Nuclear power self-stabilization control system and method

Abstract

The application discloses a nuclear power self-stabilization pressure control system and method, which comprises a safety shell (1), a pressure container (2), a core (21) and a main heat exchanger (22). The core (21) and the main heat exchanger (22) are connected through a pipeline to form a natural circulation loop. The system is characterized in that it comprises an intermediate loop (3) and a pressure relief loop (4). The intermediate loop (3) is arranged outside the safety shell (1) and penetrates the safety shell (1) and the pressure container (2) through a pipeline, so that heat exchange with the main heat exchanger (22) can be achieved. The pressure relief loop (4) comprises a water pool (41) and a pressure relief pipeline (42). The pressure relief pipeline (42) is connected to the top of the pressure container (2) and the water pool (41). An automatic pressure relief system (43) is arranged on the pressure relief pipeline (42), which can actively discharge the medium in the pressure container (2) in an accident.

Description

Technical Field

[0001] This invention relates to the field of nuclear reactor safety technology, specifically to a nuclear power plant self-stabilizing voltage control system and method. Background Technology

[0002] Integrated reactors, exemplified by nitrogen self-regulating reactors, employ nitrogen self-regulating technology. The primary loop pressure in the upper gas space of the pressure vessel is maintained by the combined partial pressure of nitrogen and water vapor. The reactor is equipped with a passive residual heat removal system (PRHRS), which relies on triple natural circulation to remove core decay heat, ensuring core flooding.

[0003] Existing passive safety systems typically reduce primary circuit pressure and achieve steam condensation by directly releasing high-temperature, high-pressure coolant from the Automatic Depressurization System (ADS) into the containment space or the water storage area within the containment. However, during accident depressurization, the ADS discharge medium includes not only liquid coolant and water vapor, but also non-condensable gases such as nitrogen released from the upper gas chamber of the pressure vessel. Because water vapor can condense rapidly upon contact with cooling water, while non-condensable gases such as nitrogen cannot undergo phase change, they gradually accumulate in the upper space of the containment or water storage device.

[0004] On the one hand, when non-condensable gases accumulate in large quantities inside the containment, they will still increase the pressure in the containment's gas phase space and reduce the steam condensation efficiency within the containment, affecting the ability to rapidly relieve pressure after an accident. On the other hand, existing water storage structures typically lack separation and diversion measures for the gas-liquid two-phase media. Non-condensable gases generated by ADS emissions can easily enter the free space of the containment during the steam release process, making centralized collection and subsequent treatment difficult. Furthermore, when the gas phase space at the top of the pool is connected to the containment space, the migration path of non-condensable gases is uncontrollable, potentially leading to localized gas accumulation within the containment, further affecting heat removal and pressure control after an accident.

[0005] On the other hand, although the existing design can ultimately maintain core safety, it still has the following shortcomings in the accident response process: First, regarding the containment pressure peak, high-temperature and high-pressure coolant is injected into the containment during a breach accident, causing the containment pressure to rise rapidly. Studies have shown that the containment pressure peak occurs approximately 1000 seconds after the accident, and excessively high pressure can threaten the structural integrity of the containment. Second, nitrogen has a low density and accumulates in the upper gas chamber of the pressure vessel. When the water level drops below the top of the built-in main heat exchanger (PHE), the natural circulation working fluid may carry nitrogen into the heat exchange tubes. Since the PHE has a direct-flow structure, once nitrogen enters the tube bundle, it will be carried deeper by the flowing working fluid and is difficult to discharge through a single point at the top, thus continuously reducing condensation efficiency and delaying the depressurization process of the primary loop. In addition, improper timing of ADS opening and closing can cause problems. If ADS is opened too early, the discharge is mainly steam, resulting in significant coolant loss; if it is opened too late, nitrogen has already entered the heat exchanger; if ADS is closed too early, nitrogen has not been vented; if it is closed too late, coolant continues to be lost and heat accumulates in the containment. Existing technologies lack precise control strategies for the characteristics of nitrogen; relying solely on pressure drop time or fixed pressure thresholds cannot determine whether nitrogen has been vented, which may lead to misjudgments.

[0006] Patent document CN115410728B discloses a fully natural circulation reactor system, including a reactor pressure vessel and in-core components. Fuel assemblies are arranged in the lower region of the vessel, and a control rod drive mechanism is located in an upper support assembly in the upper region of the vessel. Primary and secondary loop heat exchange assemblies are arranged around the upper support assembly. The main body of the primary and secondary loop heat exchange assemblies includes a guide tube assembly with openings at the upper and lower ends. The guide tube wall is a sealed structure to separate the flow paths of hot and cold fluids. The primary loop coolant in the vessel enters the fuel assembly along the inner wall of the vessel, exchanges heat, flows upward, passes through the top of the upper support assembly, and flows through the primary and secondary loop heat exchange assemblies for cooling before continuing to flow downward. Full natural circulation is achieved through the density difference of the primary loop coolant. However, this invention does not solve the problems of inaccurate start-up and shutdown of the automatic depressurization system and ineffective separation and management of non-condensable gases in nitrogen self-regulating integrated reactor accidents.

