Nuclear pressurized water reactor pipeline thermal cycle test system and method

Abstract

The invention discloses a nuclear power pressurized water reactor pipeline thermal cycle test system and method, the system comprises a test section (1), a high pressure steam generator (2), a plurality of flow control sections (3) and a voltage stabilizer (4), the test section (1), the voltage stabilizer (4) and the plurality of flow control sections (3) are connected in series to form a fluid circulation loop, the plurality of flow control sections (3) are arranged in parallel, the test section (1) further comprises a main pipe (11) and a branch pipe (12), and the main pipe (11) and the branch pipe (12) are connected in parallel. The main pipe (11) is communicated with the branch pipe (12), the high-pressure steam generator (2) is communicated with the main pipe (11) through a plurality of steam injection branches (21), the system can provide various pressure fluctuation forms for a test loop through the high-pressure steam generator, the complex working conditions in actual operation of a nuclear power plant are simulated, meanwhile, flexible switching is achieved under different temperature and flow conditions, and the test efficiency is improved. And the test working conditions of high temperature, low temperature and different flow requirements are met.

Description

Technical Field

[0001] The invention relates to the field of nuclear energy pipeline fatigue analysis research, and in particular to a nuclear pressure water reactor pipeline thermal cycle test system and method. Background Art

[0002] T-type pipes are one of the most common component forms in nuclear power plant design, used for the intersection or diversion of fluids. When high-temperature water and low-temperature water in nuclear power plants are mixed through T-type pipes, the temperature of the inner surface of the pipe changes periodically, which may cause thermal fatigue damage to the pipe.

[0003] At present, when using the test bench for nuclear power plant related pipeline simulation for experimental research, there are the following technical difficulties:

[0004] (1) The starting time of the thermal fatigue test is difficult to determine. After adjusting the pressure and temperature of the test loop to the target values, in order to capture the temperature penetration and stress step change phenomena, the test duration is often increased to avoid missing the key test phenomena, thus wasting a lot of time and manpower costs;

[0005] (2) The influence of pipeline pressure fluctuations is difficult to realize. Experimental research can usually only control parameters such as flow rate and temperature. Since the influence of pressure fluctuations cannot be introduced, it cannot reflect the actual situation of the nuclear power plant;

[0006] (3) The pipeline flow control is limited. Affected by the single pipeline design of the circulation loop and the thermal expansion and contraction of the coolant, the flow rate is maintained unchanged during the test in order to follow the principle of a single control variable. As a result, the maximum target flow rate of the circulation loop can only be based on the low-temperature test conditions.

[0007] Patent document CN113808766B discloses a nanofluid cooling loop system suitable for a pressurized water reactor nuclear power plant. The system includes a nanofluid particle size detector, a nanofluid dispersion tank, and loop pipeline equipment. A sampling box is provided at the front end of the nanofluid particle size detector. A high-frequency ultrasonic vibration disperser group, an agitator, and a filter are provided inside the nanofluid dispersion tank. The start-up of each device is controlled by a controller. The problems of difficulty in determining the start time of the thermal fatigue test, inability to simulate pipeline pressure fluctuations, and limited flow control are not solved.

[0008] Patent document CN110957055B discloses a separate flexible heat pipe cooling system suitable for a pressurized water reactor nuclear power plant, including a closed loop consisting of a hot pipe segment, a connecting pipe segment and a cooling pipe segment. The hot pipe segment adopts a flexible pipe material with good heat exchange performance and capable of bending and deforming. The hot pipe segment is arranged inside the cooling object or wrapped around the cooling object. The cooling medium exchanges heat with the cooling object in the hot pipe segment, enters the cooling pipe segment through the connecting pipe segment, is cooled by the heat exchange device in the cooling pipe segment, and then flows back to the hot pipe segment. The problems of difficulty in determining the starting time of the thermal fatigue test, inability to simulate pipeline pressure fluctuations, and limited flow control are not solved.

