furthermore, the reactor core detection device comprises a reactor core internal coolant temperature measurement assembly, the temperature measurement end of a reactor core thermocouple of the reactor core internal coolant temperature measurement assembly is inserted into the reactor core from the opening end of the cladding tube, the tail end of the temperature measurement end penetrates out of the tube wall of the cladding tube, the penetrating length is 0.1-5 mm, the number of the reactor core thermocouples in the same cladding tube is multiple, and the positions of the collected temperatures are distributed along the axis of the cladding tube.

The tail end of the temperature measuring end penetrates out of the tube wall of the cladding tube, so that the part, through which the temperature measuring end penetrates out, is conveniently welded on the cladding tube, and the temperature measuring end is fixed.

Further, the reactor core detection device includes coolant pressure measurement subassembly in the reactor core, the signal receiving terminal of intelligent pressure transmitter is inserted to the one end of the reactor core pressure tube I of coolant pressure measurement subassembly in the reactor core, and the other end of reactor core pressure tube I is inserted into from the open end of cladding pipe, and the end of this end is worn out from the pipe wall of cladding pipe, and it is 0.1 ~ 5mm to wear out length.

In a similar way, the tail end of the reactor core pressure leading pipe I penetrates out of the pipe wall of the cladding pipe, and the part, which penetrates out of the reactor core pressure leading pipe I, is conveniently welded on the cladding pipe so as to fix the temperature measuring end.

Further, the reactor core detection device comprises a reactor core inner liquid level measurement assembly, the reactor core inner liquid level measurement assembly comprises two sets of reactor core pressure leading pipes II, one ends of the reactor core pressure leading pipes II are connected to a signal receiving end of the intelligent pressure transmitter, the other ends of the reactor core pressure leading pipes II are inserted into the reactor core pressure leading pipes II from the opening ends of the cladding pipes, the tail ends of the reactor core pressure leading pipes penetrate out of the pipe walls of the cladding pipes, the penetrating lengths of the tail ends of the reactor core pressure leading pipes are 0.1-5 mm, and the reactor core pressure leading pipes II measure pressure information of two axial heights in the same cladding pipe.

Level measurement subassembly and in-core coolant pressure measurement subassembly's structure is similar in the reactor core, level measurement subassembly is by two sets of and in-core coolant pressure measurement subassembly barrel structure in the reactor core, the detecting element of function constitutes, two sets of reactor cores pressure pipe II have promptly, one side that is close to the top on a set of lateral wall from the cladding pipe is worn out, another group wears out from one side that is close to the open end on the lateral wall of cladding pipe, two sets of reactor cores pressure pipe II measure the pressure of two positions departments, with two sets of reactor cores pressure pipe II access intelligent differential pressure transmitter, convert according to intelligent differential pressure transmitter's signal, obtain reactor core level measurement result.

The tail end of the reactor core pressure leading pipe II penetrates out of the pipe wall of the cladding pipe, and the part, through which the reactor core pressure leading pipe II penetrates out, is conveniently welded on the cladding pipe so as to fix the temperature measuring end.

The pressure measuring assembly for the coolant in the reactor core and the liquid level measuring assembly in the reactor core successfully solve the problem that the pressure and the liquid level of the coolant in the reactor core cannot be measured in the previous thermal hydraulic experiment of the nuclear reactor, and the pressure and the liquid level parameters of the coolant in the reactor core can be accurately measured when a loss of coolant accident occurs to the nuclear reactor by installing the pressure measuring pulse tube on the empty cladding tube of the reactor core fuel element simulator.

Further, the core detecting device also comprises a fuel element wall temperature measuring assembly, the fuel element wall temperature measuring assembly comprises two wall temperature thermocouples, the temperature collecting ends of the wall temperature thermocouples are inserted into the cladding tube from the open end of the cladding tube, one of the temperature collecting ends is fixed on the inner wall of the cladding tube, the other temperature collecting end is fixed on the outer wall of the electric heating element, and the collecting ends of the two wall temperature thermocouples are positioned in the same radial direction of the cladding tube.

The fuel element wall temperature measuring assembly is welded on the wall surfaces of the cladding tube and the electric heating element, and can accurately measure the wall temperature change condition of the reactor core fuel element when a nuclear reactor loss of coolant accident occurs.

The device further comprises a pressure vessel detection device, the pressure vessel detection device comprises a coolant temperature measurement component in the pressure vessel, a coolant pressure measurement component in the pressure vessel and a coolant liquid level detection component in the pressure vessel, a plurality of sealing components are arranged on the outer wall of the middle cylindrical barrel, the temperature acquisition end of a barrel thermocouple of the coolant temperature measurement component in the pressure vessel, the tail end of a barrel pressure leading pipe I of the coolant pressure measurement component in the pressure vessel and the tail end of a barrel pressure leading pipe II of the coolant liquid level detection component in the pressure vessel respectively penetrate through one sealing component and then are inserted into the inner cavity of the middle cylindrical barrel from the outer wall of the middle cylindrical barrel and are positioned between the middle cylindrical barrel and the hanger simulation body, two groups of barrel pressure leading pipes II of the coolant liquid level detection component in the pressure vessel penetrate through the upper part of the middle cylindrical barrel, the other group penetrates through the lower part of the middle cylindrical barrel, and the sealing assembly seals the part, penetrated by the signal acquisition end of the pressure container detection device, of the middle cylindrical barrel.

The coolant liquid level detection assembly in the pressure vessel is composed of two groups of detection assemblies which are identical to the coolant pressure measurement assembly in the pressure vessel in structure and function, one group of detection assemblies is arranged on the upper portion of the reactor pressure vessel barrel, one group of detection assemblies is arranged on the lower portion of the reactor pressure vessel barrel, the barrel pressure leading pipe II is connected into the intelligent differential pressure transmitter, conversion is carried out according to signals of the intelligent differential pressure transmitter, and a reactor core liquid level measurement result is obtained.