[0007] Patent document CN121034682A discloses a natural circulation reactor steam supply system and nuclear power plant based on a loop thermosiphon, including a reactor pressure vessel and a reactor core, control rod drive mechanism, and in-core heat exchanger located inside the reactor pressure vessel; the heat pipe circulation loop includes a steam heat exchanger and a reflux heat exchanger, one end of the steam heat exchanger is connected to the primary medium outlet of the in-core heat exchanger, and the other end is connected to the primary medium inlet of the steam generator, the steam heat exchanger being used to transport gaseous working fluid; one end of the reflux heat exchanger is connected to the primary medium outlet of the steam generator, and the other end is connected to the primary medium inlet of the in-core heat exchanger, the reflux heat exchanger being used to transport liquid working fluid; however, it does not solve the problems of the automatic depressurization system's inaccurate start / stop and the inability to effectively separate and manage non-condensable gases during nitrogen self-regulating integrated reactor accidents.

[0008] In summary, neither of the two existing patents mentioned above has solved the problems of the automatic depressurization system's inability to accurately start and stop in nitrogen self-regulating integrated reactor accidents, nor the inability to effectively separate and manage non-condensable gases. Summary of the Invention

[0009] Based on the above-mentioned technical problems, this invention proposes a nuclear power plant self-regulating pressure control system and method to solve the problems of inaccurate start-up and shutdown of the automatic depressurization system and ineffective separation and management of non-condensable gases in nitrogen self-regulating integrated reactor accidents.

[0010] To achieve the above objectives, this invention proposes a nuclear power plant self-regulating voltage control system.

[0011] A nuclear power plant self-regulating pressure control system includes a containment vessel, a pressure vessel, a reactor core, and a main heat exchanger. The reactor core and the main heat exchanger form a natural circulation loop through pipelines. The system is characterized by including an intermediate loop, a pressure relief loop, and monitoring instruments. The intermediate loop is located outside the containment vessel and connects the containment vessel and the pressure vessel through pipelines, enabling heat exchange with the main heat exchanger. The pressure relief loop includes a water pool and a pressure relief pipeline. The pressure relief pipeline connects the top of the pressure vessel and the water pool, and is equipped with an automatic pressure relief system capable of actively releasing the medium inside the pressure vessel in the event of an accident. The monitoring instruments are located inside the pressure vessel.

[0012] Furthermore, the intermediate loop includes a steam generator, a hot pipe, and a cold pipe, wherein the hot pipe and the cold pipe are both connected to the main heat exchanger and the steam generator to form the intermediate loop.

[0013] Furthermore, the hot pipe supplies the cooling medium in the intermediate circuit to the steam generator; the cold pipe supplies the cooling medium in the intermediate circuit to the main heat exchanger.

[0014] Furthermore, the intermediate loop includes a water supply pipeline and a steam pipeline, which are connected to the steam generator for heat exchange with the steam generator.

[0015] Furthermore, the intermediate circuit includes a circulating pump and an isolation valve, wherein the circulating pump is located on the cold pipeline; and the isolation valve is located on both the hot pipeline and the cold pipeline.

[0016] Furthermore, it also includes an air-cooled circuit, which is connected to the intermediate circuit and is capable of dissipating residual heat from the reactor core.

[0017] Furthermore, the air-cooled circuit includes an air-cooled pipeline and a waste heat discharge heat exchanger. The air-cooled pipeline is connected to the hot pipeline and the cold pipeline between the isolation valve and the main heat exchanger; the air-cooled pipeline is connected to the waste heat discharge heat exchanger.

[0018] Furthermore, the air-cooled circuit includes a volume compensation box, which is connected to the air-cooled pipeline and is capable of supplying coolant to the air-cooled pipeline.

[0019] Furthermore, the monitoring instruments include a water level gauge, a pressure gauge, and a steam thermometer.

[0020] Furthermore, the water tank includes a cap and a pressure relief valve. The cap is located at the top of the water tank and is capable of confining the medium discharged from the pressure vessel within the space formed by the water tank and the cap. The pressure relief valve is located on the cap.

[0021] Furthermore, the water tank includes a gas medium storage chamber and a purification component. The gas medium storage chamber is located at the bottom of the cap and is capable of collecting the gas emitted by the pressure vessel. The purification component is located at the air inlet of the gas medium storage chamber.

[0022] To achieve the above objectives, the present invention also proposes a nuclear power plant self-regulating voltage control method, using a nuclear power plant self-regulating voltage control system according to the above description, comprising the following steps: S1: Collect monitoring data from the pressure vessel using the monitoring instrument. The monitoring data includes water level, pressure, and temperature. S2: The automatic pressure relief system starts or stops discharging the medium into the pressure vessel into the water tank according to the monitoring data.

[0023] Further, step S2 includes: setting a low water level threshold based on the vertical height of the main heat exchanger within the pressure vessel; activating the automatic pressure relief system when the water level is below the threshold; obtaining the saturation pressure based on the temperature and pressure of the pressure vessel; and deactivating the automatic pressure relief system when the difference between the pressure and the saturation pressure is less than or equal to the allowable residual pressure difference.