[0009] In summary, the above two existing patents have not solved the problems that the starting time of the thermal fatigue test is difficult to determine, the pipeline pressure fluctuation cannot be simulated, and the flow control is limited. Summary of the invention

[0010] Based on the above technical problems, the present invention proposes a nuclear pressure water reactor pipeline thermal cycle test system and method to solve the problems of difficulty in determining the start time of thermal fatigue test, inability to simulate pipeline pressure fluctuations and limited flow control.

[0011] To achieve the above object, the present invention provides a nuclear pressure water reactor pipeline thermal cycle test system.

[0012] A nuclear pressure water reactor pipeline thermal cycle test system comprises a test section, a high-pressure steam generator, a plurality of flow control sections and a pressurizer. The test section, the pressurizer and the plurality of flow control sections are connected in series to form a fluid circulation loop, and the plurality of flow control sections are arranged in parallel. The test section also comprises a main pipe and a branch pipe, the main pipe is connected to the branch pipe, and the high-pressure steam generator is connected to the main pipe through a plurality of steam injection branches.

[0013] Furthermore, the high-pressure steam generator is connected to the main pipe through a plurality of steam injection branches arranged along the axial direction of the main pipe.

[0014] Furthermore, the branch pipe includes a connecting section and a horizontal section, the connecting section is connected to the main pipe, and the angle formed by the axis of the connecting section and the axis of the main pipe is 85°-95°; the horizontal section is connected to the connecting section.

[0015] Furthermore, the axes of the plurality of steam injection branches and the connecting section are located in the same test plane, and the steam injection branches and the connecting section are arranged in parallel; and the horizontal section and the main pipe are arranged in parallel.

[0016] Furthermore, among the plurality of steam injection branches, a distance D between an axis of the steam injection branch farthest from the water inlet of the main pipe and an axis of the connecting section ranges from 1.5 m to 3 m.

[0017] Furthermore, the branch pipe is made of stainless steel, and the surface roughness Ra is 5 μm-20 μm.

[0018] Furthermore, the flow control section includes a heat exchange section and a heating section, and the heat exchange section and the heating section are arranged in series.

[0019] Furthermore, the heat exchange section includes a front main pipeline and a front branch pipeline, and the front main pipeline and the front branch pipeline are arranged in parallel.

[0020] Furthermore, the heat exchange section includes a heat exchanger, and the heat exchanger is arranged on the front main pipe.

[0021] Furthermore, the diameter D of the front main pipe is 11 The diameter D of the front branch pipe 12 The ratio ranges from 1:1.05 to 1:0.95.

[0022] Furthermore, the heating section includes a rear main pipeline and a rear branch pipeline, and the rear main pipeline and the rear branch pipeline are arranged in parallel.

[0023] Furthermore, the heating section comprises a heater, and the heater is arranged on the rear main pipe.

[0024] Furthermore, the diameter D of the front main pipe is 11 The diameter D of the rear main pipe 21 The ratio range is 1:1.05 to 1:0.95; the diameter D of the rear main pipe 21 The diameter D of the rear branch pipe 22 The ratio range is 1.95:1 to 2.05:1.

[0025] Furthermore, the pressure stabilizer is arranged on the pipeline between the test section and the plurality of flow control sections.

[0026] Furthermore, the distance between the plurality of steam injection branches is 10 cm-15 cm; and the length of the connecting section is 0.3 m-1.5 m.

[0027] The present invention also proposes a nuclear pressure water reactor pipeline thermal cycle test method, using the above-mentioned nuclear pressure water reactor pipeline thermal cycle test system, comprising:

[0028] S1: Open the valve of the flow control section and start the pressure stabilizer;

[0029] S2: When the coolant in the loop reaches the preset temperature, pressure and flow rate values, the valve of the high-pressure steam generator is opened and steam is injected into the main pipe of the test section through the plurality of steam injection branches.