Furthermore, the sealing assembly matched with the barrel thermocouple comprises a tube seat, a compression nut, a compression rod and a red copper gasket which are coaxial with each other, one end of the tube seat is welded on the outer wall of the middle cylindrical barrel, the other end of the tube seat is recessed into a spherical concave surface, a central hole of the compression nut is a two-stage stepped hole, a large-diameter section of the compression nut is sleeved on one side, far away from the middle cylindrical barrel, of the side wall of the tube seat and is in threaded connection with the tube seat, one end of the compression rod is connected with a pressing block, one end, far away from the compression rod, of the pressing block protrudes outwards to form a spherical convex surface and is matched with the spherical concave surface, the other end of the compression rod penetrates through a small-diameter section of the compression nut, the hole bottom of the large-diameter section is in contact with one end, far away from the tube seat, of the tube seat, one end of a through hole in the tube seat penetrates through the wall of the middle cylindrical barrel, and the other end of the tube seat penetrates through the pressing block and the compression rod in sequence;

the red copper gasket is positioned between the spherical convex surface of the pressing block and the spherical concave surface of the tube seat, and the pressing nut is rotated to enable the pressing block to press on the red copper gasket when the pressing nut moves towards the middle cylindrical barrel, so that the red copper gasket is deformed, one side of the red copper gasket, which faces the tube seat, is internally tangent to the spherical concave surface, and the spherical convex surface is internally tangent to one side of the red copper gasket, which is far away from the tube seat;

the temperature acquisition end of the cylinder thermocouple sequentially passes through the through hole on the pressure rod, the through hole on the pressure block, the red copper gasket and the through hole on the tube seat and is positioned in the inner cavity of the middle cylindrical cylinder after the cylinder wall of the middle cylindrical cylinder.

During installation, the tube seat is welded on the reactor pressure vessel cylinder body firstly, the temperature acquisition end of the cylinder thermocouple enters the reactor pressure vessel cylinder body through the through hole in the middle of the tube seat, the red copper gasket is sleeved on the cylinder thermocouple, then the compression nut is screwed on the tube seat, the pressing block is in contact with the red copper gasket, then the compression nut is screwed, and the red copper gasket is deformed through the pressing block, so that the temperature measurement of high-temperature and high-pressure fluid is realized.

The sealing assembly matched with the barrel pressure guiding pipe I or the barrel pressure guiding pipe II comprises a pipe seat and a compression nut which are coaxial, one end of the pipe seat is welded on the outer wall of the middle cylindrical barrel, the other end of the pipe seat is concaved inwards to form a spherical concave surface, a center hole of the compression nut is a second-stage stepped hole, a large-diameter section of the compression nut is sleeved on one side, far away from the middle cylindrical barrel, of the side wall of the pipe seat and is in threaded connection with the pipe seat, the ends of the barrel pressure guiding pipe I and the barrel pressure guiding pipe II protrude outwards to form a spherical convex surface and are matched with the spherical concave surface, one end, far away from the ends, of the barrel pressure guiding pipe I and the barrel pressure guiding pipe II penetrates through a small-diameter section of the corresponding compression nut, the hole bottom of the large-diameter section is in contact with one end, far away from the pipe seat, and one end of a through hole in the pipe seat penetrates through the barrel wall of the middle cylindrical barrel.

During installation, the tube seat is welded on the reactor pressure vessel barrel firstly, then the compression nut is screwed on the tube seat, the ends of the barrel pressure guiding tube I and the barrel pressure guiding tube II are opposite to the spherical concave surface of the corresponding tube seat, then the compression nut is screwed, the hole bottom of the large-diameter section of the compression nut presses the end to move towards the middle cylindrical barrel, so that the spherical convex surface on the end is in contact with the spherical concave surface on the barrel pressure guiding tube I, the end is compressed on the spherical concave surface, and the compression nut is in a close contact state. The end heads of the cylinder body pressure guiding pipe I and the cylinder body pressure guiding pipe II are spherical sealing surfaces, and the end heads of the cylinder body pressure guiding pipe I and the cylinder body pressure guiding pipe II are in close contact with the spherical sealing surfaces of the corresponding pipe seats through compression nuts, so that the pressure measurement of high-temperature and high-pressure fluid is realized.

The invention discloses a novel reactor pressure vessel experiment simulator under the condition of nuclear reactor loss of coolant accident, which is developed based on the actual requirements of the third generation advanced nuclear reactor research and development and according to the requirements of the safety thermal hydraulic experiment of the nuclear reactor core. The reactor pressure vessel experiment simulator under the nuclear reactor loss of coolant accident condition comprises a reactor pressure vessel simulator, a reactor core fuel element simulator, a hanging basket simulator, a pressure vessel internal coolant temperature measuring assembly, a pressure vessel internal coolant pressure measuring assembly, a pressure vessel internal liquid level measuring assembly, a reactor core internal coolant temperature measuring assembly, a reactor core internal coolant pressure measuring assembly, a reactor core internal collapse liquid level measuring assembly, a fuel element wall temperature measuring assembly and the like. The electronic device involved in the present invention: the temperature measuring assembly for the coolant in the pressure vessel, the pressure measuring assembly for the coolant in the pressure vessel, the liquid level detecting assembly for the coolant in the pressure vessel, the temperature measuring assembly for the coolant in the reactor core, the pressure measuring assembly for the coolant in the reactor core, the liquid level measuring assembly for the liquid level in the reactor core, the wall temperature measuring assembly for the fuel element and the intelligent differential pressure transmitter are all the prior art and can be purchased in the market.

Compared with the prior art, the invention has the following advantages and beneficial effects:

1. the reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor can accurately measure the loss of coolant accident of the nuclear reactor, and acquires the change rule of parameters such as temperature, pressure, liquid level and the like of a coolant in the reactor pressure vessel and the safety characteristic of a reactor fuel element, and meanwhile, the detection device for the reactor core assembly simulator can not influence the sealing performance between the cylinder body of the reactor pressure vessel and the reactor core because of an acquisition path, and a sealing device does not need to be installed on the detection device for the reactor core assembly simulator, so that the sealing cost is reduced.

2. The reactor pressure vessel experiment simulator for the pressurized water reactor loss of coolant accident has mature manufacturing process and lower cost; the experimental simulator successfully solves the problem of measuring key reactor thermal safety parameters such as the temperature, the pressure, the liquid level and the wall temperature of a fuel element of the reactor core coolant of the nuclear reactor under the loss of coolant accident, obtains the change rule of the advanced reactor core thermal parameters under the loss of coolant accident condition, lays a solid foundation for researching and developing an advanced nuclear reactor system, and effectively promotes the research and development process of the advanced nuclear reactor system;

3. the reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor is characterized in that a coolant temperature measuring assembly in the pressure vessel, a coolant pressure measuring assembly in the pressure vessel and a liquid level measuring assembly in the pressure vessel are arranged around a cylinder body of the reactor pressure vessel, so that the change rule of the pressure, the temperature and the liquid level of the coolant in the pressure vessel when the loss of coolant accident occurs to a nuclear reactor can be accurately measured.