[0024] Based on the above technical solution, the present invention has at least the following beneficial effects: 1. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting an intermediate loop outside the containment vessel and connecting it to the containment vessel and pressure vessel via pipelines, and exchanging heat with the main heat exchanger, the heat from the primary loop can be transferred outwards through an independent intermediate loop. This reduces the risk of radioactive cooling media directly entering the external heat exchange system and isolates the internal and external heat transfer processes, improving system operational safety. Furthermore, by setting a pressure relief loop connecting the top of the pressure vessel and the water pool, and installing an automatic pressure relief system on the relief pipeline, the high-pressure medium inside the pressure vessel can be actively directed to the water pool for release in the event of an accident, reducing the internal pressure of the pressure vessel and mitigating the pressure rise caused by the release of high-temperature, high-pressure media during an accident, thus improving accident response capability. Monitoring instruments are installed to monitor the internal state parameters of the pressure vessel in real time, providing a data basis for judging the reactor's operating status and accident response. By constructing independent heat transfer paths and accident pressure release paths, this invention achieves coordinated operation of reactor heat removal and pressure regulation processes, improving the state maintenance capability and safety margin of the nitrogen self-regulating integrated reactor in the face of accident disturbances.

[0025] 2. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting up an intermediate loop outside the containment, including a steam generator, hot pipes, and cold pipes, the intermediate loop is connected to the main heat exchanger inside the pressure vessel and the external steam generator, enabling the transfer of core heat to the secondary side via the main heat exchanger. By installing a circulating pump and isolation valve on the intermediate loop, the cooling medium in the intermediate loop can be controlled to open and close according to the operating status, improving the controllability of the heat transfer process. By setting up an air-cooled loop connected to the intermediate loop and a residual heat removal heat exchanger, the decay heat generated by the core under accident conditions can be transferred to the external environment via the main heat exchanger and the intermediate loop, reducing reliance on active safety equipment and improving the reliability of long-term residual heat removal after an accident. By setting up a volume compensation tank to replenish the cooling medium in the air-cooled pipes, the volume changes in the loop caused by temperature variations or coolant loss can be mitigated, improving the operational stability of the air-cooled loop.

[0026] 3. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting a capping structure at the top of the pool, the high-temperature and high-pressure medium discharged from the pressure vessel through the automatic pressure relief system can be confined within the closed space formed by the pool and the capping structure, preventing the pressure relief medium from directly diffusing into the containment space, thereby reducing the risk of rapid pressure rise in the containment during an accident. By setting a gas medium storage chamber at the bottom of the capping structure, the gas-liquid mixture such as steam and nitrogen entering the pool can be collected centrally after condensation and separation, preventing the disorderly migration of non-condensable gases within the containment and reducing the adverse effects of non-condensable gases entering the heat exchange area on heat exchange performance. By setting a purification component at the gas medium storage chamber inlet, the gas entering the storage chamber is filtered, reducing the possibility of aerosols or other pollutants migrating with the gas during accident pressure relief, and improving the ability to manage accident products and control the environment within the containment.

[0027] 4. This invention proposes a nuclear power plant self-regulating pressure control system and method. By collecting monitoring data such as water level, pressure, and temperature inside the pressure vessel, and controlling the automatic pressure relief system to start or stop based on the monitoring data, the pressure relief process of the pressure vessel during an accident can be dynamically adjusted according to the actual operating status. This avoids the problems of premature pressure relief leading to excessive coolant loss or delayed pressure relief leading to continuous pressure rise caused by fixed pressure threshold control. By setting a low water level threshold based on the height of the main heat exchanger and determining the pressure relief termination condition based on the residual pressure difference between the pressure and the saturation pressure, the automatic pressure relief process can simultaneously consider the core cooling requirements and the primary loop depressurization status, improving the accuracy of ADS action timing judgment, reducing the risk of nitrogen entering the main heat exchanger and continuous coolant loss, thereby improving the stability of the accident response process. Attached Figure Description

[0028] The accompanying drawings, which form part of this specification, are used to provide a further understanding of the invention. The illustrative embodiments of the invention and their descriptions are used to explain the invention and do not constitute an undue limitation of the invention. In the drawings: Figure 1 A cross-sectional view of a nuclear power plant self-regulating voltage control system according to one embodiment is shown; Figure 2 A schematic diagram of the vertical height of the main heat exchanger in one embodiment is shown.

[0029] The above figures include the following reference numerals: 1. Containment vessel; 2. Pressure vessel; 3. Intermediate circuit; 4. Pressure relief circuit; 5. Air-cooled circuit; 6. Monitoring instruments; 21. Reactor core; 22. Main heat exchanger; 31. Steam generator; 32. Hot piping; 33. Cold piping; 34. Water supply line; 35. Steam line; 36. Circulating pump; 37. Isolation valve; 41. Water tank; 42. Pressure relief pipeline; 43. Automatic pressure relief system; 411. Gas medium storage chamber; 412. Purification component; 413. Pressure relief valve; 414. Top cover; 51. Air-cooled piping; 52. Waste heat discharge heat exchanger; 53. Volume compensation tank; 54. Air-cooled tower; 55. Waste heat discharge isolation valve. Detailed Implementation

[0030] It should be noted that, unless otherwise specified, the embodiments and features described in the present invention can be combined with each other. The present invention will now be described in detail with reference to the accompanying drawings and embodiments.

[0031] The present invention will be further described in detail below with reference to specific embodiments. These embodiments should not be construed as limiting the scope of protection claimed by the present invention. The term "comprising" indicates the presence of a feature, but does not exclude the presence or addition of one or more other features. The terms "lateral," "upper," "lower," "front," "rear," "left," "right," "top," "bottom," "inner," and "outer," etc., indicate the orientation or positional relationship based on the orientation or positional relationship shown in the accompanying drawings, and are only for the purpose of description, and do not indicate or imply that the device or element referred to must have a specific orientation, or be constructed and operated in a specific orientation, and therefore should not be construed as a limitation of the present invention. In addition, the terms "first" and "second" are used for descriptive purposes only and should not be construed as indicating or implying relative importance.

