Method and apparatus for controlling a nuclear reactor

Abstract

The application discloses a nuclear reactor control method and device, wherein the method comprises: in the case that the duration of the nuclear reactor shutdown reaches a preset duration, obtaining the first outlet working medium state of the once-through steam generator; in the case that the first outlet working medium state indicates that the outlet of the once-through steam generator is in a non-subcooled water state, switching the operation mode of the secondary loop to a low-load feedwater mode, and performing a first control operation on the secondary loop; after the first control operation is completed, obtaining the first state information of the primary loop; in the case that the first state information of the primary loop reaches a long-term primary loop residual heat removal operation condition, controlling the secondary loop to switch to a long-term primary loop residual heat removal mode. The embodiment of the application can avoid damage to the once-through steam generator.

Description

Technical Field

[0001] This application belongs to the field of nuclear power technology, and in particular relates to a control method and apparatus for a nuclear reactor. Background Technology

[0002] During nuclear reactor operation, the over-the-loop steam generator (OTSG) serves as a critical heat exchange device between the primary and secondary loops, and its operational status directly affects the safety and stability of the nuclear reactor. The primary loop carries the heat generated by the reactor core and transfers it to the feedwater in the secondary loop via the OTG generator. The feedwater in the secondary loop undergoes preheating, evaporation, and superheating phase change processes within the OTG generator to form steam, which then powers subsequent equipment.

[0003] When a nuclear reactor accident triggers an emergency shutdown, the main feedwater system often fails due to power outages or equipment isolation. In this situation, the secondary loop residual heat removal mode is activated, continuously removing core residual heat by maintaining a low flow rate (typically 5%–10% of the rated feedwater flow rate). However, under this low-flow condition, when the OTSG outlet is in a non-subcooled water state (such as saturated water or steam), there is a significant risk of flow instability, which may cause severe pipe vibration within the OSTG, leading to mechanical damage to the OSTG and, in severe cases, even threatening the overall operational safety of the reactor. Summary of the Invention

[0004] This application provides a method and apparatus for controlling a nuclear reactor, aiming to solve the problem that existing nuclear reactor control methods easily lead to flow instability at the outlet of the DC steam generator after an emergency shutdown of the nuclear reactor, thereby causing damage to the DC steam generator.

[0005] In a first aspect, embodiments of this application provide a control method for a nuclear reactor, the nuclear reactor including a primary loop and a secondary loop, comprising: The nuclear reactor includes a primary loop and a secondary loop, comprising: If the nuclear reactor shutdown lasts for a preset duration, the first outlet working fluid status of the DC steam generator is obtained; the first outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state. When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, the operating mode of the second loop is switched to low-load feedwater mode, and a first control operation is performed on the second loop; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state. After completing the first control operation, the first state information of the loop is obtained; When the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition, the secondary loop is controlled to switch to the long-term primary loop residual heat discharge mode. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

[0006] In one embodiment of this application, switching the operating mode of the secondary circuit to a low-load water supply mode and performing a first control operation on the secondary circuit includes: Adjust the water flow rate of the second circuit to be higher than the low set value of the water flow rate; Based on the average temperature of the first circuit, the pressure of the second circuit is adjusted to a preset pressure range.

[0007] In one embodiment of this application, after obtaining the first state information of the loop, the method further includes: If the first state information of the first loop does not meet the long-term primary loop waste heat discharge operation conditions, the second outlet working fluid state of the DC steam generator and the cooling state of the first loop are obtained. When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop meets the preset cooling status, the second loop is controlled to maintain the low-load feedwater mode, and the process returns to obtaining the first status information of the first loop until the first status information of the first loop reaches the long-term primary loop waste heat discharge operation condition.

[0008] In one embodiment of this application, after obtaining the second outlet working fluid state of the DC steam generator and the cooling state of the primary loop, the method further includes: If the outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop does not meet the preset cooling status, the feedwater flow rate of the second loop is reduced, and the process returns to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0009] In one embodiment of this application, after obtaining the second outlet working fluid state of the DC steam generator and the cooling state of the primary loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop does not meet the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water flow rate of the second circuit and heat the first circuit, and return to obtain the first status information of the first circuit until the first status information of the first circuit reaches the long-term waste heat discharge operation condition of the first circuit.

[0010] In one embodiment of this application, after obtaining the flow rate amplitude of the second loop, the method further includes: If the flow rate is less than or equal to the preset flow rate, the primary loop is heated, and the process returns to obtain the first state information of the primary loop until the first state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0011] In one embodiment of this application, after obtaining the second outlet working fluid state of the DC steam generator and the cooling state of the primary loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling status of the first loop meets the preset conditions, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water supply flow rate of the second loop and return to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0012] In one embodiment of this application, after obtaining the flow rate amplitude of the second loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop meets the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is less than or equal to the preset flow rate, maintain the low-load water supply mode and return to the process of obtaining the first state information of the first loop until the first state information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0013] In one embodiment of this application, after obtaining the first outlet working fluid state of the DC steam generator, the method further includes: When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, the second status information of the first loop is obtained. When the second state information of the first loop reaches the long-term primary loop waste heat discharge operation condition, the second loop is controlled to switch to the long-term primary loop waste heat discharge mode.

[0014] Secondly, embodiments of this application provide a control device for a nuclear reactor, the nuclear reactor including a primary loop and a secondary loop, comprising: The first acquisition module is used to acquire the first outlet working fluid status of the DC steam generator when the nuclear reactor shutdown time reaches a preset time; the outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state. The first control module is used to switch the DC steam generator to a low-load feedwater mode and perform a first control operation on the second loop when the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state. The second acquisition module is used to acquire the first state information of the loop after completing the first control operation; The second control module is used to control the second loop to switch to the long-term primary loop residual heat discharge mode when the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

[0015] Thirdly, embodiments of this application provide an electronic device, including: a processor and a memory storing computer program instructions; When the processor executes the computer program instructions, it implements the nuclear reactor control method as described in the first aspect.

