A method and computer program product for monitoring the risk of a pressurized thermal shock in a nuclear power plant

Abstract

The application is suitable for the technical field of nuclear power plants, and provides a pressure-bearing thermal shock risk monitoring method and computer program product for a nuclear power plant. The method comprises the following steps: after an accident occurs in the nuclear power plant, the current pressure of a reactor coolant system, the current subcooling degree of a reactor outlet coolant, the operation state of a main pump, and the connection state of a residual heat removal system are obtained; if it is determined according to the operation state of the main pump that there is no main pump running in the nuclear power plant, and it is determined according to the connection state that the residual heat removal system is not connected with the reactor coolant system, a first risk monitoring curve is obtained; and whether there is a pressure-bearing thermal shock risk is determined according to the current pressure, the current subcooling degree, and the first risk monitoring curve. According to the current pressure of the reactor coolant system of the nuclear power plant and the current subcooling degree of the reactor outlet coolant of the nuclear power plant after an accident occurs in the nuclear power plant, the method can determine whether there is a pressure-bearing thermal shock risk, and the safety of the nuclear power plant is improved.

Description

Technical Field

[0001] This application belongs to the field of nuclear power plant technology, and in particular relates to a method and computer program product for monitoring the pressure thermal shock risk of nuclear power plants. Background Technology

[0002] In the safe operation of nuclear power plants, the risk of pressure thermal shock is a key hidden danger that threatens the integrity of the primary circuit pressure boundary. It may cause brittle fracture of main equipment such as pressure vessels or pressurizer fluctuating pipes, thereby leading to serious safety accidents.

[0003] Current technologies cannot determine the presence of pressurized thermal shock risk at a nuclear power plant after an accident, and primarily address this risk by optimizing equipment hardware configurations, such as adjusting pressurizer safety valve settings or improving primary loop system design to prevent the initial occurrence of pressurized thermal shock. However, such methods focus only on passive protection at the equipment level and fail to provide dynamic risk monitoring capabilities after a nuclear power plant accident. Especially under accident conditions, when a main steam pipe rupture leads to severe supercooling in the primary loop, the sudden drop in coolant temperature may exacerbate material embrittlement; or when a high-pressure safety injection system malfunctions, causing severe overpressure in the primary loop, the abnormal increase in system pressure will further amplify the risk; at the same time, equipment used for risk prevention, such as safety valves, may fail to operate or malfunction, rendering existing protection mechanisms ineffective. Due to the lack of monitoring methods for real-time operating conditions after an accident, operators cannot perform quantitative risk assessments based on key parameters such as primary loop pressure and core outlet coolant supercooling, combined with the actual reactor status (including the main pump operating status and residual heat removal system connection status), making it difficult to identify pressurized thermal shock risks in a timely manner and take targeted measures.

[0004] It can be seen that existing technology cannot determine whether a nuclear power plant is at risk of pressure thermal shock after an accident, which reduces the safety of nuclear power plants. Summary of the Invention

[0005] In view of this, embodiments of this application provide a method and computer program product for monitoring the pressure thermal shock risk of nuclear power plants, in order to solve the technical problem of low safety in nuclear power plants in the prior art.

[0006] In a first aspect, embodiments of this application provide a method for monitoring the pressure thermal shock risk of a nuclear power plant, including: After an accident at a nuclear power plant, the following information is obtained: the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system. The connection status describes whether the residual heat removal system is connected to the reactor coolant system. If, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system, then a first risk monitoring curve is obtained; the first risk monitoring curve is pre-plotted based on the brittle fracture risk of the pressure vessel of the nuclear power plant. Based on the current pressure, the current subcooling, and the first risk monitoring curve, determine whether the nuclear power plant faces a risk of pressure thermal shock.

[0007] Optional, also includes: If, based on the operating status of the main pump, it is determined that at least one main pump in the nuclear power plant is in operation, then a second risk monitoring curve is obtained; the second risk monitoring curve is pre-plotted based on the brittle fracture risk of the pressurizer surge tube of the nuclear power plant. Based on the current pressure, the current subcooling, and the second risk monitoring curve, determine whether the nuclear power plant is at risk of pressure thermal shock.