[0030] Furthermore, the step S2 includes:

[0031] The branch pipe is monitored by an infrared temperature detector to obtain the position of the thermal stratification boundary line of the branch pipe. When the position of the thermal stratification boundary line of the branch pipe reaches a preset position, the valves of the plurality of steam injection branches are opened for injection and the time is recorded.

[0032] Furthermore, the step S2 further includes:

[0033] Steam is continuously injected into the main pipe at a time interval of 1s-2s. The test duration is 0.2h-1h. During the steam injection process, the temperature data of the branch pipe is continuously recorded by the infrared temperature detector.

[0034] Based on the above technical solution, the present invention has at least the following beneficial effects:

[0035] 1. The present invention proposes a nuclear pressurized water reactor pipeline thermal cycle test system and method, which significantly improves the functionality and adaptability of the test system through a high-pressure steam generator, a parallel pipeline design, and precise control of temperature and flow; the high-pressure steam generator can provide a variety of pressure fluctuation forms for the test loop to simulate the complex working conditions in the actual operation of the nuclear power plant, thereby more realistically reflecting the thermal fatigue characteristics of the pipeline under different pressure conditions; through the parallel design of multiple flow control sections, the system can flexibly switch under different temperature and flow conditions to meet high temperature, low temperature and different flow requirements of the test conditions. This design not only breaks through the flow limitation of the traditional single pipeline design, but also realizes precise control of the flow through the layout of pipelines with different diameters, ensuring the reliability and scientificity of the test data.

[0036] 2. The present invention proposes a nuclear pressure water reactor pipeline thermal cycle test system and method, which realizes accurate determination of the test start time through an infrared monitoring and control system; in traditional tests, since it is difficult to determine the start time of the test, it is usually necessary to extend the test duration to ensure that key phenomena are captured, which leads to a large waste of time and manpower costs. The present invention uses an infrared monitoring system to monitor the temperature changes of the test section in real time, and accurately determines the test start time based on the position of the thermal stratification boundary line, thereby significantly reducing the test duration and improving the test efficiency. Accurate start time determination also provides a more reliable basis for subsequent data recording and analysis, ensuring the accuracy and repeatability of the test results; in addition, the system adjusts the temperature of the cooling medium through a heater and a heat exchanger, further optimizing the test conditions so that it can adapt to a variety of complex test requirements.

[0037] 3. The present invention proposes a nuclear pressure water reactor pipeline thermal cycle test system and method. The parallel pipeline design and pipeline layout of different diameters significantly improve the flexibility of the test system and the flow control accuracy. By setting pipelines of different diameters and using flow meters, the flow measurement is more accurate, the measurement error is significantly reduced, and the reliability and scientificity of the test data are improved. This design not only optimizes the test process, but also provides more accurate test conditions for nuclear power plant pipeline thermal fatigue research. BRIEF DESCRIPTION OF THE DRAWINGS

[0038] The accompanying drawings constituting a part of the present invention are used to provide a further understanding of the present invention. The exemplary embodiments of the present invention and their descriptions are used to explain the present invention and do not constitute an improper limitation of the present invention. In the accompanying drawings:

[0039] Figure 1 A schematic structural diagram of a nuclear pressure water reactor pipeline thermal cycle test system according to an embodiment is shown;

[0040] Figure 2 A schematic diagram of a branch pipe structure of an embodiment is shown;

[0041] Figure 3 A schematic diagram of pressure fluctuations during steam injection branch injection according to an embodiment is shown.

[0042] The above-mentioned drawings include the following reference numerals:

[0043] 1. Test section; 2. High-pressure steam generator; 3. Flow control section; 4. Voltage stabilizer; 5. Infrared temperature detector; 6. Cooling medium;

[0044] 11. Main pipe; 12. Branch pipe;

[0045] 21. Steam injection branch;

[0046] 31. Heat exchange section; 32. Heating section; 33. Main flow meter; 34. Check valve; 35. Loop pump; 36. Branch flow meter;

[0047] 121. connecting section; 122. horizontal section;

[0048] 311. Front main pipeline; 312. Front branch pipeline; 313. Heat exchanger;

[0049] 321. Rear main pipeline; 322. Rear branch pipeline; 323. Heater. DETAILED DESCRIPTION

[0050] It should be noted that, in the absence of conflict, the embodiments of the present invention and the features in the embodiments can be combined with each other. The present invention will be described in detail below with reference to the accompanying drawings and in combination with the embodiments.