[0032] In this description, it should be noted that, unless otherwise explicitly specified and limited, the terms "installation," "connection," and "linking" should be interpreted broadly. For example, they can refer to a fixed connection, a detachable connection, or an integral connection; they can refer to a direct connection or an indirect connection through an intermediate medium; and they can refer to the internal connection of two components. Those skilled in the art can understand the specific meaning of these terms in this invention based on the specific circumstances. Furthermore, in the description of this invention, unless otherwise stated, "a plurality of" means two or more.

[0033] Example

[0034] This invention proposes a nuclear power plant self-regulating pressure control system and method to solve the problems of inaccurate start-up and shutdown of the automatic depressurization system and ineffective separation and management of non-condensable gases in nitrogen self-regulating integrated reactor accidents.

[0035] A nuclear power plant self-regulating voltage control system, such as Figure 1The reactor includes a containment vessel 1, a pressure vessel 2, a reactor core 21, and a main heat exchanger 22. The reactor core 21 and the main heat exchanger 22 form a natural circulation loop through pipelines. The reactor is characterized by including an intermediate loop 3, a pressure relief loop 4, and monitoring instruments 6. The intermediate loop 3 is located outside the containment vessel 1 and connects the containment vessel 1 and the pressure vessel 2 through pipelines, enabling heat exchange with the main heat exchanger 22. The pressure relief loop 4 includes a water tank 41 and a pressure relief pipeline 42. The pressure relief pipeline 42 connects the top of the pressure vessel 2 and the water tank 41, and is equipped with an automatic pressure relief system 43, which can actively discharge the medium inside the pressure vessel 2 in case of an accident. The monitoring instruments 6 are located inside the pressure vessel 2.

[0036] Furthermore, such as Figure 1 As shown, the intermediate loop 3 includes a steam generator 31, a hot pipe 32, and a cold pipe 33. Both the hot pipe 32 and the cold pipe 33 are connected to the main heat exchanger 22 and the steam generator 31 to form the intermediate loop 3. The hot pipe 32 inputs the cooling medium in the intermediate loop 3 into the steam generator 31. The cold pipe 33 inputs the cooling medium in the intermediate loop 3 into the main heat exchanger 22.

[0037] Specifically, the main heat exchanger 22 is located inside the pressure vessel 2, and the steam generator 31 is located outside the containment 1. The cooling medium in the intermediate loop 3 enters the main heat exchanger 22 through the cold pipe 33, and its temperature rises after indirect heat exchange with the primary loop cooling medium. It is then transported to the steam generator 31 through the hot pipe 32. Inside the steam generator 31, the high-temperature cooling medium exchanges heat with the feedwater side, transferring heat to the cooling medium, causing the cooling medium to be heated and generate steam, thus completing the heat energy output. After heat exchange, the temperature of the cooling medium decreases, and it returns to the main heat exchanger 22 through the cold pipe 33, forming a continuous cycle.

[0038] In this embodiment, by arranging the steam generator 31 outside the containment 1, the complexity of the internal equipment arrangement of the pressure vessel 2 can be reduced, and the steam generator 31 can be easily inspected, replaced and maintained. The intermediate loop 3 also provides a unified heat exchange interface for the subsequent air-cooled loop. Under normal operating conditions, it undertakes the transfer of heat generated by the core 21. Under accident conditions, it can work in coordination with the air-cooled loop 5 to improve the flexibility and expansion capability of the system heat removal.

[0039] Furthermore, the intermediate loop 3 includes a water supply line 34 and a steam line 35, which are connected to the steam generator 31 for heat exchange with the steam generator 31.

[0040] Furthermore, such as Figure 1As shown, in this embodiment, both the water supply pipeline 34 and the steam pipeline 35 are connected to the steam generator 31, forming the cooling medium and steam delivery channel of the steam generator 31. Specifically, the cooling medium enters the steam generator 31 through the water supply pipeline 34, and indirectly exchanges heat with the high-temperature cooling medium in the intermediate loop 3 inside the steam generator 31. After absorbing heat, it undergoes a phase change to form steam. The generated steam is output to the subsequent steam turbine power generation system or other thermal energy utilization system through the steam pipeline 35, realizing the transfer of reactor thermal energy to external loads.

[0041] In this embodiment, the water supply line 34 is used to continuously replenish the cooling medium to the steam generator 31, ensuring that the steam generator 31 has a stable heat exchange medium and can maintain continuous heat exchange with the high-temperature cooling medium in the intermediate loop 3; the steam line 35 is used to continuously output the steam generated by the steam generator 31, thereby maintaining a stable temperature gradient and heat exchange capacity on both sides of the steam generator 31 and improving the heat transfer efficiency of the intermediate loop 3.