4. The reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor solves the problem that the traditional high-pressure temperature measuring assembly cannot measure the temperature of the coolant in the reactor core, and the temperature of the coolant at the top, upper part, middle part and lower part of the fuel element of the reactor core can be accurately measured by mounting the temperature measuring thermocouple on the empty cladding tube of the fuel element simulator of the reactor core when the loss of coolant accident occurs to the nuclear reactor;

5. the reactor pressure vessel experiment simulator, the in-core coolant pressure measuring assembly and the in-core liquid level measuring assembly for the pressurized water reactor loss of coolant accident successfully solve the problem that the pressure and the liquid level of the reactor core coolant in the conventional nuclear reactor thermal hydraulic experiment cannot be measured.

6. The invention relates to a reactor pressure vessel experiment simulator for pressurized water reactor loss of coolant accident, wherein a fuel element wall temperature measuring assembly is welded on the wall surfaces of a cladding tube and an electric heating element, and the wall temperature change condition of a reactor core fuel element when the loss of coolant accident of a nuclear reactor occurs can be accurately measured.

Drawings

The accompanying drawings, which are included to provide a further understanding of the embodiments of the invention and are incorporated in and constitute a part of this application, illustrate embodiment(s) of the invention and together with the description serve to explain the principles of the invention. In the drawings:

FIG. 1 is a schematic structural view of the present invention;

FIG. 2 is a schematic structural diagram of a core assembly simulator;

FIG. 3 is a schematic diagram of a cladding tube mounted core thermocouple configuration;

FIG. 4 is a schematic structural diagram of a cladding tube installation core pressure guiding tube I15;

FIG. 5 is a schematic structural diagram of a cladding tube installation core pressure guiding tube II 16;

FIG. 6 is a schematic view showing the installation position of the wall-temperature thermocouple 17;

FIG. 7 is an enlarged view of circle A of FIG. 1;

fig. 8 is an enlarged view shown by circle B in fig. 1.

Reference numbers and corresponding part names in the drawings:

1-middle cylindrical barrel, 2-reactor core fuel element simulator, 3-chlorophytum simulator, 4-pressure vessel coolant temperature measuring component, 5-pressure vessel coolant pressure measuring component, 6-pressure vessel coolant liquid level detecting component, 7-reactor core coolant temperature measuring component, 8-reactor core coolant pressure measuring component, 9-reactor core liquid level measuring component, 10-fuel element wall temperature measuring component, 11-electric heating element, 12-cladding tube, 13-location grid, 14-reactor core thermocouple, 15-reactor core pressure leading tube I, 16-reactor core pressure leading tube II, 17-wall temperature thermocouple, 18-upper head, 19-lower head, 20-barrel thermocouple, 21A-tube seat, 22A-gland nut, 21B-a pipe seat, 22B-a compression nut, 23-a compression bar, 24-a pressing block, 25-a red copper gasket, 26-a cylinder body pressure guiding pipe I, 27-a cylinder body pressure guiding pipe II, 28-end heads.

Detailed Description

In order to make the objects, technical solutions and advantages of the present invention more apparent, the present invention is further described in detail below with reference to examples and accompanying drawings, and the exemplary embodiments and descriptions thereof are only used for explaining the present invention and are not meant to limit the present invention.

Example 1

As shown in fig. 1-8, the reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor of the present invention comprises a reactor pressure vessel experimental simulator body and a reactor core detection device, wherein the reactor pressure vessel experimental simulator body comprises a pressure vessel simulator body and a hanging flange simulator body 3 located in an inner cavity of the pressure vessel simulator body, a reactor core assembly simulator body is arranged in the hanging flange simulator body, the pressure vessel simulator body comprises an upper end enclosure 18, a middle cylindrical barrel 1 and a lower end enclosure 19 which are welded in sequence, the reactor core assembly simulator body comprises a cladding tube 12, an electric heating element 11 disposed in the cladding tube 12 and a positioning grid 13 for fixing each cladding tube 12, the reactor core detection device is used for collecting parameter information inside the reactor core assembly simulator during experiment, the top of the cladding tube 12 is a closed end, the bottom of the cladding tube 12 is a shell tube, and the bottom of the cladding tube 12 extends downwards and penetrates through the lower end enclosure 19, the cladding tube 12 is hermetically welded with a through hole on the lower end enclosure 19, which is convenient for the bottom of the cladding tube 12 to penetrate out, and the data acquisition end of the reactor core detection device is inserted into the cladding tube 12 from the open end of the cladding tube 12.