[0051] The present invention is further described in detail below in conjunction with specific embodiments, and these embodiments cannot be understood as limiting the scope of protection claimed by the present invention. The term "including" when used indicates the existence of a feature, but does not exclude the existence or addition of one or more other features; the orientation or position relationship indicated by the terms "lateral", "upper", "lower", "front", "rear", "left", "right", "top", "bottom", "inside", "outside", etc. is based on the orientation or position relationship shown in the drawings, which is only for the convenience of description, and does not indicate or imply that the device or element referred to must have a specific orientation, be constructed and operated in a specific orientation, and therefore cannot be understood as a limitation on the present invention; in addition, the terms "first" and "second" are only used for descriptive purposes, and cannot be understood as indicating or implying relative importance.

[0052] In the description, it should be noted that, unless otherwise clearly specified and limited, the terms "installed", "connected", and "connected" should be understood in a broad sense. For example, it can be a fixed connection, a detachable connection, or an integral connection; it can be directly connected, or indirectly connected through an intermediate medium, or it can be the internal communication of two components. For ordinary technicians in this field, the specific meanings of the above terms in the present invention can be understood according to specific circumstances. In addition, in the description of the present invention, unless otherwise specified, "plurality" means two or more.

[0053] Example

[0054] The present invention provides a nuclear pressure water reactor pipeline thermal cycle test system, such as Figure 1 As shown in the figure, it includes a test section 1, a high-pressure steam generator 2, multiple flow control sections 3 and a pressure stabilizer 4. The test section 1, the pressure stabilizer 4 and the multiple flow control sections 3 are connected in series to form a fluid circulation loop, and the multiple flow control sections 3 are arranged in parallel. The test section 1 also includes a main pipe 11 and a branch pipe 12. The main pipe 11 is connected to the branch pipe 12, and the high-pressure steam generator 2 is connected to the main pipe 11 through multiple steam injection branches 21.

[0055] Furthermore, in the present invention, "connectivity" refers to connecting different containers or devices through pipes, pipelines, etc., so as to realize the transmission and distribution of fluids such as gases or liquids; in such a connected system, fluids such as gases or liquids can flow from one container to another or from one device to another under the action of pressure difference.

[0056] Furthermore, if Figure 2 As shown in the figure, the test section 1 consists of a main pipe 11 and a branch pipe 12. The main pipe 12 is connected to the lower part of the main pipe 11 near the water outlet. The branch pipe 12 includes a connecting section 121 and a horizontal section 122. The connecting section 121 is connected to the main pipe 11, and the horizontal section 122 is connected to the connecting section 121.

[0057] Preferably, in this embodiment, the branch pipe 12 is made of stainless steel with a surface roughness Ra of 15 μm; the angle formed by the axis of the connecting section 121 and the axis of the main pipe is 90°, and the horizontal section 122 is arranged parallel to the main pipe. Since the length of the connecting section 121 is too small, thermal stratification will not occur, and it is difficult to achieve heat penetration when it is too long. The length of the connecting section 121 is set to 1 m. In other embodiments, the angle, roughness and length of the branch pipe can be set to any value within a range.

[0058] Furthermore, if Figure 1 As shown in , a plurality of steam injection branches 21 are arranged at the top of the main pipe 11 along the axial direction, and the high-pressure steam generator 2 is connected to the main pipe through the steam injection branch 21 , and each of the steam injection branch 21 is provided with a valve to control the opening.