[0042] Furthermore, such as Figure 1 As shown, the intermediate loop 3 is equipped with a circulating pump 36 and isolation valves 37. The circulating pump 36 is located on the cold pipe 33 and is used to pump the cooling medium in the cold pipe 33 into the main heat exchanger 22. The two isolation valves 37 are respectively located on the hot pipe 32 and the cold pipe 33, and can be used to switch between the intermediate loop and the air-cooled loop. Under normal operating conditions, both isolation valves 37 are in the open state. Driven by the circulating pump 36, the cooling medium in the intermediate loop 3 flows sequentially through the main heat exchanger 22, the hot pipe 32, the steam generator 31, and the cold pipe. The circuit 33 forms a stable forced circulation, continuously transferring the heat absorbed by the primary loop of the main heat exchanger 22 to the steam generator 31 for heat exchange. When the reactor enters an accident condition or requires equipment maintenance, the corresponding two isolation valves 37 can be closed according to the operating conditions to isolate the steam generator 31 from the rest of the intermediate circuit 3, preventing the cooling medium from continuing to flow into the steam generator 31, thereby avoiding the continued transfer of heat to the steam generator 31 under accident conditions and reducing the continuous impact of high-temperature media on the steam generator 31 and its connected equipment.

[0043] In this embodiment, the circulating pump 36 is used to maintain the circulating flow of the intermediate loop 3 under normal operating conditions, while the isolation valve 37 is used to realize the flow path switching and equipment isolation of the intermediate loop 3. The two work together to enable the intermediate loop 3 to take into account both the high-efficiency heat transfer requirements under normal operating conditions and the safety isolation requirements under accident conditions, thereby improving the operational stability of the entire reactor thermal system.

[0044] In one embodiment, the pressure relief circuit 4 includes only a water tank 41, a pressure relief pipeline 42, and an automatic pressure relief system 43. When the automatic pressure relief system 43 discharges the medium in the pressure vessel 2 into the water tank 41, nitrogen and other media directly enter the water tank 41, some vapor is condensed and retained in the water tank 41, and the remaining nitrogen and other media directly enter the containment 1.

[0045] Furthermore, the pressure relief circuit 4 includes a water tank 41, a pressure relief pipeline 42, and an automatic pressure relief system 43. In one embodiment, the water tank 41 is located inside the containment vessel 1 and is connected to the top of the pressure vessel 2 through the pressure relief pipeline 42. A cap 414 is provided on the top of the water tank 41, and the cap 414 and the water tank 41 form a relatively closed liquid storage space. After the automatic pressure relief system 43 is activated, the high-temperature and high-pressure medium in the pressure vessel 2 enters the water tank 41 through the pressure relief pipeline 42.

[0046] Specifically, when the high-temperature and high-pressure medium enters the water tank 41, the liquid coolant flows directly into the water tank. After the steam comes into contact with the water, it condenses rapidly, transferring heat to the water, thereby reducing the temperature of the pressure relief medium and slowing down the rapid increase in pressure inside the containment under accident conditions. Since the cap 414 covers the top of the water tank 41, it can restrict the rapid diffusion of the pressure relief medium into the free space of the containment, allowing the steam to complete the condensation process inside the water tank 41 first, thereby improving the buffering and heat absorption capacity of the water tank 41 for the pressure relief medium.

[0047] Furthermore, a pressure relief valve 413 is installed on the cap 414. When the pressure in the enclosed space reaches the set value, the pressure relief valve 413 automatically opens, gradually releasing the gas in the enclosed space into the containment 1. This prevents the pressure inside the water tank 41 from continuously rising and affecting the normal operation of the automatic pressure relief system 43, and also prevents the pressure inside the containment 1 from suddenly increasing in a short period of time, which could damage the integrity of the containment 1. When the pressure returns to the set range, the pressure relief valve 413 closes again, keeping the enclosed space in a stable pressure state and improving the stability of the emergency pressure relief process.

[0048] In such Figure 1 In the preferred embodiment shown, in addition to the addition of a cap 414 and a pressure relief valve 413 to the water tank 41, the water tank 41 is further provided with a gas medium storage chamber 411 and a purification component 412; the gas medium storage chamber 411 is located at the bottom of the cap 414 and above the liquid surface of the water tank 41, forming a gas-liquid separation space together with the water tank 41; when the automatic pressure relief system 43 is activated, the mixed medium of steam, coolant, and nitrogen discharged from the pressure vessel 2 enters the water tank 41, and the airflow direction corresponding to the medium is as follows: Figure 1 As shown in the middle water tank 41, steam is condensed by the pool water to form liquid coolant and flows back into the water tank 41. Non-condensable gases, due to their lower density, gradually accumulate above the liquid surface and enter the gas medium storage chamber 411 for centralized storage.

[0049] Furthermore, the purification component 412 is disposed at the air inlet of the gas medium storage chamber 411. When the gas enters the gas medium storage chamber 411, it first passes through the purification component 412. The purification component 412 can filter the droplets, aerosols and particulate matter entrained in the gas, thereby reducing the possibility of impurities entering the gas medium storage chamber 411, improving the stability of the gas storage environment, and reducing the risk of subsequent processing equipment being contaminated. In this preferred embodiment, by setting the gas medium storage chamber 411, the non-condensable gas generated during the depressurization process can be spatially separated from the condensed liquid coolant, avoiding a large amount of gas directly diffusing into the containment 1. At the same time, it is beneficial to conduct unified monitoring, emission or further treatment of the non-condensable gas, and improve the depressurization medium management capability under accident conditions.

[0050] Furthermore, such as Figure 1 As shown in the figure, in this embodiment, the nuclear power plant self-regulating control system also includes an air-cooled circuit 5, which is connected to the intermediate circuit 3 and is used to discharge the residual heat generated by the reactor core 21 under accident conditions.