As shown in figure 1, the reactor pressure vessel cylinder body is composed of an upper seal head 18, a lower seal head 19 and a middle cylinder body 1 through welding. The sizes of the upper seal head 18, the lower seal head 19 and the middle cylindrical barrel 1 are subjected to scale reduction and simulation according to the size of the prototype reactor pressure vessel. The prototype reactor head 18 is arranged with control rod drive mechanisms for controlling the nuclear power of the reactor. The nuclear reactor thermal hydraulic experiment generally adopts electric heating power to simulate nuclear heat release, so a control rod driving mechanism is not simulated in the nuclear reactor thermal hydraulic experiment. The reactor core nuclear measurement assembly is arranged on the lower end socket of the prototype reactor and used for measuring reactor core nuclear data of the reactor, and the simulation reactor core nuclear measurement assembly is not required to be considered in a thermal hydraulic experiment of the reactor. The operating pressure of the nuclear reactor is generally 15MPa, and the operating temperature is about 340 ℃. In order to truly simulate the physical phenomena and characteristics in the reactor pressure vessel when the nuclear reactor has an accident, the design pressure and the design temperature of the reactor pressure vessel cylinder are kept consistent with those of a prototype. The reactor core fuel element simulator 2 is an important component of the reactor pressure vessel simulator and mainly simulates nuclear heat release of a nuclear reactor fuel element. Electrical heating is typically used to simulate the heat emitted by nuclear reactor fuel elements, taking into account nuclear radiation and research costs. As shown in fig. 2, the core fuel element simulator 2 is mainly composed of an electric heating element 11, a cladding tube 12, and a spacer 13. The electric heating element 11 mainly simulates the nuclear fission reaction heat of the fuel element of the nuclear reactor, and different nuclear powers are simulated by controlling the heating power of the electric heating element 11. The outer diameter of the electric heating element 11 is consistent with the outer diameter of the uranium dioxide pellet of the nuclear reactor fuel element, and the heating length and the heating position are also the same as those of the prototype. An electric heating element 11 is arranged in the cladding tube 12. The cladding tube 12 is closed at one end and open at the other. The closed end of the cladding tube 12 penetrates through the hanging basket simulator body 3 from a mounting hole at the joint of the reactor pressure vessel cylinder and the lower end socket 19 to enter the reactor pressure vessel cylinder, and the open end of the cladding tube 12 is arranged outside the reactor pressure vessel cylinder. After the depth of the cladding tube 12 is determined, the cladding tube 12 is welded to the side of the reactor pressure vessel barrel near the lower end cap 19. After the cladding tube 12 is welded, the data acquisition end of the core detection device is inserted into the cladding tube 12 from the open end of the cladding tube 12, and then the heating section of the electric heating element 11 is inserted into the cladding tube 12 from the open end thereof. The spacer grid 13 is the same as the fuel element of the prototype reactor core, and mainly functions to fix the electric heating element 11 and simulate the flow field in the prototype reactor. The hanging basket simulator 3 has two main functions, namely: fixing and supporting the core; II, secondly: under the action of isolation and flow guide, coolant entering the reactor pressure vessel from a loop flows downwards along the outer wall of the hanging basket simulator 3, and the reactor core is turned for 180 degrees and then is carried out from the lower part of the reactor core, so that the prototype reactor is realized.

By adopting the traditional reactor core detection device, the temperature and the pressure of the coolant in the reactor core and the liquid level of the coolant can be measured only by passing through the reactor pressure vessel cylinder, the hanging basket simulator 3 and the reactor core fuel element simulator. In order to prevent the fluid in the reactor pressure vessel barrel from leaking out of the pressure vessel or flowing into the reactor core in series, high-temperature and high-pressure sealing needs to be performed on two parts of the reactor pressure vessel barrel and the nacelle simulator 3, which are penetrated by the reactor core detection device. The reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor, which is designed by the invention, sets the bottom of the cladding pipe of the reactor core fuel element simulator in the reactor core as an open end and extends downwards to penetrate out of the lower end enclosure 19, meanwhile, the cladding tube 12 and the through hole which is arranged on the lower end enclosure 19 and is convenient for the bottom of the cladding tube 12 to penetrate out are welded in a sealing way, and the temperature, pressure, coolant level and other parameter acquisition ends of the core detection device are inserted into the cladding tube 12 from the opening end to acquire corresponding parameters inside the cladding tube 12, thereby the reactor core detection device does not need to penetrate through the reactor pressure vessel cylinder and the hanging basket simulator 3, thereby avoiding the occurrence of adverse conditions such as sealing failure of the penetrating part and the like, but the independent sealing chambers are respectively arranged between the reactor pressure vessel cylinder and the hanging basket analog body 3 and between the hanging basket analog body 3 and the cladding tube 12, thereby preventing fluid in the reactor pressure vessel barrel from leaking out of the pressure vessel or streaming into the core.

In conclusion, the reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor is designed, so that the loss of coolant accident of the nuclear reactor can be accurately measured, the change rule of parameters such as temperature, pressure and liquid level of a coolant in the reactor pressure vessel and the safety characteristic of a reactor fuel element are acquired, meanwhile, the detection device for the reactor core assembly simulator does not influence the sealing performance between the reactor pressure vessel cylinder and the reactor core due to the acquisition path, a sealing device does not need to be installed on the detection device for the reactor core assembly simulator, and the sealing cost is reduced.

Example 2

In this embodiment, an embodiment of the core testing apparatus will be further described based on embodiment 1.

As shown in fig. 3, the reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor of the present invention includes a temperature measuring assembly 7 for coolant in the reactor core, wherein the temperature measuring end of a thermocouple 14 of the temperature measuring assembly 7 for coolant in the reactor core is inserted into the open end of a cladding tube 12, the tail end of the temperature measuring end penetrates out of the wall of the cladding tube 12 and has a length of 1mm, a plurality of thermocouples 14 are located in the same cladding tube 12, and the temperature collecting positions are distributed along the axis of the cladding tube 12.

The tail end of the temperature measuring end penetrates out of the tube wall of the cladding tube 12, so that the part, through which the temperature measuring end penetrates out, is conveniently welded on the cladding tube 12, and the temperature measuring end is fixed.

The pressurized water reactor fuel assembly generally takes a 17 x 17 square configuration with 264 fuel rods, 24 control rod guide tubes and 1 instrumentation tube. Both the control rod guide tube and the instrumentation tube do not heat up, and therefore are typically simulated using an empty cladding tube. The in-core coolant temperature measurement assembly 7 measures the temperature of the in-core coolant by using the empty cladding tube in the reactor core fuel element simulator 2. Four core thermocouples 14 are arranged in the same cladding tube 12, and the method for installing the coolant temperature measuring assembly 7 in the core is as follows:

four holes are arranged on the hole wall of the empty cladding tube 12

The holes are arranged in the upper part of the shell,

the holes are evenly distributed on the cladding tube 12 at intervals along the axial direction of the cladding tube 12, and the holes are farthest from the opening end

The hole is located at the top of the cladding tube 12, closest to the open end

The side of the bore on the sidewall of the cladding tube 12 near the open end;

(II) installing the temperature measuring end of a core thermocouple 14 from the open end of the empty cladding tube 12, and then installing the temperature measuring end of the core thermocouple from the open end of the empty cladding tube

The large opening penetrates out, and the penetrating length is 1 mm;

welding the reactor core thermocouple 14 on the empty cladding tube 12 from the outside of the empty cladding tube 12;

and fourthly, sequentially welding other three thermocouples on the empty cladding tube 12 according to the method, wherein the reactor core thermocouples 14 respectively penetrate through one thermocouple

Four core thermocouples 14 measure the temperature of the coolant at the inlet, middle, upper and outlet of the core fuel element, respectively;

and (V) extending the cladding tube 12 provided with the in-core coolant temperature measuring assembly 7 into the core position in the pressure vessel cylinder, and then welding one side of the side wall of the cladding tube 12 close to the lower end enclosure on one side of the middle cylindrical cylinder 1 close to the lower end enclosure.