[0059] Furthermore, the axes of the multiple steam injection branches 21 and the connecting section 121 are located in the same test plane, and the multiple steam injection branches 21 are arranged parallel to each other.

[0060] Specifically, in this embodiment, Figure 1 As shown in FIG. 3 , three steam injection branches 21 are arranged on the main pipe 11 , the distance between the steam injection branches 21 is 12 cm, and the distance between the steam injection branch 21 farthest from the water inlet of the main pipe 11 and the axis of the branch pipe is 1.8 m.

[0061] Furthermore, the flow control section 3 includes a heat exchange section 31 and a heating section 32, and the heat exchange section 31 and the heating section 32 are arranged in series. The heat exchange section 31 includes a front main pipeline 311 and a front branch pipeline 312, and the front main pipeline 311 and the front branch pipeline 312 are arranged in parallel; the heating section 32 includes a rear main pipeline 321 and a rear branch pipeline 322, and the rear main pipeline 321 and the rear branch pipeline 322 are arranged in parallel.

[0062] Furthermore, a loop pump 35 is provided between the heat exchange section 31 and the heating section 32 , a heater 323 is provided on the rear main line 321 , and a heat exchanger 313 is provided on the front main line 311 .

[0063] Furthermore, a main flow meter 33 and a check valve 34 are arranged between the flow control section 3 and the test section 1 , the check valve 34 is arranged upstream of the main flow meter 33 , and the branch flow meter 36 is arranged on the rear branch pipeline 322 .

[0064] Preferably, the diameter D of the front main pipe 311 is 11 The diameter D of the front branch pipe 312 12 The ratio is 1:1; the diameter D of the front main pipe 311 11The diameter D of the rear main pipe 321 21 The ratio is 1:1; the diameter D of the rear main pipe 321 21 The diameter D of the rear branch pipe 322 22 The ratio is 2:1.

[0065] Specifically, in this embodiment, Figure 1 As shown in , two flow control sections 3 are set, and the two flow control sections 3 are set in parallel. The two flow control sections each include a heat exchange section 31, a heating section 32, a main flow meter 33, a check valve 34, a loop pump 35, a branch flow meter 36, a front main line 311, a front branch line 312, a heat exchanger 313, a rear main line 321, a rear branch line 322 and a heater 323; the diameter D of the front main line 311 of the two flow control sections 3 11 The diameter D of the front branch pipe 312 12 The ratio is 1:1, and the diameter D of the front main pipe 311 is 11 The diameter D of the rear main pipe 321 21 The ratio is 1:1, and the diameter D of the rear main pipe 321 is 21 The diameter D of the rear branch pipe 322 22 The ratio is 2:1; Figure 1 The diameter D of the rear main pipe 321 of the flow control section 3 on the left side of the middle 21 The diameter D of the rear main pipe 321 of the right flow control section 3 is 21 The ratio is 2:1.

[0066] Furthermore, the pressure stabilizer 4 is arranged on the pipeline between the test section 1 and the plurality of flow control sections 3 .

[0067] Furthermore, the cooling medium 6 may be deionized water.

[0068] To achieve the above object, the present invention also proposes a nuclear pressure water reactor pipeline thermal cycle test method, using a nuclear pressure water reactor pipeline thermal cycle test system according to the above, comprising the following steps:

[0069] S1: Open the valve of the flow control section 3 and start the voltage regulator 4;

[0070] S2: When the coolant in the loop reaches the preset temperature, pressure and flow rate values, the valve of the high-pressure steam generator 2 is opened and steam is injected into the main pipe 11 of the test section 1 through the multiple steam injection branches 21.

[0071] Furthermore, the heater 323 of the regulator 4 and the flow control section 3 is turned on to heat the internal cooling medium 6 simultaneously. During the heating process, the temperature t of the cooling medium 6 in the regulator 4 must be kept constant. 1 Greater than the temperature of the circulating coolant t2 , and satisfy 10℃≤t 1 -t 2 ≤100℃, in this embodiment, t 1 -t 2 The reference value is 20℃.