[0051] Furthermore, such as Figure 1 As shown, the air-cooled circuit 5 includes an air-cooled pipe 51 and a waste heat discharge heat exchanger 52. An air-cooled tower 54 is provided outside the waste heat discharge heat exchanger 52. Two air-cooled pipes 51 are respectively connected to the hot pipe 32 and the cold pipe 33 between the isolation valve 37 and the main heat exchanger 22. The waste heat discharge heat exchanger 52 is connected to the air-cooled pipe 51 and is used to transfer the heat in the intermediate circuit 3 to the outside air. A waste heat discharge isolation valve is provided on the air-cooled pipe 51 to control the coolant entering the main heat exchanger 22. Furthermore, when the reactor enters an accident condition, or when the steam generator 31 fails to exchange heat normally, the two isolation valves 37 corresponding to the steam generator 31 can be closed, and the residual heat discharge isolation valve 55 can be opened to isolate the intermediate loop 3 from the steam generator 31 and put the air-cooled loop 5 into operation. At this time, the high-temperature cooling medium in the intermediate loop 3 enters the air-cooled loop 51 through the hot pipe 32 and flows into the residual heat discharge heat exchanger 52. In the residual heat discharge heat exchanger 52, it exchanges heat with the outside air and releases the decay heat generated by the core 21 to the environment. After completing the heat exchange, the cooling medium returns to the cold pipe 33 through the air-cooled loop 51 and re-enters the main heat exchanger 22, realizing the continuous removal of residual heat from the core under accident conditions.

[0052] Furthermore, in this embodiment, the air-cooled circuit 5 also includes a volume compensation tank 53, which is connected to the air-cooled pipeline 51. Since the cooling medium in the air-cooled circuit 5 will undergo thermal expansion or contraction with temperature changes during accident conditions, and a small amount of cooling medium may be lost during system operation, the volume compensation tank 53 is used to compensate for the volume changes of the cooling medium in the air-cooled circuit 5. When the pressure or liquid level in the air-cooled circuit 5 decreases, the volume compensation tank 53 can replenish the cooling medium to the air-cooled pipeline 51 to maintain the liquid-filled state in the air-cooled circuit 5. When the cooling medium expands in volume due to temperature increases, the volume compensation tank 53 can also absorb some of the expanded medium, mitigating pressure fluctuations in the circuit, thereby ensuring the long-term stable operation of the air-cooled circuit 5 and improving the reliability and continuity of the waste heat discharge system under accident conditions.

[0053] Furthermore, the monitoring instrument 6 includes a water level gauge, a pressure gauge, and a steam thermometer.

[0054] To achieve the above objectives, the present invention also proposes a nuclear power plant self-regulating voltage control method, using a nuclear power plant self-regulating voltage control system according to the above description, comprising the following steps: S1: The monitoring data inside the pressure vessel 2 is collected by the monitoring instrument 6, and the monitoring data includes water level, pressure and temperature; S2: The automatic pressure relief system 43 starts or stops discharging the medium into the pressure vessel 2 into the water tank 41 according to the monitoring data.

[0055] Further, step S2 includes: setting a low water level threshold based on the vertical height of the main heat exchanger 22 within the pressure vessel 2; activating the automatic pressure relief system 43 when the water level is lower than the water level threshold; obtaining the saturation pressure based on the temperature and pressure of the pressure vessel 2; and deactivating the automatic pressure relief system 43 when the difference between the pressure and the saturation pressure is less than or equal to the allowable residual pressure difference.

[0056] Further, a low water level threshold is set based on the vertical height of the main heat exchanger 22 within the pressure vessel 2, including: The water level collected by the water level gauge inside pressure vessel 2 is monitored in real time. The expression for the low water level threshold is as follows: , Where L is the low water level threshold. For example Figure 2 The vertical distance from the top of the main heat exchanger 22 to the bottom of the pressure vessel 2 is shown. For safety margin, the safety margin is set in the range of 0.2m-0.5m. Preferably, in this embodiment, the safety margin is set to 0.3m. When the real-time monitored water level is lower than the low water level threshold, the automatic pressure relief system 43 is activated. The above steps can ensure that the automatic pressure relief system 43 is activated when the water level is still higher than the top of the main heat exchanger. At this time, the nitrogen gas chamber at the top of the pressure vessel is isolated from the heat exchange tube by liquid water or a vapor-liquid two-phase mixture. Nitrogen will not be carried into the pressure relief pipeline. The discharged medium is mainly a mixture of steam and nitrogen in the top gas chamber.

[0057] Further, based on the temperature of the pressure vessel 2, the saturation pressure is obtained. When the difference between the pressure and the saturation pressure is less than or equal to the allowable residual pressure difference, the automatic pressure relief system 43 is shut down. This includes: setting a corresponding saturation pressure for the collected temperature of the pressure vessel 2 based on a lookup table or empirical value, setting the residual pressure difference, and constructing the following inequality based on the collected pressure, saturation pressure, and residual pressure difference: , in For monitoring the primary circuit pressure, For monitoring the temperature inside the pressure vessel, For saturation pressure, The residual pressure difference ranges from 0.1 MPa to 0.2 MPa; in this embodiment, the residual pressure difference is taken as 1.5 MPa. According to Dalton's law of partial pressures, the primary circuit pressure... , For nitrogen partial pressure, when When, it indicates the partial pressure of nitrogen. The effects are negligible, meaning the nitrogen has been largely purged. Continuing to release nitrogen at this point would only result in a loss of coolant and heating of the containment vessel, offering no positive benefits. The ADS should be shut down immediately.