The problem that the temperature of the coolant in the reactor core cannot be measured by a traditional high-pressure temperature measuring assembly is solved, and the temperature of the coolant at the top, the upper part, the middle part and the lower part of the reactor core fuel element can be accurately measured when a nuclear reactor is in a loss of coolant accident by utilizing the thermocouple for measuring the temperature installed on the empty cladding tube of the reactor core fuel element simulator.

Example 3

In this example, a second embodiment of the core testing apparatus will be further described in addition to example 1.

As shown in fig. 4, the reactor pressure vessel experiment simulator for loss of coolant accident of pressurized water reactor of the present invention includes a pressure measuring assembly 8 for coolant in the reactor core, wherein one end of a pressure tube i 15 for coolant in the reactor core of the pressure measuring assembly 8 is connected to a signal receiving end of an intelligent pressure transmitter, the other end of the pressure tube i 15 for the reactor core is inserted into the opening end of a cladding tube 12, and the end of the pressure tube is extended out of the wall of the cladding tube 12 and is extended out by 1 mm.

In the same way, the tail end of the reactor core pressure leading pipe I15 penetrates out of the pipe wall of the cladding pipe 12, so that the part, through which the reactor core pressure leading pipe I15 penetrates out, is conveniently welded on the cladding pipe 12, and the temperature measuring end is fixed.

The method for installing the in-core coolant pressure measuring assembly 8 comprises the following steps:

one, the empty cladding tube is opened

The aperture of (a);

(II) mixing

The upper part of the reactor core pressure guiding pipe I15 is bent by a 90-degree right angle;

thirdly, one end of the reactor core pressure leading pipe I15, which is provided with a right-angled bend, is arranged from the open end of the empty cladding pipe 12 and penetrates out of the opening of the empty cladding pipe, and the penetrating length is 1 mm;

welding a reactor core pressure leading pipe I15 on the empty cladding tube 12 from the outside of the empty cladding tube 12;

stretching the cladding tube 12 of the coolant pressure measuring assembly 8 in the reactor core into the position of the reactor core in the pressure vessel cylinder, and then welding one side, close to the lower end enclosure, of the side wall of the cladding tube 12 on one side, close to the lower end enclosure, of the middle cylindrical cylinder 1;

(VI),

After the right-angle bend on the reactor core pressure guiding pipe I15 penetrates out of the cladding pipe 12, the tail end of the pressure guiding pipe is welded with the pressure guiding pipe with the diameter changing function

The small-diameter end of the reducing pressure guiding pipe is welded with the tail end of the right-angled bend;

and (seventhly), the downstream of the reactor core pressure leading pipe I15 is connected with an intelligent pressure transmitter for pressure measurement.

Example 4

In this example, a third embodiment of the core testing apparatus will be further described with reference to example 1.

As shown in FIG. 5, the reactor pressure vessel experiment simulator for the loss of coolant accident of the pressurized water reactor comprises a reactor core internal liquid level measuring assembly 9, wherein the reactor core internal liquid level measuring assembly 9 comprises two groups of reactor core pressure guiding pipes II 16, one ends of the reactor core pressure guiding pipes II 16 are connected to a signal receiving end of an intelligent pressure transmitter, the other ends of the reactor core pressure guiding pipes II 16 are inserted into an opening end of a cladding pipe 12, the tail ends of the ends penetrate out of the pipe wall of the cladding pipe 12, the penetrating length of the tail ends is 0.1-5 mm, and the reactor core pressure guiding pipes II 16 measure pressure information of two axial heights in the same cladding pipe 12.

The structure of in-core liquid level measurement subassembly 9 and in-core coolant pressure measurement subassembly 8 is similar, in-core liquid level measurement subassembly 9 by two sets of and 8 tube structures of in-core coolant pressure measurement subassembly, the detection subassembly of function constitutes, two sets of reactor core pressure pipes II 16 have promptly, a set of one side of being close to the top on the lateral wall from cladding pipe 12 is worn out, another group wears out from the one side that is close to the open end on the lateral wall of cladding pipe 12, two sets of reactor core pressure pipes II 16 measure the pressure of two positions departments, press pipe II 16 to insert intelligent differential pressure transmitter with two sets of reactor cores, convert according to intelligent differential pressure transmitter's signal, obtain reactor core liquid level measurement result.

The tail end of the reactor core pressure guiding pipe II 16 penetrates out of the wall of the cladding pipe 12, so that the part, through which the reactor core pressure guiding pipe II 16 penetrates out, is conveniently welded on the cladding pipe 12, and the temperature measuring end is fixed.

Since the in-core liquid level measurement module 9 employs two in-core coolant pressure measurement modules 8, the installation method thereof is the same as the installation method of the in-core coolant pressure measurement module 8 in embodiment 3, except that two in-core coolant pressure measurement modules 8 are provided

Two core pressure guiding pipes II 16 are installed, and the specific installation method is as follows:

two cladding tubes are arranged on an empty cladding tube

One on the side of the side wall of the cladding tube 12 near the top and one on the side of the side wall near the open end;

(II) mixing two

The upper parts of the reactor core pressure guiding pipes II 16 are bent by 90 degrees at right angles;

(III) one end of a reactor core pressure guiding pipe II 16 provided with a right-angled bend is arranged from the open end of the empty cladding pipe 12, and is arranged from one end of the empty cladding pipe

The hole is penetrated out, and the penetrating length is 1 mm;

welding a reactor core pressure leading pipe II 16 on the empty cladding tube 12 from the outside of the empty cladding tube 12;

(V) repeating the step (III) and the step (IV), and fixing one end of the right-angled bend of the other reactor core pressure guiding pipe II 16 on the cladding pipe 12;

stretching the cladding tube 12 of the coolant pressure measuring assembly 8 in the reactor core into the position of the reactor core in the pressure vessel cylinder, and then welding one side, close to the lower end enclosure, of the side wall of the cladding tube 12 on one side, close to the lower end enclosure, of the middle cylindrical cylinder 1;

(seven),

After the right angle bend on the reactor core pressure guiding pipe II 16 penetrates out of the cladding pipe 12, the tail ends of the pressure guiding pipes are welded with pressure guiding pipes with variable diameters

The small-diameter end of the reducing pressure guiding pipe is welded with the tail end of the right-angled bend;

and (seventhly), the downstream of the reactor core pressure guiding pipe II 16 is connected with an intelligent pressure transmitter for pressure measurement, and the conversion is carried out according to the signal of the intelligent differential pressure transmitter to obtain the reactor core liquid level measurement result.