[0072] Furthermore, when the temperature and pressure of the cooling medium 6 in the loop reach the target value, the loop pump 5 is adjusted to make the flow rate of the loop reach the target value; the target value range of the loop flow rate is 400m 3 / h-500m 3 / h; In this embodiment, the target value of the cooling medium 6 is 300°C, the target value of the pressure is 15.5MPa, and the target value of the loop flow is 450m 3 / h.

[0073] Furthermore, setting Figure 1 The two flow control sections 3 shown in the figure are selected to test the flow control section 3 on the left. At this time, the valves of the other pipelines of the rear main pipeline 321 and the front main pipeline 311 are closed. When the loop pump 35 adopts a variable frequency pump and the operating frequency reaches the maximum, the flow rate of the cooling medium 6 in the circulation loop reaches the maximum value Q max Due to the thermal expansion and contraction of the cooling medium 6, the viscosity of the coolant is small at high temperature and the resistance is reduced, resulting in a larger flow rate at low temperature. At this time, the front branch pipeline 312 and the rear branch pipeline 322 of the flow control section 3 are opened to increase the maximum flow rate of the circulation loop by 1.2Q max As the temperature rises, the flow rate of the coolant in the loop gradually increases. At this time, the loop flow rate can be adjusted by closing the front branch pipeline 312 and the rear branch pipeline 322 of the flow control section 3 or by reducing the frequency of the loop pump 5 until the loop flow rate reaches the maximum value Q max .

[0074] Furthermore, if Figure 2 As shown in the figure, the branch pipe 12 is monitored by the infrared temperature detector 5 to obtain the position of the thermal stratification boundary line of the branch pipe 12. When the position of the thermal stratification boundary line of the branch pipe 12 reaches a preset position, the valves of the plurality of steam injection branches 21 are opened for injection and the time is recorded.

[0075] Furthermore, steam is continuously injected into the main pipe 11 at intervals of 1s-2s, and the injection pressure waveform is as follows: Figure 3 As shown in FIG, multiple steam injection branches 21 are injected together, the pressure range of steam injection through the steam injection branch 21 is 0.2Mpa-0.6Mpa, the test duration is 0.2h-1h, and the pressure waveform of steam injection can be Figure 3 In this embodiment, one of a, b, c and d is used. Figure 3The test is performed according to the pressure waveform diagram c in the figure. The multiple steam injection branches 21 perform an injection pressure value ranging from 0.2Mpa to 0.4Mpa together. The injection pressure value is changed every 5s, and steam with pressures of 0.2Mpa and 0.4Mpa are injected alternately. During the steam injection process, the temperature data of the branch pipe 12 is continuously recorded by the infrared temperature detector 5.

[0076] Furthermore, the pressure p of the high-pressure steam generator 2 2 The pressure p maintained by the regulator 4 1 Satisfy 1.2p 1 ≤p 2 ≤1.5p 1 , set endpoint p 2 ≥1.2p 1 In order to ensure that each steam injection branch 21 can inject high-temperature and high-pressure steam into the main pipe 11 of the test section 1, the endpoint p is set 2 ≤1.5p 1 It is designed based on the instantaneous maximum working pressure of 1.5 times of the stainless steel pipe. In this embodiment, p 2 =1.3p 1 .

[0077] In summary, it can be seen from the above description that the above embodiments of the present invention achieve the following technical effects:

[0078] 1. The present invention proposes a nuclear pressurized water reactor pipeline thermal cycle test system and method, which significantly improves the functionality and adaptability of the test system through a high-pressure steam generator, a parallel pipeline design, and precise control of temperature and flow; the high-pressure steam generator can provide a variety of pressure fluctuation forms for the test loop to simulate the complex working conditions in the actual operation of the nuclear power plant, thereby more realistically reflecting the thermal fatigue characteristics of the pipeline under different pressure conditions; through the parallel design of multiple flow control sections, the system can flexibly switch under different temperature and flow conditions to meet high temperature, low temperature and different flow requirements of the test conditions. This design not only breaks through the flow limitation of the traditional single pipeline design, but also realizes precise control of the flow through the layout of pipelines with different diameters, ensuring the reliability and scientificity of the test data.