[0058] Furthermore, after the automatic pressure relief system 43 is shut down in the event of an accident, only a small, non-isolated rupture (maximum 3cm) remains in the system. At this point, the nitrogen has been vented, the condensation efficiency of the main heat exchanger 22 is restored, and the air-cooled loop 5 independently discharges the decay heat from the core 21. The heat is discharged to the outside of the containment 1 through the air-cooled loop 5. The pressure of the primary loop is close to the pressure of the containment 1, and the leakage from the rupture approaches zero, thus achieving long-term safety.

[0059] In summary, as can be seen from the above description, the embodiments of the present invention achieve the following technical effects: 1. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting an intermediate loop outside the containment vessel and connecting it to the containment vessel and pressure vessel via pipelines, and exchanging heat with the main heat exchanger, the heat from the primary loop can be transferred outwards through an independent intermediate loop. This reduces the risk of radioactive cooling media directly entering the external heat exchange system and isolates the internal and external heat transfer processes, improving system operational safety. Furthermore, by setting a pressure relief loop connecting the top of the pressure vessel and the water pool, and installing an automatic pressure relief system on the relief pipeline, the high-pressure medium inside the pressure vessel can be actively directed to the water pool for release in the event of an accident, reducing the internal pressure of the pressure vessel and mitigating the pressure rise caused by the release of high-temperature, high-pressure media during an accident, thus improving accident response capability. Monitoring instruments are installed to monitor the internal state parameters of the pressure vessel in real time, providing a data basis for judging the reactor's operating status and accident response. By constructing independent heat transfer paths and accident pressure release paths, this invention achieves coordinated operation of reactor heat removal and pressure regulation processes, improving the state maintenance capability and safety margin of the nitrogen self-regulating integrated reactor in the face of accident disturbances.

[0060] 2. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting up an intermediate loop outside the containment, including a steam generator, hot pipes, and cold pipes, the intermediate loop is connected to the main heat exchanger inside the pressure vessel and the external steam generator, enabling the transfer of core heat to the secondary side via the main heat exchanger. By installing a circulating pump and isolation valve on the intermediate loop, the cooling medium in the intermediate loop can be controlled to open and close according to the operating status, improving the controllability of the heat transfer process. By setting up an air-cooled loop connected to the intermediate loop and a residual heat removal heat exchanger, the decay heat generated by the core under accident conditions can be transferred to the external environment via the main heat exchanger and the intermediate loop, reducing reliance on active safety equipment and improving the reliability of long-term residual heat removal after an accident. By setting up a volume compensation tank to replenish the cooling medium in the air-cooled pipes, the volume changes in the loop caused by temperature variations or coolant loss can be mitigated, improving the operational stability of the air-cooled loop.

[0061] 3. This invention proposes a nuclear power plant self-regulating pressure control system and method. By setting a capping structure at the top of the pool, the high-temperature and high-pressure medium discharged from the pressure vessel through the automatic pressure relief system can be confined within the closed space formed by the pool and the capping structure, preventing the pressure relief medium from directly diffusing into the containment space, thereby reducing the risk of rapid pressure rise in the containment during an accident. By setting a gas medium storage chamber at the bottom of the capping structure, the gas-liquid mixture such as steam and nitrogen entering the pool can be collected centrally after condensation and separation, preventing the disorderly migration of non-condensable gases within the containment and reducing the adverse effects of non-condensable gases entering the heat exchange area on heat exchange performance. By setting a purification component at the gas medium storage chamber inlet, the gas entering the storage chamber is filtered, reducing the possibility of aerosols or other pollutants migrating with the gas during accident pressure relief, and improving the ability to manage accident products and control the environment within the containment.

[0062] 4. This invention proposes a nuclear power plant self-regulating pressure control system and method. By collecting monitoring data such as water level, pressure, and temperature inside the pressure vessel, and controlling the automatic pressure relief system to start or stop based on the monitoring data, the pressure relief process of the pressure vessel during an accident can be dynamically adjusted according to the actual operating status. This avoids the problems of premature pressure relief leading to excessive coolant loss or delayed pressure relief leading to continuous pressure rise caused by fixed pressure threshold control. By setting a low water level threshold based on the height of the main heat exchanger and determining the pressure relief termination condition based on the residual pressure difference between the pressure and the saturation pressure, the automatic pressure relief process can simultaneously consider the core cooling requirements and the primary loop depressurization status, improving the accuracy of ADS action timing judgment, reducing the risk of nitrogen entering the main heat exchanger and continuous coolant loss, thereby improving the stability of the accident response process.

[0063] The above description is merely a preferred embodiment of the present invention and is not intended to limit the invention. Various modifications and variations can be made to the present invention by those skilled in the art. Any modifications, equivalent substitutions, improvements, etc., made within the spirit and principles of the present invention should be included within the scope of protection of the present invention.

[0064] It should be noted that, in this document, relational terms such as "first" and "second" are used merely to distinguish one entity or operation from another, and do not necessarily require or imply any such actual relationship or order between these entities or operations. Furthermore, the terms "comprising," "including," or any other variations thereof are intended to cover non-exclusive inclusion, such that a process, method, article, or apparatus that comprises a list of elements includes not only those elements but also other elements not expressly listed, or elements inherent to such a process, method, article, or apparatus. Without further limitations, an element defined by the phrase "comprising one..." does not exclude the presence of other identical elements in the process, method, article, or apparatus that includes the element.