Example 5

In this example, a fourth embodiment of the core testing device will be further described with reference to example 1.

As shown in fig. 6, the reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to the present invention further includes a fuel element wall temperature measuring assembly 10, wherein the fuel element wall temperature measuring assembly 10 includes two wall temperature thermocouples 17, temperature collecting ends of the wall temperature thermocouples 17 are inserted into the cladding tube 12 from an opening end of the cladding tube 12, one of the temperature collecting ends is fixed on an inner wall of the cladding tube 12, the other temperature collecting end is fixed on an outer wall of the electric heating element 11, and the collecting ends of the two wall temperature thermocouples 17 are located in the same radial direction of the cladding tube 12.

The fuel element wall temperature measurement assembly 10 is installed as shown in fig. 6, with the following steps:

cutting off the cladding tube 12 at a position 5mm upstream of a set wall temperature measuring position;

welding a wall temperature thermocouple 17 on the inner wall surface of the cladding tube 12;

thirdly, welding the cut cladding tube 12 along the cut part;

welding another wall temperature thermocouple 17 on the electric heating element 11;

(V) inserting the electric heating element 11 of the wall temperature thermocouple 17 into the cladding tube 12 from the open end of the cladding tube 12;

and (VI) welding the cladding tube 12 on one side of the reactor pressure vessel cylinder close to the lower end socket.

In step (four) of this embodiment, it is necessary to ensure that the temperature-collecting ends of the two wall-temperature thermocouples 17 are located at the same radial direction of the cladding tube 12 when the electric heating element is installed in the cladding tube 12, so that the two wall-temperature thermocouples 17 collect the temperature at the same height position.

The fuel element wall temperature measuring assembly is welded on the wall surfaces of the cladding tube and the electric heating element, and can accurately measure the wall temperature change condition of the reactor core fuel element when a nuclear reactor loss of coolant accident occurs.

Example 6

This example is a combination of example 2 and example 5.

In the experiment, because the core fuel element simulators in the core are distributed in a checkerboard symmetry manner, the in-core coolant temperature measuring assembly 7, the in-core coolant pressure measuring assembly 8, the in-core liquid level measuring assembly 9 and the fuel element wall temperature measuring assembly 10 are arranged in symmetrical positions, and only one position is selected for arrangement, and meanwhile, the core detecting device or the core detecting devices in the embodiments 2 to 5 can be selectively installed in one cladding tube 12.

The figures corresponding to examples 2 to 4 do not show the electric heating element 11.

Example 7

As shown in figure 1, the reactor pressure vessel experimental simulator for the loss of coolant accident of the pressurized water reactor further comprises a pressure vessel detection device, wherein the pressure vessel detection device comprises a coolant temperature measurement component 4 in the pressure vessel, a coolant pressure measurement component 5 in the pressure vessel and a coolant liquid level detection component 6 in the pressure vessel, a plurality of sealing components are arranged on the outer wall of a middle cylindrical barrel 1, the temperature acquisition end of a barrel thermocouple 20 of the coolant temperature measurement component 4 in the pressure vessel, the tail end of a barrel pressure leading pipe I26 of the coolant pressure measurement component 5 in the pressure vessel and the tail end of a barrel pressure leading pipe II 27 of the coolant liquid level detection component 6 in the pressure vessel respectively penetrate through one sealing component, then the sealing components are inserted into the inner cavity of the middle cylindrical barrel 1 from the outer wall of the middle cylindrical barrel 1 and are positioned between the middle cylindrical barrel 1 and a hanger simulator 3, and the cylinder pressure leading pipes II 27 of the coolant liquid level detection assembly 6 in the pressure vessel are divided into two groups, one group penetrates through the upper part of the middle cylinder body 1, the other group penetrates through the lower part of the middle cylinder body 1, and the sealing assembly seals the part, penetrated by the signal acquisition end of the pressure vessel detection device, of the middle cylinder body 1.

The coolant liquid level detection assembly 6 in the pressure vessel is composed of two groups of detection assemblies which are identical to the coolant pressure measurement assembly 5 in the pressure vessel in structure and function, one group of detection assemblies is arranged on the upper portion of the reactor pressure vessel barrel, one group of detection assemblies is arranged on the lower portion of the reactor pressure vessel barrel, the barrel pressure leading pipe II 27 is connected into the intelligent differential pressure transmitter, conversion is carried out according to signals of the intelligent differential pressure transmitter, and a reactor core liquid level measurement result is obtained.

The temperature measuring assemblies 4 for the coolant in the pressure vessel are arranged on the periphery of the reactor pressure vessel barrel and are centrosymmetric along the axis of the reactor pressure vessel barrel so as to measure the temperature of the coolant in the reactor pressure vessel.

Example 8

The present embodiment is a description of the structure of the sealing assembly engaged with the cartridge thermocouple 20.

As shown in fig. 7, the sealing assembly cooperating with the barrel thermocouple 20 includes a tube seat 21A, a gland nut 22A, a pressure bar 23 and a red copper washer 25 coaxial with each other, one end of the tube seat 21A is welded on the outer wall of the middle cylindrical barrel 1, the other end of the tube seat 21A is recessed into a spherical concave surface, the center hole of the gland nut 22A is a two-stage stepped hole, the large diameter section thereof is sleeved on one side of the side wall of the tube seat 21A far from the middle cylindrical barrel 1 and is in threaded connection with the tube seat 21A, one end of the pressure bar 23 is connected with a press block 24, one end of the press block 24 far from the pressure bar 23 protrudes outward into a spherical convex surface and cooperates with the spherical concave surface, the other end of the pressure bar 23 penetrates through the small diameter section of the gland nut 22A, the bottom of the large diameter section is in contact with one end of the press block 24 far from the tube seat 21, one end of the through hole on the tube seat 21A penetrates through the barrel wall of the middle cylindrical barrel 1, the other end of the tube seat 21A sequentially penetrates through the pressing block 24 and the pressing rod 23;