[0016] Fourthly, embodiments of this application provide a computer-readable storage medium storing computer program instructions, which, when executed by a processor, implement the nuclear reactor control method as described in the first aspect.

[0017] Fifthly, embodiments of this application provide a computer program product in which instructions, when executed by a processor of an electronic device, cause the electronic device to perform the nuclear reactor control method as described in the first aspect.

[0018] The nuclear reactor control method and apparatus proposed in this application, when the nuclear reactor shutdown duration reaches a preset duration, acquires the working fluid state at the first outlet of the DC steam generator. Based on the working fluid state at the first outlet, it determines whether the DC steam generator outlet is in a subcooled water state. If the DC steam generator outlet is not in a subcooled water state, it indicates that there is flow instability at the DC steam generator outlet, which can easily damage the DC steam generator. Therefore, the operating mode of the secondary loop is switched to a low-load feedwater mode, and a first control operation is performed on the secondary loop to bring the DC steam generator outlet into a subcooled water state, reducing the flow instability at the DC steam generator outlet and thus avoiding damage to the DC steam generator. Subsequently, the first state information of the primary loop is acquired. If the first state information of the primary loop meets the long-term primary loop residual heat removal operating conditions, it indicates that the primary loop is in a physical state capable of autonomously and safely removing core residual heat, and the secondary loop can be switched to the long-term primary loop residual heat removal mode. Therefore, by controlling the secondary loop after an emergency shutdown of the nuclear reactor, the flow instability at the outlet of the DC steam generator can be reduced while safely removing residual heat from the primary loop core, thus avoiding damage to the DC steam generator. Attached Figure Description

[0019] To more clearly illustrate the technical solutions of the embodiments of this application, the accompanying drawings used in the embodiments of this application will be briefly introduced below. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0020] Figure 1 This is a schematic flowchart of a nuclear reactor control method provided in an embodiment of this application; Figure 2 This is a schematic diagram of the primary loop average temperature-pressure curve provided in an embodiment of this application; Figure 3 This is a schematic diagram illustrating the implementation process of the nuclear reactor control method provided in the embodiments of this application; Figure 4 This is a schematic diagram of the structure of the control device for a nuclear reactor provided in an embodiment of this application; Figure 5 This is a schematic diagram of the structure of the electronic device provided in the embodiments of this application. Detailed Implementation

[0021] The features and exemplary embodiments of various aspects of this application will be described in detail below. To make the objectives, technical solutions, and advantages of this application clearer, the application will be further described in detail below with reference to the accompanying drawings and specific embodiments. It should be understood that the specific embodiments described herein are only intended to explain this application and not to limit it. For those skilled in the art, this application can be implemented without some of these specific details. The following description of the embodiments is merely to provide a better understanding of this application by illustrating examples.

[0022] 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..." does not exclude the presence of additional identical elements in the process, method, article, or apparatus that includes said element.

[0023] In all specific embodiments of this application, when processing data related to user identity or characteristics, such as user information, user behavior data, user historical data, and user location information, user permission or consent is obtained first. Furthermore, the collection, use, and processing of this data comply with relevant laws, regulations, and standards. Additionally, when embodiments of this application require access to sensitive personal information, separate permission or consent from the user is obtained through pop-ups or redirects to confirmation pages. Only after obtaining the user's separate permission or consent is the necessary user-related data required for the proper functioning of these embodiments obtained.

[0024] To address the problems of the prior art, this application provides a method and apparatus for controlling a nuclear reactor. The control method for a nuclear reactor provided in this application will be described first.

[0025] Figure 1 A schematic flowchart of a nuclear reactor control method according to an embodiment of this application is shown. Figure 1 As shown in the embodiment of this application, the control method for a nuclear reactor includes a primary loop and a secondary loop, comprising the following steps 101-104, wherein: Step 101: If the nuclear reactor shutdown lasts for a preset duration, obtain the first outlet working fluid status of the DC steam generator; the first outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state.

[0026] The preset duration can be set according to the actual situation and is not specifically limited here. For example, considering that human intervention is not possible for a period of time after a nuclear reactor accident, the preset duration can be set to 30 minutes.

[0027] The state of the first outlet working medium can be obtained by installing temperature and pressure sensors on the outlet pipe of the DC steam generator: based on the detected outlet working medium temperature and pressure, combined with the thermodynamic property table of the working medium, it can be determined whether the working medium is in a subcooled water state.

[0028] In this embodiment, the operating status of the nuclear reactor after the accident is divided into a short-term stage (within the preset time period after the accident) and a long-term stage (after the preset time period after the accident) according to a preset time period.

[0029] In the short term of an accident, the core heat of the primary loop can be removed through emergency shutdown and secondary loop residual heat removal mode. When the accident lasts for a long period, the working fluid status of the first outlet of the DC steam generator is obtained, and the secondary loop is controlled based on the working fluid status of the first outlet of the DC steam generator.

[0030] Step 102: When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, the operating mode of the second loop is switched to low-load feedwater mode, and a first control operation is performed on the second loop; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state.

[0031] Non-subcooled water state includes two-phase flow state and superheated steam state. That is, if the outlet of the DC steam generator is in two-phase flow state or superheated steam state, the outlet of the DC steam generator can be in non-subcooled water state.

[0032] Low-load water supply mode can refer to the mode in which the secondary circuit operates at a water supply flow rate lower than the set value. The specific value of the low water supply flow rate set value can depend on the design of the OTSG. The basis for the value is that the OTSG will not experience flow instability when the water supply flow rate is higher than the low set value during normal operation. For example, according to a certain OTSG flow instability test, when the OTSG water supply flow rate is lower than 15% of the rated flow rate, its water supply flow rate fluctuation exceeds the equipment's tolerance range. The low water supply flow rate set value of this OTSG is 15% of the rated flow rate.