[0079] 2. The present invention proposes a nuclear pressure water reactor pipeline thermal cycle test system and method, which realizes accurate determination of the test start time through an infrared monitoring and control system; in traditional tests, since it is difficult to determine the start time of the test, it is usually necessary to extend the test duration to ensure that key phenomena are captured, which leads to a large waste of time and manpower costs. The present invention uses an infrared monitoring system to monitor the temperature changes of the test section in real time, and accurately determines the test start time based on the position of the thermal stratification boundary line, thereby significantly reducing the test duration and improving the test efficiency. Accurate start time determination also provides a more reliable basis for subsequent data recording and analysis, ensuring the accuracy and repeatability of the test results; in addition, the system adjusts the temperature of the cooling medium through a heater and a heat exchanger, further optimizing the test conditions so that it can adapt to a variety of complex test requirements.

[0080] 3. The present invention proposes a nuclear pressure water reactor pipeline thermal cycle test system and method. The parallel pipeline design and pipeline layout of different diameters significantly improve the flexibility of the test system and the flow control accuracy. By setting pipelines of different diameters and using flow meters, the flow measurement is more accurate, the measurement error is significantly reduced, and the reliability and scientificity of the test data are improved. This design not only optimizes the test process, but also provides more accurate test conditions for nuclear power plant pipeline thermal fatigue research.

[0081] The above description is only a preferred embodiment of the present invention and is not intended to limit the present invention. For those skilled in the art, the present invention may have various modifications and variations. Any modification, equivalent replacement, improvement, etc. made within the spirit and principle of the present invention shall be included in the protection scope of the present invention.

[0082] It should be noted that, in this article, relational terms such as first and second, etc. are only used to distinguish one entity or operation from another entity or operation, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Moreover, the terms "include", "comprise" or any other variants thereof are intended to cover non-exclusive inclusion, so that a process, method, article or device including a series of elements includes not only those elements, but also other elements not explicitly listed, or also includes elements inherent to such process, method, article or device. In the absence of further restrictions, the elements defined by the sentence "comprise a ..." do not exclude the existence of other identical elements in the process, method, article or device including the elements.

[0083] It should be noted that, in the description of this specification, the description with reference to the terms "one embodiment", "some embodiments", "example", "specific example", or "some examples" etc. means that the specific features, structures, materials or characteristics described in conjunction with the embodiment or example are included in at least one embodiment or example of the present invention. In this specification, the schematic representations of the above terms do not necessarily refer to the same embodiment or example. Moreover, the specific features, structures, materials or characteristics described may be combined in any one or more embodiments or examples in a suitable manner. In addition, those skilled in the art may combine and combine the different embodiments or examples described in this specification and the features of the different embodiments or examples, unless they are contradictory.

Claims

1. A nuclear pressure water reactor pipeline thermal cycle test system, characterized in that: The invention comprises a test section (1), a high-pressure steam generator (2), a plurality of flow control sections (3) and a pressure stabilizer (4); the test section (1), the pressure stabilizer (4) and the plurality of flow control sections (3) are connected in series to form a fluid circulation loop; the plurality of flow control sections (3) are arranged in parallel; the test section (1) also comprises a main pipe (11) and a branch pipe (12); the main pipe (11) is connected to the branch pipe (12); the high-pressure steam generator (2) is connected to the main pipe (11) via a plurality of steam injection branches (21).

2. The test system according to claim 1, characterized in that: The high-pressure steam generator (2) is connected to the main pipe (11) through a plurality of steam injection branches (21) arranged along the axial direction of the main pipe (11).