the red copper gasket 25 is positioned between the spherical convex surface of the pressing block 24 and the spherical concave surface of the tube seat 21A, and the pressing nut 22A is rotated to enable the pressing block 24 to press the red copper gasket 25 when the pressing nut 22A moves towards the middle cylindrical barrel 1, so that the red copper gasket 25 is deformed, one side of the red copper gasket 25, which faces the tube seat 21A, is internally tangent to the spherical concave surface, and the spherical convex surface is internally tangent to one side of the red copper gasket 25, which is far away from the tube seat 21A;

the temperature acquisition end of the cylinder thermocouple 20 sequentially passes through the through hole on the pressure rod 23, the through hole on the pressure block 24, the red copper gasket 25, the through hole on the tube seat 21A and the wall of the middle cylinder 1 and then is positioned in the inner cavity of the middle cylinder 1.

During installation, the tube seat 21A is welded on a reactor pressure vessel cylinder, the temperature acquisition end of the cylinder thermocouple 20 enters the reactor pressure vessel cylinder through a through hole in the middle of the tube seat 21A, the copper gasket 25 is sleeved on the cylinder thermocouple 20, the compression nut 22A is screwed on the tube seat 21A, the pressing block 24 is in contact with the copper gasket 25, the compression nut 22A is screwed, and the copper gasket is deformed through the pressing block 24, so that the temperature measurement of high-temperature and high-pressure fluid is realized.

Example 9

In this embodiment, the structure of the seal assembly engaged with the cylinder pressure introduction pipe i 26 and the cylinder pressure introduction pipe ii 27 will be described.

As shown in fig. 8, the seal assembly to be fitted to the cylinder pressure introduction pipe i 26 or the cylinder pressure introduction pipe ii 27 includes a pipe seat 21B and a gland nut 22B coaxial with each other, one end of the tube seat 21B is welded on the outer wall of the middle cylindrical barrel 1, the other end of the tube seat 21B is concavely provided with a spherical concave surface, the central hole of the compression nut 22B is a two-stage stepped hole, the large-diameter section is sleeved on one side of the side wall of the tube seat 21B far away from the middle cylindrical barrel 1 and is in threaded connection with the tube seat 21B, the ends of the cylinder body pressure guiding pipe I26 and the cylinder body pressure guiding pipe II 27 are protruded outwards to form a spherical convex surface, and is matched with the spherical concave surface, one end of the cylinder pressure leading pipe I26 and one end of the cylinder pressure leading pipe II 27 far away from the end head are both penetrated out of the corresponding small-diameter section of the compression nut 22B, and the hole bottom of the large-diameter section is contacted with one end of the end head far away from the tube seat 21B, and one end of the through hole on the tube seat 21B penetrates through the tube wall of the middle cylindrical tube body 1.

During installation, the tube seat 21B is welded on the reactor pressure vessel barrel, then the compression nut 22B is screwed on the tube seat 21B, the ends of the barrel pressure leading tube I26 and the barrel pressure leading tube II 27 are opposite to the corresponding spherical concave surface of the tube seat 21B, then the compression nut 22B is screwed, the hole bottom of the large-diameter section of the compression nut 22B presses the end to move towards the middle cylindrical barrel 1, so that the spherical convex surface on the end is in contact with the spherical concave surface on the barrel pressure leading tube I26, and the end is compressed on the spherical concave surface and is in a close contact state. The ends of the cylinder body pressure guiding pipe I26 and the cylinder body pressure guiding pipe II 27 are spherical sealing surfaces, and the ends of the cylinder body pressure guiding pipe I26 and the cylinder body pressure guiding pipe II 27 are in close contact with the corresponding spherical sealing surfaces of the pipe seat 221 through the compression nut 22B, so that the pressure measurement of high-temperature and high-pressure fluid is realized.

The above-mentioned embodiments are intended to illustrate the objects, technical solutions and advantages of the present invention in further detail, and it should be understood that the above-mentioned embodiments are merely exemplary embodiments of the present invention, and are not intended to limit the scope of the present invention, and any modifications, equivalent substitutions, improvements and the like made within the spirit and principle of the present invention should be included in the scope of the present invention.

Claims (7)

1. A reactor pressure vessel experiment simulator for pressurized water reactor loss of coolant accident, including reactor pressure vessel experiment simulator body and reactor core detection device, reactor pressure vessel experiment simulator body includes pressure vessel simulator body and is located the hanging flange simulator body (3) of pressure vessel simulator internal chamber, is provided with the reactor core subassembly in hanging the flange simulator body and simulates the body, and pressure vessel simulator body is including welded upper cover (18), middle cylinder (1) and low head (19) in proper order, and the reactor core subassembly simulates the body and includes cladding pipe (12), arranges electric heating element (11) in cladding pipe (12) and is used for the location grillwork (13) of fixed each cladding pipe (12), reactor core detection device is used for gathering the inside parameter information of reactor core subassembly simulator when experimenting, its characterized in that: the top of the cladding tube (12) is a closed end, the bottom of the cladding tube is an open end, the bottom of the cladding tube (12) extends downwards and penetrates out of the lower end enclosure (19), the cladding tube (12) is hermetically welded with a through hole in the lower end enclosure (19) for facilitating the penetration of the bottom of the cladding tube (12), and the data acquisition end of the reactor core detection device is inserted into the cladding tube (12) from the open end of the cladding tube (12);

the reactor core detection device comprises a reactor core internal coolant temperature measurement assembly (7), wherein the temperature measurement end of a reactor core thermocouple (14) of the reactor core internal coolant temperature measurement assembly (7) is inserted into the reactor core from the opening end of a cladding tube (12), the tail end of the temperature measurement end penetrates out of the wall of the cladding tube (12) and is 0.1-5 mm in length, the reactor core thermocouples (14) in the same cladding tube (12) are provided with a plurality of temperature measurement ends, and the temperature collection positions of the temperature measurement ends are distributed along the axis of the cladding tube (12).

2. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to claim 1, wherein: the reactor core detection device comprises a reactor core internal coolant pressure measuring assembly (8), one end of a reactor core pressure leading pipe I (15) of the reactor core internal coolant pressure measuring assembly (8) is connected to a signal receiving end of the intelligent pressure transmitter, the other end of the reactor core pressure leading pipe I (15) is inserted into the reactor core pressure leading pipe from the opening end of the cladding pipe (12), the tail end of the reactor core pressure leading pipe I (15) penetrates out of the pipe wall of the cladding pipe (12), and the penetrating length of the reactor core pressure leading pipe I is 0.1-5 mm.

3. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to claim 1, wherein: the reactor core detection device comprises a reactor core liquid level measurement assembly (9), the reactor core liquid level measurement assembly (9) comprises two sets of reactor core pressure leading pipes II (16), one ends of the reactor core pressure leading pipes II (16) are connected into a signal receiving end of an intelligent pressure transmitter, the other ends of the reactor core pressure leading pipes II (16) are inserted into an opening end of a cladding pipe (12), the tail ends of the ends penetrate out of the pipe wall of the cladding pipe (12) and are 0.1-5 mm in length, and the reactor core pressure leading pipes II (16) measure pressure information of two axial heights inside the same cladding pipe (12).

4. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to any one of claims 1 to 3, wherein: the core detection device also comprises a fuel element wall temperature measuring assembly, the fuel element wall temperature measuring assembly comprises two wall temperature thermocouples (17), the temperature collecting ends of the wall temperature thermocouples (17) are inserted into the cladding tube (12) from the open end of the cladding tube (12), one of the temperature collecting ends is fixed on the inner wall of the cladding tube (12), the other temperature collecting end is fixed on the outer wall of the electric heating element (11), and the collecting ends of the two wall temperature thermocouples (17) are positioned in the same radial direction of the cladding tube (12).

5. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to any one of claims 1 to 3, wherein: the device comprises a pressure vessel, and is characterized by further comprising a pressure vessel detection device, wherein the pressure vessel detection device comprises a pressure vessel coolant temperature measurement component (4), a pressure vessel coolant pressure measurement component (5) and a pressure vessel coolant liquid level detection component (6), a plurality of sealing components are arranged on the outer wall of the middle cylindrical barrel (1), the temperature acquisition end of a barrel thermocouple (20) of the pressure vessel coolant temperature measurement component (4), the tail end of a barrel pressure leading pipe I (26) of the pressure vessel coolant pressure measurement component (5) and the tail end of a barrel pressure leading pipe II (27) of the pressure vessel coolant liquid level detection component (6) are respectively inserted into one sealing component and are communicated with the inner cavity of the middle cylindrical barrel (1) through the sealing components, and the barrel pressure leading pipes II (27) of the pressure vessel coolant liquid level detection component (6) are provided with two groups, one group is arranged on the upper part of the middle cylindrical barrel body (1), the other group is arranged on the lower part of the middle cylindrical barrel body (1), and the sealing component seals the position where the middle cylindrical barrel body (1) is communicated with the signal acquisition end of the pressure container detection device.

6. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to claim 5, wherein: the sealing assembly matched with the cylinder thermocouple (20) comprises a tube seat (21A), a compression nut (22A), a compression bar (23) and a red copper gasket (25) which are coaxial with each other, one end of the tube seat (21A) is welded on the outer wall of the middle cylindrical cylinder (1), the other end of the tube seat (21A) is concaved inwards to form a spherical concave surface, a central hole of the compression nut (22A) is a two-stage stepped hole, a large-diameter section of the compression nut is sleeved on one side, far away from the middle cylindrical cylinder (1), of the side wall of the tube seat (21A) and is in threaded connection with the tube seat (21A), one end of the compression bar (23) is connected with a pressing block (24), one end, far away from the compression bar (23), of the pressing block (24) protrudes outwards to form a spherical convex surface and is matched with the spherical concave surface, the other end of the compression bar (23) penetrates through a small-diameter section of the compression nut (22A), and the hole bottom of the large-diameter section is in contact with one end, far away from the tube seat (21), on the pressing block (24), one end of a through hole on the tube seat (21A) penetrates through the wall of the middle cylindrical barrel (1), and the other end of the tube seat (21A) penetrates through the pressing block (24) and the pressing rod (23) in sequence;

the red copper gasket (25) is positioned between the spherical convex surface of the pressing block (24) and the spherical concave surface of the tube seat (21A), the pressing nut (22A) is rotated, when the pressing nut (22A) moves towards the middle cylindrical barrel body (1), the pressing block (24) presses on the red copper gasket (25), and the red copper gasket (25) is deformed, so that one side, facing the tube seat (21A), of the red copper gasket (25) is internally tangent to the spherical concave surface, and the spherical convex surface is internally tangent to one side, away from the tube seat (21A), of the red copper gasket (25);

the temperature acquisition end of the cylinder thermocouple (20) sequentially penetrates through the through hole in the pressure lever (23), the through hole in the pressure block (24), the red copper gasket (25), the through hole in the tube seat (21A) and the wall of the middle cylinder (1) to be positioned in the inner cavity of the middle cylinder (1).

7. The reactor pressure vessel experimental simulator for loss of coolant accident of pressurized water reactor according to claim 5, wherein: the sealing assembly matched with the cylinder body pressure guiding pipe I (26) or the cylinder body pressure guiding pipe II (27) comprises a pipe seat (21B) and a compression nut (22B) which are coaxial, one end of the pipe seat (21B) is welded on the outer wall of the middle cylindrical cylinder body (1), the other end of the pipe seat (21B) is concaved inwards to form a spherical concave surface, a central hole of the compression nut (22B) is a two-stage stepped hole, a large-diameter section of the compression nut is sleeved on one side, far away from the middle cylindrical cylinder body (1), of the side wall of the pipe seat (21B) and is in threaded connection with the pipe seat (21B), the ends of the cylinder body pressure guiding pipe I (26) and the cylinder body pressure guiding pipe II (27) are protruded outwards to form a spherical convex surface and are matched with the spherical concave surface, one end, far away from the ends, of the cylinder body pressure guiding pipe I (26) and the cylinder body pressure guiding pipe II (27) penetrates out of a small-diameter section of the corresponding compression nut (22B), and the hole bottom of the large-diameter section is contacted with one end, far away from the pipe seat (21B), one end of the through hole on the tube seat (21B) penetrates through the wall of the middle cylindrical barrel body (1).

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