[0033] The first control operation can be an operation that controls the water flow rate of the secondary loop, an operation that controls the pressure of the secondary loop, or an operation that controls both the water flow rate and pressure of the secondary loop. For example, the first control operation could be to increase the water flow rate of the secondary loop and to increase the pressure of the secondary loop.

[0034] Step 103: After completing the first control operation, obtain the first state information of the first loop.

[0035] The first state information of the primary loop may include at least one of the following: temperature, pressure, coolant flow rate, and core residual heat power.

[0036] Step 104: When the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition, control the secondary loop to switch to the long-term primary loop residual heat discharge mode. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

[0037] When the first state information of the primary loop includes the temperature and pressure of the primary loop, the long-term waste heat discharge operation conditions may include the temperature of the primary loop being lower than a preset temperature threshold and the pressure of the primary loop being lower than a preset pressure threshold. The preset temperature threshold and preset pressure threshold can be set according to actual conditions and are not specifically limited here.

[0038] In some embodiments, when the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition, it indicates that the primary loop does not need to be cooled by a DC steam generator to safely discharge the core residual heat; therefore, the secondary loop is switched to the long-term primary loop residual heat discharge mode to perform long-term stable cooling of the primary loop.

[0039] In this embodiment, when the reactor shutdown lasts for a preset duration, the working fluid status at the first outlet of the DC steam generator is acquired. The status of the working fluid at the first outlet is used to determine whether the DC steam generator outlet is in a subcooled water state. If the DC steam generator outlet is not in a subcooled water state, it indicates that there is flow instability at the DC steam generator outlet, which can easily damage the DC steam generator. Therefore, the operating mode of the secondary loop is switched to low-load feedwater mode, and a first control operation is performed on the secondary loop to bring the DC steam generator outlet into a subcooled water state, reducing flow instability at the DC steam generator outlet and thus avoiding damage to the DC steam generator. Subsequently, the first state information of the primary loop is acquired. If the first state information of the primary loop meets the long-term primary loop residual heat removal operating conditions, it indicates that the primary loop is in a physical state capable of autonomously and safely removing core residual heat, and the secondary loop can be switched to the long-term primary loop residual heat removal mode. Therefore, by controlling the secondary loop after an emergency shutdown of the nuclear reactor, the flow instability at the outlet of the DC steam generator can be reduced while safely removing residual heat from the primary loop core, thus avoiding damage to the DC steam generator.

[0040] In some embodiments, switching the operating mode of the secondary loop to a low-load water supply mode and performing a first control operation on the secondary loop may include: Adjust the water flow rate of the second circuit to be higher than the low set value of the water flow rate; Based on the average temperature of the first circuit, the pressure of the second circuit is adjusted to a preset pressure range.

[0041] The preset pressure range can be calculated from the saturation pressure curve corresponding to the primary loop temperature. By controlling the secondary loop pressure, the saturation temperature of the working fluid in the secondary loop is matched with the temperature in the primary loop, ensuring the stability of the heat exchange process in the DC steam generator and avoiding premature vaporization of the working fluid due to excessive pressure or insufficient heat exchange efficiency due to excessively low pressure.

[0042] In some implementations, during low-load water supply mode operation, the water supply pump is activated, and the water supply flow rate is adjusted to exceed a lower setpoint (e.g., 15% of the rated water supply flow rate). The OTSG pressure is then adjusted within the shaded range based on the primary loop average temperature-pressure curve. The primary loop average temperature-pressure curve can be found in [reference needed]. Figure 2 , Figure 2 The blue curve represents the pressure curve of saturated water at different temperatures, while the red area represents the shaded region.

[0043] It should be noted that the primary loop average temperature-pressure curve can be predetermined based on the design characteristics of the DC steam generator through thermal-hydraulic safety analysis.

[0044] In some implementations, by controlling the pressure of the secondary loop, the saturation temperature of the working fluid in the secondary loop is matched with the temperature of the primary loop, ensuring the stability of the heat exchange process in the DC steam generator and avoiding premature vaporization of the working fluid due to excessively high pressure or insufficient heat exchange efficiency due to excessively low pressure.

[0045] In this embodiment, by adjusting the feedwater flow rate to a value greater than the lower setpoint, sufficient cooling capacity is ensured, preventing non-subcooled water from appearing at the outlet of the DC steam generator. Simultaneously, by adjusting the pressure to a preset range, temperature and pressure matching between the primary and secondary loops is guaranteed, improving heat exchange efficiency. This effectively avoids flow instability within the DC steam generator, thereby reducing the risk of pipe vibration within the generator.

[0046] If the initial state information of the primary loop does not meet the operating conditions for long-term residual heat removal, directly performing subsequent operations may cause fluctuations in the state of the nuclear reactor, which may lead to flow instability or insufficient residual heat removal efficiency.

[0047] This application further proposes that, after obtaining the first state information of the primary loop, the control method for the nuclear reactor provided in the embodiments of this application may further include: If the first state information of the first loop does not meet the long-term primary loop waste heat discharge operation conditions, the second outlet working fluid state of the DC steam generator and the cooling state of the first loop are obtained. When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop meets the preset cooling status, the second loop is controlled to maintain the low-load feedwater mode, and the process returns to obtaining the first status information of the first loop until the first status information of the first loop reaches the long-term primary loop waste heat discharge operation condition.

[0048] The state of the second outlet working medium can be monitored by temperature or pressure sensors to detect the phase state of the working medium at the outlet of the DC steam generator.

[0049] The preset cooling state can be determined based on the cooling rate of the primary loop. For example, when the cooling rate of the primary loop is less than a rate threshold, it is a safe cooling state, i.e., the preset cooling state. The rate threshold can be set according to the actual situation and is not specifically limited here.