3. The test system according to claim 2, characterized in that: The branch pipe (12) comprises a connecting section (121) and a horizontal section (122). The connecting section (121) is in communication with the main pipe (11), and the angle formed by the axis of the connecting section (121) and the axis of the main pipe (11) is 85°-95°; The horizontal section (122) is connected to the connecting section (121).

4. The test system according to claim 3, characterized in that: The axes of the plurality of steam injection branches (21) and the connecting section (121) are located in the same test plane, and the steam injection branches (21) and the connecting section (121) are arranged in parallel; the horizontal section (122) and the main pipe (11) are arranged in parallel.

5. The test system according to claim 4, characterized in that: Among the multiple steam injection branches (21), the distance D between the axis of the steam injection branch (21) farthest from the water inlet of the main pipe (11) and the axis of the connecting section (121) is in the range of 1.5 m to 3 m.

6. The test system according to claim 1, characterized in that: The branch pipe (12) is made of stainless steel, and its surface roughness Ra is 5 μm-20 μm.

7. The test system according to claim 1, characterized in that: The flow control section (3) comprises a heat exchange section (31) and a heating section (32). The heat exchange section (31) and the heating section (32) are arranged in series.

8. The test system according to claim 7, characterized in that: The heat exchange section (31) comprises a front main pipeline (311) and a front branch pipeline (312). The front main pipeline (311) and the front branch pipeline (312) are arranged in parallel.

9. The test system according to claim 8, characterized in that: The heat exchange section (31) comprises a heat exchanger (313), The heat exchanger (313) is arranged on the front main pipe (311).

10. The test system according to claim 9, characterized in that: The diameter D of the front main pipe (311) 11 The diameter D of the front branch pipeline (312) 12 The ratio ranges from 1:1.05 to 1:0.

95.

11. The test system according to claim 10, characterized in that: The heating section (32) comprises a rear main pipeline (321) and a rear branch pipeline (322). The rear main pipeline (321) and the rear branch pipeline (322) are arranged in parallel.

12. The test system according to claim 11, characterized in that: The heating section (32) comprises a heater (323), The heater (323) is arranged on the rear main pipe (321).

13. The test system according to claim 12, characterized in that: The diameter D of the front main pipe (311) 11 The diameter D of the rear main pipe (321) 21 The ratio ranges from 1:1.05 to 1:0.95; The diameter D of the rear main pipe (321) 21 The diameter D of the rear branch pipeline (322) 22 The ratio range is 1.95:1 to 2.05:

1.

14. The test system according to claim 13, characterized in that: The pressure stabilizer (4) is arranged on the pipeline between the test section (1) and the plurality of flow control sections (3).

15. The test system according to claim 3, characterized in that: The distance between the plurality of steam injection branches (21) is 10 cm to 15 cm; The length of the connecting section (121) is 0.3m-1.5m.

16. A test method based on the nuclear pressure water reactor pipeline thermal cycle test system according to any one of claims 1 to 15, characterized in that: include: S1: Open the valve of the flow control section (3) and start the voltage regulator (4); S2: When the coolant in the loop reaches the preset temperature, pressure and flow rate values, the valve of the high-pressure steam generator (2) is opened and steam is injected into the main pipe (11) of the test section (1) through the multiple steam injection branches (21).

17. The method according to claim 16, characterized in that: The step S2 includes: The branch pipe (12) is monitored by an infrared temperature detector (5) to obtain the position of the thermal stratification boundary line of the branch pipe (12); when the position of the thermal stratification boundary line of the branch pipe (12) reaches a preset position, the valves of the plurality of steam injection branches (21) are opened to perform injection and the time is recorded.

18. The method according to claim 17, characterized in that: The step S2 further includes: Steam is continuously injected into the main pipe (11) at intervals of 1 second to 2 seconds. The test duration is 0.2 hours to 1 hour. During the steam injection process, the temperature data of the branch pipe (12) is continuously recorded by the infrared temperature detector (5).

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