[0065] It should be noted that, in the description of this specification, the references to terms such as "one embodiment," "some embodiments," "example," "specific example," or "some examples," etc., indicate that a specific feature, structure, material, or characteristic described in connection with that embodiment or example is included in at least one embodiment or example of the present invention. In this specification, the illustrative expressions of the above terms do not necessarily refer to the same embodiment or example. Furthermore, the specific features, structures, materials, or characteristics described may be combined in any suitable manner in one or more embodiments or examples. Moreover, without contradiction, those skilled in the art can combine and integrate the different embodiments or examples described in this specification, as well as the features of different embodiments or examples.

Claims

1. A nuclear power plant self-regulating pressure control system, comprising a containment vessel (1), a pressure vessel (2), a reactor core (21), and a main heat exchanger (22), wherein the reactor core (21) and the main heat exchanger (22) form a natural circulation loop through pipelines, characterized in that: The system includes an intermediate circuit (3), a pressure relief circuit (4), and a monitoring instrument (6). The intermediate circuit (3) is located outside the containment vessel (1) and connects the containment vessel (1) and the pressure vessel (2) through a pipeline. It can exchange heat with the main heat exchanger (22). The pressure relief circuit (4) includes a water tank (41) and a pressure relief pipeline (42). The pressure relief pipeline (42) connects the top of the pressure vessel (2) and the water tank (41). The pressure relief pipeline (42) is equipped with an automatic pressure relief system (43) that can actively discharge the medium inside the pressure vessel (2) in case of an accident. The monitoring instrument (6) is located inside the pressure vessel (2).

2. The nuclear power plant self-regulating voltage control system according to claim 1, characterized in that, The intermediate circuit (3) includes a steam generator (31), a hot pipe (32), and a cold pipe (33). The hot pipe (32) and the cold pipe (33) are both connected to the main heat exchanger (22) and the steam generator (31) to form the intermediate loop (3).

3. The nuclear power plant self-regulating voltage control system according to claim 2, characterized in that, The heat pipe (32) inputs the cooling medium in the intermediate circuit (3) into the steam generator (31). The cold pipe (33) inputs the cooling medium in the intermediate loop (3) into the main heat exchanger (22).

4. The nuclear power plant self-regulating voltage control system according to claim 2, characterized in that, The intermediate circuit (3) includes a water supply pipeline (34) and a steam pipeline (35). The water supply line (34) and the steam line (35) are connected to the steam generator (31) for heat exchange with the steam generator (31).

5. The nuclear power plant self-regulating voltage control system according to claim 2, characterized in that, The intermediate circuit (3) includes a circulation pump (36) and an isolation valve (37). The circulating pump (36) is installed on the cold pipe (33); The isolation valve (37) is respectively installed on the hot pipe (32) and the cold pipe (33).

6. The nuclear power plant self-regulating voltage control system according to claim 5, characterized in that, It also includes an air-cooled circuit (5). The air-cooled circuit (5) is connected to the intermediate circuit (3) and can discharge the residual heat of the reactor core.

7. The nuclear power plant self-regulating voltage control system according to claim 6, characterized in that, The air-cooled circuit (5) includes an air-cooled pipe (51) and a waste heat discharge heat exchanger (52). The air-cooled pipe (51) is connected to the hot pipe (32) and the cold pipe (33) between the isolation valve (37) and the main heat exchanger (22); The air-cooled pipeline (51) is connected to the waste heat discharge heat exchanger (52).

8. The nuclear power plant self-regulating voltage control system according to claim 7, characterized in that, The air-cooled circuit (5) includes a volume compensation box (53). The volume compensation box (53) is connected to the air-cooled pipeline (51) and can deliver coolant to the air-cooled pipeline (51).

9. The nuclear power plant self-regulating voltage control system according to claim 1, characterized in that, The monitoring instruments (6) include a water level gauge, a pressure gauge and a steam thermometer.

10. The nuclear power plant self-regulating voltage control system according to claim 1, characterized in that, The water tank (41) includes a cap (414) and a pressure relief valve (413). The cap (414) is located on the top of the water tank (41) and can confine the medium discharged from the pressure vessel (2) within the space formed by the water tank (41) and the cap (414). The pressure relief valve (413) is located on the cap (414).

11. The nuclear power plant self-regulating voltage control system according to claim 10, characterized in that, The water tank (41) includes a gas medium storage chamber (411) and a purification component (412). The gas medium storage chamber (411) is located at the bottom of the cap (414) and can collect the gas discharged from the pressure vessel (2); The purification component (412) is located at the air inlet of the gas medium storage chamber (411).

12. A control method based on the nuclear power plant self-regulating voltage control system according to any one of claims 1-11, characterized in that, include: S1: Collect monitoring data inside the pressure vessel (2) according to the monitoring instrument (6), the monitoring data including water level, pressure and temperature; S2: The automatic pressure relief system (43) opens or stops discharging the medium into the pressure vessel (2) into the water tank (41) according to the monitoring data.

13. The method according to claim 12, characterized in that, Step S2 includes: The low water level threshold is set according to the vertical height of the main heat exchanger (22) inside the pressure vessel (2). When the water level is lower than the water level threshold, the automatic pressure relief system (43) is activated. The saturation pressure is obtained based on the temperature of the pressure vessel (2). When the difference between the pressure and the saturation pressure is less than or equal to the allowable residual pressure difference, the automatic pressure relief system (43) is shut down.

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