[0050] Specifically, when the first state information of the primary loop does not meet the long-term residual heat removal conditions, the system first checks whether the outlet of the once-through steam generator is in a subcooled water state. If the outlet working fluid is in a subcooled water state and the primary loop cooling rate is within a preset range (e.g., 0.5℃-1℃ per minute), the system determines that the current thermodynamic state of the nuclear reactor is stable and continues to maintain the low-load feedwater mode. At this time, by maintaining the low-load feedwater, the system ensures that the inside of the once-through steam generator maintains a single-phase flow state, avoiding pipe vibrations caused by non-subcooled water.

[0051] Simultaneously, the temperature and pressure parameters of the primary loop (i.e., the first state information) are collected in real time. When the first state information meets the conditions for long-term waste heat discharge, the mode switching is triggered, and the system switches to the long-term waste heat discharge mode.

[0052] In this embodiment, the operating mode of the secondary loop is dynamically adjusted according to the working fluid state at the outlet of the DC steam generator and the cooling status of the primary loop, so as to ensure that the outlet of the DC steam generator is always kept in the subcooled water state, avoiding the risk of flow instability, and at the same time achieving stable cooling of the primary loop.

[0053] When the outlet of the DC steam generator is in a subcooled water state and the cooling state of the primary loop does not meet the preset cooling state, maintaining the low-load feedwater mode may result in the primary loop residual heat not being effectively discharged, thereby affecting the establishment of the core safety state.

[0054] Therefore, the nuclear reactor control method provided in this application embodiment may further include: If the outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop does not meet the preset cooling status, the feedwater flow rate of the second loop is reduced, and the process returns to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0055] Specifically, when it is detected that the outlet of the DC steam generator is in a subcooled water state but the cooling rate of the primary loop is lower than the preset value, it indicates that the cooling rate of the primary loop is too fast, which may pose a safety hazard. Therefore, the feedwater flow rate of the secondary loop is reduced to decrease the heat absorbed by the secondary loop from the primary loop, thereby reducing the cooling rate of the primary loop and allowing it to cool down rapidly while ensuring safety. The process then returns to the step of obtaining the first state information of the primary loop until the first state information of the primary loop meets the long-term primary loop waste heat discharge operation conditions.

[0056] For example, the water flow rate of the second loop can be reduced by 5% to slow down the cooling rate of the first loop, and the status information of the first loop can be continuously monitored. When the pressure, temperature and flow parameters of the first loop meet the conditions for long-term waste heat discharge, the operating mode can be switched immediately.

[0057] By dynamically adjusting the matching relationship between the feedwater flow rate and the primary loop cooling status, the problem of delayed discharge of residual heat from the primary loop due to excessive heat absorption in the secondary loop is avoided, ensuring the rapid establishment of a safe core state.

[0058] In this embodiment, when the DC steam generator outlet is in a subcooled water state but the primary loop cooling status does not meet expectations, the cooling rate of the primary loop is adjusted by reducing the secondary loop feedwater flow rate. This is to avoid overcooling the primary loop and prevent thermal shock from damaging the equipment. At the same time, by continuously monitoring the primary loop status and making timely adjustments, the safety and stability of the nuclear reactor during shutdown are ensured.

[0059] When the outlet working fluid is in a non-subcooled state and the cooling state of the primary loop does not meet the preset cooling state, it indicates that the secondary loop is in a non-subcooled water state and the cooling rate of the primary loop is relatively fast. At this time, there is a difference between the states of the primary and secondary loops, which can be further determined by obtaining the flow rate amplitude of the secondary loop. Therefore, the nuclear reactor control method provided in this application embodiment may further include: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop does not meet the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water flow rate of the second circuit and heat the first circuit, and return to obtain the first status information of the first circuit until the first status information of the first circuit reaches the long-term waste heat discharge operation condition of the first circuit.

[0060] Flow amplitude can refer to the range of flow fluctuations, specifically the maximum change in water supply flow rate from its average value. Flow amplitude characterizes the severity of instability in the secondary loop flow.

[0061] The preset range can be set according to actual conditions and is not specifically limited here. Specifically, the value of the preset range depends on the design of the OTSG. It refers to the fact that when the OTSG feedwater flow rate fluctuation exceeds a certain range, it may cause flow and pressure oscillations and changes in wall temperature circulation, leading to thermal fatigue and mechanical vibration problems in the structure, affecting the safe operation of the equipment. For example, according to a certain OTSG flow instability test, a flow rate fluctuation exceeding 10% will endanger equipment safety.

[0062] Specifically, in the dual abnormal scenario where the outlet working fluid status indicator of the DC steam generator is in a non-subcooled water state and the cooling state of the primary loop does not meet the preset cooling state, the flow rate amplitude of the secondary loop is obtained. If the flow rate amplitude of the secondary loop is greater than the preset amplitude, it indicates that the feedwater flow rate of the secondary loop needs to be increased to alleviate the non-subcooled water phenomenon at the OTG outlet. However, this will further accelerate the cooling of the primary loop. Therefore, this application increases the feedwater flow rate of the secondary loop while heating the primary loop through the voltage regulator electric heater to avoid the problem of the primary loop cooling rate being too fast. Then, the step of obtaining the first state information of the primary loop is executed until the first state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0063] In this embodiment, by adjusting the secondary circuit water supply flow and heating the primary circuit, while ensuring the safe cooling of the primary circuit, the flow instability at the outlet of the DC steam generator is effectively reduced, and the risk of mechanical damage caused by pipeline vibration is avoided.

[0064] When the flow rate of the secondary loop does not exceed the preset range, continuing to increase the water supply flow rate may cause the operating parameters of the secondary loop to deviate from the stable range. At the same time, the primary loop has the problem of excessive cooling rate. In this case, the problem of excessive cooling rate of the primary loop can be solved by directly heating the primary loop.

[0065] Therefore, the nuclear reactor control method provided in this application embodiment may further include: If the flow rate is less than or equal to the preset flow rate, the primary loop is heated, and the process returns to obtain the first state information of the primary loop until the first state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0066] Specifically, in the dual abnormal scenario where the outlet working fluid status indicator of the DC steam generator is in a non-subcooled water state and the cooling status of the primary loop does not meet the preset cooling status, the flow rate amplitude of the secondary loop is obtained. If the flow rate amplitude of the secondary loop is less than the preset amplitude, it indicates that further increasing the feedwater flow rate of the secondary loop may cause the operating parameters of the secondary loop to deviate from the stable range. However, the primary loop still has the problem of excessive cooling rate. Therefore, the primary loop is heated only by the voltage regulator electric heater to avoid excessive cooling rate of the primary loop, and the execution of the step of obtaining the first status information of the primary loop is returned until the first status information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0067] In this embodiment, when the flow rate of the secondary loop is limited, the primary loop is heated to accelerate the achievement of long-term waste heat discharge conditions. This ensures the safety of the primary loop while maintaining a stable flow rate in the secondary loop.

[0068] When the outlet of the DC steam generator is in a non-subcooled water state and the cooling state of the primary loop meets the preset conditions, maintaining the low-load feedwater mode may result in insufficient feedwater flow in the secondary loop, which will not be able to effectively regulate the outlet working fluid state and thus affect the stability of the OTG.

[0069] Therefore, the nuclear reactor control method provided in this application embodiment may further include: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling status of the first loop meets the preset conditions, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water supply flow rate of the second loop and return to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0070] Specifically, when the outlet of the DC steam generator is in a non-subcooled water state, it indicates that the feedwater flow rate of the second loop needs to be increased to alleviate the non-subcooled water state of the second loop. At this time, if the cooling rate of the first loop has met the preset conditions, it is necessary to determine whether the feedwater flow rate of the second loop can be increased based on the flow rate amplitude of the second loop. If the flow rate amplitude is greater than the preset amplitude, the feedwater flow rate of the second loop is increased, and the process returns to obtain the first state information of the first loop until the first state information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0071] In this embodiment, the secondary circuit feedwater flow rate is dynamically adjusted based on the outlet status of the DC steam generator and the primary circuit cooling situation. When a non-subcooled water state occurs, increasing the feedwater flow rate helps to adjust the non-subcooled water state at the OTG outlet, which in turn helps to restore the DC steam generator outlet to a subcooled water state as quickly as possible.

[0072] When the flow rate of the secondary loop is less than the preset threshold, directly increasing the water supply flow rate of the secondary loop may cause flow instability risk and affect the safety of the waste heat discharge process.

[0073] Therefore, the nuclear reactor control method provided in this application embodiment may further include: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop meets the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is less than or equal to the preset flow rate, maintain the low-load water supply mode and return to the process of obtaining the first state information of the first loop until the first state information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0074] Specifically, when it is detected that the outlet of the DC steam generator is in a non-subcooled water state but the primary loop has achieved stable cooling, the flow rate amplitude of the secondary loop is acquired. If the flow rate amplitude of the secondary loop does not exceed a preset threshold, it indicates that the secondary loop's regulation capability has reached its limit, but the nuclear reactor is still under control. At this time, the existing low flow rate operating parameters are maintained to avoid pressure fluctuations on the secondary side of the steam generator due to a forced increase in flow rate. The process then returns to acquiring the first state information of the primary loop until the first state information of the primary loop meets the long-term primary loop waste heat removal operating conditions.

[0075] In this embodiment, stable control of the secondary loop is achieved by accurately determining the flow fluctuation threshold. Under the premise of meeting the cooling requirements of the primary loop, oscillations caused by excessive flow regulation under low load conditions are avoided, ensuring that the nuclear reactor can safely transition to the autonomous waste heat removal stage, thereby reducing the risk of non-subcooled water instability in the DC steam generator during the transition process.

[0076] When the outlet working fluid status indicator of the DC steam generator is in the subcooled water state, it indicates that the cooling of the entire nuclear reactor is relatively safe and stable. Therefore, the nuclear reactor control method provided in this application embodiment may further include: When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, the second status information of the first loop is obtained. When the second state information of the first loop reaches the long-term primary loop waste heat discharge operation condition, the second loop is controlled to switch to the long-term primary loop waste heat discharge mode.

[0077] Specifically, first obtain the working fluid status at the outlet of the OTSG. If the outlet of the OTSG is subcooled water, the operation of the secondary loop waste heat discharge mode can continue until the long-term primary loop waste heat discharge operation conditions are met.

[0078] In this embodiment, the operating mode of the secondary loop is switched according to the working fluid state at the outlet of the OTG. When the working fluid state at the outlet of the OTG is supercooled water, the secondary loop residual heat removal mode is maintained for a short period of time so as to safely remove the core residual heat of the primary loop when the OTG outlet is stable.

[0079] The following describes the implementation process of the nuclear reactor control method provided in the embodiments of this application: This application embodiment divides the reactor operation status after an accident into short-term control (within a preset time period after the accident) and long-term control (after a preset time period after the accident). In the short term, the reactor core residual heat is still removed through emergency shutdown and secondary loop residual heat removal system. When the accident occurs 30 minutes later, the operator intervenes and starts the low-load feedwater system to increase the feedwater flow rate. Subsequently, by dynamically adjusting the secondary loop feedwater flow rate and pressure, the OTSG is kept to operate at a large flow rate and the OTSG outlet is kept to be subcooled water or stable vapor-liquid two-phase flow to eliminate the risk of flow instability.

[0080] Specifically: In the short term (within the preset time frame after an accident), the residual heat in the reactor core will still be removed through emergency shutdown and automatic control of the secondary loop residual heat removal system.

[0081] Upon receiving an emergency reactor shutdown signal, the reactor can be started and maintained at a low flow rate (e.g., 5% to 10% of the rated feedwater flow rate) through emergency shutdown, turbine tripping, main feedwater isolation, and secondary loop waste heat removal mode, thus establishing a small loop of secondary loop waste heat removal mode.

[0082] And monitor relevant parameters of the nuclear reactor, including: Primary loop: temperature, pressure, flow rate; Secondary loop: The secondary loop waste heat discharge system flow rate, OTSG outlet temperature and pressure, and determine the working fluid status at the OTSG outlet (subcooled water, two-phase flow, superheated steam).

[0083] In the long-term phase (after a preset time since the accident), adjustments are made according to the "phased active control method" (i.e., the nuclear reactor control method provided in the embodiments of this application), such as... Figure 3 As shown, Figure 3 This is a schematic diagram illustrating the implementation process of the nuclear reactor control method provided in this application embodiment.

[0084] Step 1: First, confirm the working fluid status of the OTG outlet based on the monitoring parameters. If the OTG outlet is subcooled water, the secondary loop waste heat discharge mode can continue to operate automatically until the long-term primary loop waste heat discharge operation conditions are met. If the OTG outlet is not subcooled water (vapor-liquid two-phase, superheated steam), the secondary loop waste heat discharge system is shut down and switched to low-load feedwater mode.

[0085] Step 2: During low-load water supply mode operation, the water supply pump is started, and the water supply flow rate is adjusted to be greater than the low setpoint (i.e., the low setpoint water flow rate) (e.g., 15% of the rated water supply flow rate). The OTSG pressure is then adjusted within the shaded area according to the primary loop average temperature-pressure curve. The primary loop average temperature-pressure curve is shown below. Figure 2 As shown, the blue curve represents the pressure curve of saturated water at different temperatures.

[0086] Step 3: Monitor reactor operating parameters in real time to determine whether the reactor has met the operating conditions for long-term primary loop residual heat removal. If so, switch to long-term primary loop residual heat removal mode; otherwise, proceed to the next step.

[0087] Step 4: Simultaneously determine the OTSG outlet status and the primary loop cooling rate: OTSG outlet status: If the outlet is subcooled water, maintain the status quo and return to step 3 to continue monitoring reactor operating parameters; if it is not subcooled water (vapor-liquid two-phase, superheated steam), the flow rate fluctuation amplitude is automatically monitored by the OTSG feedwater flow meter to see if it exceeds the fluctuation threshold (e.g., feedwater flow rate fluctuation amplitude exceeds 10%). If the flow rate fluctuation does not exceed the fluctuation threshold, maintain the status quo and return to step 3 to continue monitoring reactor operating parameters. If it exceeds the fluctuation threshold, an alarm is automatically triggered, the feedwater flow rate is increased by 5%, and then the process returns to step 3 to continue monitoring reactor operating parameters. Primary loop cooling rate: If the primary loop cooling rate does not exceed the rate threshold (e.g., 50℃ / h), maintain the status quo and return to step three to continue monitoring reactor operating parameters; if it exceeds the threshold, determine whether the OTSG outlet is subcooled water. If it is subcooled water, reduce the feedwater flow rate by 5% and then return to step three to continue monitoring reactor operating parameters; if it is not subcooled water, automatically monitor the flow rate fluctuation amplitude according to the OTSG feedwater flow meter to see if it exceeds the fluctuation threshold. If the flow rate fluctuation does not exceed the fluctuation threshold, turn on the pressurizer electric heater; if the flow rate fluctuation exceeds the fluctuation threshold, increase the feedwater flow rate by 5% and turn on the pressurizer electric heater, and then return to step three to continue monitoring reactor operating parameters.

[0088] Figure 4 A structural diagram of the control device for a nuclear reactor provided in an embodiment of this application is shown. Figure 4 As shown, the control device 300 for the nuclear reactor includes a primary loop and a secondary loop, comprising: The first acquisition module 301 is used to acquire the first outlet working fluid status of the DC steam generator when the nuclear reactor shutdown time reaches a preset time; the outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state. The first control module 302 is used to switch the DC steam generator to a low-load feedwater mode and perform a first control operation on the second loop when the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state. The second acquisition module 303 is used to acquire the first state information of the loop after completing the first control operation; The second control module 304 is used to control the second loop to switch to the long-term primary loop residual heat discharge mode when the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

[0089] In one embodiment of this application, the first control module 302 includes: The first adjustment submodule is used to adjust the water supply flow rate of the second circuit to a value higher than the low set value of the water supply flow rate; The second adjustment submodule is used to adjust the pressure of the second circuit to a preset pressure range based on the average temperature of the first circuit.

[0090] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The third acquisition module is used to acquire the second outlet working fluid status of the DC steam generator and the cooling status of the first loop when the first state information of the first loop does not meet the long-term primary loop waste heat discharge operation conditions. The third control module is used to control the second loop to maintain the low-load feedwater mode when the outlet working fluid status indicates that the outlet of the DC steam generator is in the subcooled water state and the cooling status of the first loop meets the preset cooling status, and then return to the process of obtaining the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0091] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The fourth control module, when the second outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state and the cooling status of the first loop does not meet the preset cooling status, reduces the feedwater flow of the second loop and returns to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0092] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The fourth acquisition module is used to acquire the flow rate amplitude of the second loop when the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state and the cooling state of the first loop does not meet the preset cooling state. The fifth control module is used to increase the water supply flow of the second circuit and heat the first circuit when the flow rate amplitude is greater than the preset amplitude, and return to obtain the first status information of the first circuit until the first status information of the first circuit reaches the long-term waste heat discharge operation condition of the first circuit.

[0093] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The sixth control module is used to heat the primary loop when the flow rate amplitude is less than or equal to the preset amplitude, and return to obtain the first state information of the primary loop until the first state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0094] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The fifth acquisition module is used to acquire the flow rate amplitude of the second loop when the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state and the cooling status of the first loop meets the preset conditions. The seventh control module is used to increase the water supply flow of the second loop when the flow rate amplitude is greater than the preset amplitude, and return to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0095] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The sixth acquisition module is used to acquire the flow rate amplitude of the second loop when the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state and the cooling state of the first loop meets the preset cooling state. The eighth control module is used to maintain the low-load water supply mode when the flow rate amplitude is less than or equal to the preset amplitude, and return to the process of obtaining the first state information of the first loop until the first state information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

[0096] In one embodiment of this application, the control device 300 for the nuclear reactor further includes: The seventh acquisition module is used to acquire the second state information of the primary loop when the first outlet working fluid state indicates that the outlet of the DC steam generator is in a subcooled water state. The ninth control module is used to control the second loop to switch to the long-term primary loop waste heat discharge mode when the second state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

[0097] The nuclear reactor control device 300 provided in this application embodiment can realize the various processes implemented in the aforementioned nuclear reactor control method embodiment and achieve the same technical effect. To avoid repetition, it will not be described again here.

[0098] Figure 5 A schematic diagram of the hardware structure of the electronic device provided in an embodiment of this application is shown.

[0099] The electronic device may include a processor 401 and a memory 402 storing computer program instructions.

[0100] Specifically, the processor 401 may include a central processing unit (CPU), an application-specific integrated circuit (ASIC), or one or more integrated circuits that can be configured to implement the embodiments of this application.

[0101] Memory 402 may include mass storage for data or instructions. For example, and not limitingly, memory 402 may include a hard disk drive (HDD), floppy disk drive, flash memory, optical disk, magneto-optical disk, magnetic tape, or Universal Serial Bus (USB) drive, or a combination of two or more of these. Where appropriate, memory 402 may include removable or non-removable (or fixed) media. Where appropriate, memory 402 may be internal or external to the integrated gateway disaster recovery device. In a particular embodiment, memory 402 is non-volatile solid-state memory.

[0102] Memory may include read-only memory (ROM), random access memory (RAM), disk storage media devices, optical storage media devices, flash memory devices, and electrical, optical, or other physical / tangible memory storage devices. Therefore, typically, memory includes one or more tangible (non-transitory) computer-readable storage media (e.g., memory devices) encoded with software including computer-executable instructions, and when the software is executed (e.g., by one or more processors), it is operable to perform the operations described with reference to the methods according to the first or second aspect of this disclosure.

[0103] The processor 401 implements any of the information auditing methods described in the above embodiments by reading and executing computer program instructions stored in the memory 402.

[0104] In one example, the electronic device may also include a communication interface 403 and a bus 410. For example, Figure 5 As shown, the processor 401, memory 402, and communication interface 403 are connected through bus 410 and complete communication with each other.

[0105] The communication interface 403 is mainly used to realize communication between various modules, devices, units and / or equipment in the embodiments of this application.

[0106] Bus 410 includes hardware, software, or both, that couples components of an information auditing method or verification device together. For example, and not as a limitation, the bus may include an Accelerated Graphics Port (AGP) or other graphics bus, an Enhanced Industry Standard Architecture (EISA) bus, a Front Side Bus (FSB), HyperTransport (HT) interconnect, an Industry Standard Architecture (ISA) bus, an Infinite Bandwidth Interconnect, a Low Pin Count (LPC) bus, a memory bus, a Microchannel Architecture (MCA) bus, a Peripheral Component Interconnect (PCI) bus, a PCI-Express (PCI-X) bus, a Serial Advanced Technology Attachment (SATA) bus, a Video Electronics Standards Association Local (VLB) bus, or other suitable buses, or combinations of two or more of these. Where appropriate, bus 410 may include one or more buses. Although specific buses are described and illustrated in embodiments of this application, any suitable bus or interconnect is contemplated herein.

[0107] Furthermore, in conjunction with the nuclear reactor control methods described in the above embodiments, this application embodiment can provide a computer storage medium for implementation. This computer storage medium stores computer program instructions; when these computer program instructions are executed by a processor, they implement any of the nuclear reactor control methods described in the above embodiments.

[0108] Alternatively, this application embodiment can provide a computer program product for implementation, wherein the instructions in the computer program product, when executed by the processor of an electronic device, cause the electronic device to implement any of the nuclear reactor control methods in the above embodiments.

[0109] It should be clarified that this application is not limited to the specific configurations and processes described above and shown in the figures. For the sake of brevity, detailed descriptions of known methods are omitted here. In the above embodiments, several specific steps are described as examples. However, the method process of this application is not limited to the specific steps described. Those skilled in the art can make various changes, modifications, and additions, or change the order of steps, after understanding the spirit of this application.

[0110] The functional blocks shown in the above-described structural diagram can be implemented as hardware, software, firmware, or a combination thereof. When implemented in hardware, they can be, for example, electronic circuits, application-specific integrated circuits (ASICs), appropriate firmware, plug-ins, function cards, etc. When implemented in software, the elements of this application are programs or code segments used to perform the required tasks. Programs or code segments can be stored on a machine-readable medium or transmitted over a transmission medium or communication link via data signals carried on a carrier wave. "Machine-readable medium" can include any medium capable of storing or transmitting information. Examples of machine-readable media include electronic circuits, semiconductor memory devices, ROM, flash memory, erasable ROM (EROM), floppy disks, CD-ROMs, optical disks, hard disks, fiber optic media, radio frequency (RF) links, etc. Code segments can be downloaded via computer networks such as the Internet, intranets, etc.

[0111] It should also be noted that the exemplary embodiments mentioned in this application describe methods or systems based on a series of steps or apparatus. However, this application is not limited to the order of the above steps; that is, the steps can be performed in the order mentioned in the embodiments, or in a different order, or several steps can be performed simultaneously.

[0112] The aspects of this disclosure have been described above with reference to flowchart illustrations and / or block diagrams of methods, apparatus (systems), and computer program products according to embodiments of this disclosure. It should be understood that each block in the flowchart illustrations and / or block diagrams, and combinations of blocks in the flowchart illustrations and / or block diagrams, can be implemented by computer program instructions. These computer program instructions can be provided to a processor of a general-purpose computer, a special-purpose computer, or other programmable data processing apparatus to produce a machine such that these instructions, executable via the processor of the computer or other programmable data processing apparatus, enable the implementation of the functions / actions specified in one or more blocks of the flowchart illustrations and / or block diagrams. Such a processor can be, but is not limited to, a general-purpose processor, a special-purpose processor, a special application processor, or a field-programmable logic circuit. It is also understood that each block in the block diagrams and / or flowcharts, and combinations of blocks in the block diagrams and / or flowcharts, can also be implemented by special-purpose hardware performing the specified functions or actions, or can be implemented by a combination of special-purpose hardware and computer instructions.

[0113] The above description is merely a specific implementation of this application. Those skilled in the art will clearly understand that, for the sake of convenience and brevity, the specific working processes of the systems, modules, and units described above can be referred to the corresponding processes in the foregoing method embodiments, and will not be repeated here. It should be understood that the protection scope of this application is not limited thereto. Any person skilled in the art can easily conceive of various equivalent modifications or substitutions within the technical scope disclosed in this application, and these modifications or substitutions should all be covered within the protection scope of this application.

Claims

1. A method for controlling a nuclear reactor, characterized in that, The nuclear reactor includes a primary loop and a secondary loop, comprising: If the nuclear reactor shutdown lasts for a preset duration, the first outlet working fluid status of the DC steam generator is obtained; the first outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state. When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, the operating mode of the second loop is switched to low-load feedwater mode, and a first control operation is performed on the second loop; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state. After completing the first control operation, the first state information of the loop is obtained; When the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition, the secondary loop is controlled to switch to the long-term primary loop residual heat discharge mode. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

2. The method according to claim 1, characterized in that, The step of switching the operating mode of the secondary circuit to a low-load water supply mode and performing a first control operation on the secondary circuit includes: Adjust the water flow rate of the second circuit to be higher than the low set value of the water flow rate; Based on the average temperature of the first circuit, the pressure of the second circuit is adjusted to a preset pressure range.

3. The method according to claim 1, characterized in that, After obtaining the first state information of the loop, the method further includes: If the first state information of the first loop does not meet the long-term primary loop waste heat discharge operation conditions, the second outlet working fluid state of the DC steam generator and the cooling state of the first loop are obtained. When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop meets the preset cooling status, the second loop is controlled to maintain the low-load feedwater mode, and the process returns to obtaining the first status information of the first loop until the first status information of the first loop reaches the long-term primary loop waste heat discharge operation condition.

4. The method according to claim 3, characterized in that, After obtaining the working fluid state at the second outlet of the DC steam generator and the cooling state of the primary loop, the method further includes: If the outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, and the cooling status of the first loop does not meet the preset cooling status, the feedwater flow rate of the second loop is reduced, and the process returns to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

5. The method according to claim 3, characterized in that, After obtaining the working fluid state at the second outlet of the DC steam generator and the cooling state of the primary loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop does not meet the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water flow rate of the second circuit and heat the first circuit, and return to obtain the first status information of the first circuit until the first status information of the first circuit reaches the long-term waste heat discharge operation condition of the first circuit.

6. The method according to claim 5, characterized in that, After obtaining the flow rate amplitude of the second loop, the method further includes: If the flow rate is less than or equal to the preset flow rate, the primary loop is heated, and the process returns to obtain the first state information of the primary loop until the first state information of the primary loop reaches the long-term primary loop waste heat discharge operation condition.

7. The method according to claim 3, characterized in that, After obtaining the working fluid state at the second outlet of the DC steam generator and the cooling state of the primary loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling status of the first loop meets the preset conditions, the flow rate amplitude of the second loop is obtained. If the flow rate is greater than the preset value, increase the water supply flow rate of the second loop and return to obtain the first status information of the first loop until the first status information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

8. The method according to claim 7, characterized in that, After obtaining the flow rate amplitude of the second loop, the method further includes: When the second outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state, and the cooling state of the first loop meets the preset cooling state, the flow rate amplitude of the second loop is obtained. If the flow rate is less than or equal to the preset flow rate, maintain the low-load water supply mode and return to the process of obtaining the first state information of the first loop until the first state information of the first loop reaches the long-term waste heat discharge operation condition of the first loop.

9. The method according to claim 1, characterized in that, After obtaining the first outlet working fluid state of the DC steam generator, the method further includes: When the first outlet working fluid status indicates that the outlet of the DC steam generator is in a subcooled water state, the second status information of the first loop is obtained. When the second state information of the first loop reaches the long-term primary loop waste heat discharge operation condition, the second loop is controlled to switch to the long-term primary loop waste heat discharge mode.

10. A control device for a nuclear reactor, characterized in that, The nuclear reactor includes a primary loop and a secondary loop, comprising: The first acquisition module is used to acquire the first outlet working fluid status of the DC steam generator when the nuclear reactor shutdown time reaches a preset time; the outlet working fluid status is used to indicate whether the outlet of the DC steam generator is in a subcooled water state. The first control module is used to switch the DC steam generator to a low-load feedwater mode and perform a first control operation on the second loop when the first outlet working fluid status indicates that the outlet of the DC steam generator is in a non-subcooled water state; wherein, the first control operation is used to control the outlet of the DC steam generator to enter the subcooled water state. The second acquisition module is used to acquire the first state information of the loop after completing the first control operation; The second control module is used to control the second loop to switch to the long-term primary loop residual heat discharge mode when the first state information of the primary loop reaches the long-term primary loop residual heat discharge operation condition. The long-term primary loop residual heat discharge operation condition is used to indicate that the primary loop is in a physical state in which it can autonomously and safely discharge core residual heat.

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