[0008] Optionally, the second risk monitoring curve is plotted in the following manner: Obtain at least two first preset cooling degrees of the coolant at the reactor core outlet; For each of the first preset cooling degrees, assuming that the water level in the regulator connected to the regulator oscillation tube fluctuates abnormally, causing all the subcooled water in the heat pipe section connected to the regulator oscillation tube to flow into the regulator oscillation tube, calculate the critical pressure that causes the regulator oscillation tube to undergo brittle fracture under the condition of the first preset cooling degree. The second risk monitoring curve is determined based on the critical pressure corresponding to each of the first preset cooling degrees; the horizontal axis of the second risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0009] Optionally, determining whether the nuclear power plant faces a risk of pressure thermal shock based on the current pressure, the current subcooling, and the second risk monitoring curve includes: Determine the first operating point corresponding to the current pressure and the current subcooling; Based on the relative position of the first operating point and the second risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

[0010] Optional, also includes: If, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is connected to the reactor coolant system, then the pressure threshold for pressure shock risk is obtained. Based on the current pressure and the pressure threshold for pressure-induced thermal shock risk, determine whether the nuclear power plant is at risk of pressure-induced thermal shock.

[0011] Optionally, obtaining the pressure threshold for thermal shock risk includes: The opening setting value of the safety valve of the waste heat discharge system is determined as the pressure threshold for the pressure-bearing thermal shock risk.

[0012] Optionally, determining whether the nuclear power plant faces a pressure shock risk based on the current pressure and the pressure threshold for pressure shock risk includes: If the current pressure is greater than or equal to the pressure threshold for pressure-bearing thermal shock risk, then the nuclear power plant is determined to be at risk of pressure-bearing thermal shock. If the current pressure is less than the pressure threshold for pressure-bearing thermal shock risk, then it is determined that the nuclear power plant does not have a pressure-bearing thermal shock risk.

[0013] Optionally, the first risk monitoring curve is plotted in the following manner: At least two preset cooling temperatures are obtained for the cold pipe section; the cold pipe section is connected to the pressure vessel; For each of the preset cooling temperatures, assuming that the temperature of the cold pipe section decreases from the hot shutdown temperature to the preset cooling degree at a preset rate, the critical pressure that causes the pressure vessel to fracture brittlely is determined. Obtain a first relationship between the temperature of the cold pipe section and the temperature at the core outlet of the nuclear power plant; Based on the critical pressure corresponding to each preset cooling temperature and the first relationship, determine the critical pressure corresponding to each core outlet temperature corresponding to each preset cooling temperature. Obtain a second relationship between the temperature at the core outlet and the subcooling of the coolant at the core outlet; Based on the critical pressure corresponding to each of the core outlet temperatures and the second relationship, the critical pressure corresponding to each of the second preset cooling degrees of the core outlet coolant corresponding to the core outlet temperature is determined. The first risk monitoring curve is determined based on the critical pressure corresponding to each of the second preset cooling degrees; the horizontal axis of the first risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0014] Optionally, determining whether the nuclear power plant faces a risk of pressure thermal shock based on the current pressure, the current subcooling, and the first risk monitoring curve includes: Determine the second operating point corresponding to the current pressure and the current subcooling; Based on the relative position of the second operating point and the first risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

[0015] Secondly, embodiments of this application provide a pressure thermal shock risk monitoring device for nuclear power plants, comprising: The first acquisition unit is used to acquire, after an accident occurs at the nuclear power plant, the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system; the connection status is used to describe whether the residual heat removal system is connected to the reactor coolant system. The second acquisition unit is used to acquire a first risk monitoring curve if, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system; the first risk monitoring curve is plotted based on the brittle fracture risk of the pressure vessel of the nuclear power plant. The determination unit is used to determine whether the nuclear power plant is at risk of pressure thermal shock based on the current pressure, the current subcooling degree, and the first risk monitoring curve.

[0016] Thirdly, embodiments of this application provide an electronic device, including a memory, a processor, and a computer program stored in the memory and executable on the processor. When the processor executes the computer program, it implements the steps of the pressure thermal shock risk monitoring method for nuclear power plants as described in any of the first aspects above.

[0017] Fourthly, embodiments of this application provide a computer-readable storage medium storing a computer program that, when executed by a processor, implements the steps of the pressure thermal shock risk monitoring method for nuclear power plants as described in any of the first aspects above.

[0018] Fifthly, embodiments of this application provide a computer program product that, when run on a control device, causes the control device to perform the steps of the pressure thermal shock risk monitoring method for nuclear power plants as described in any of the first aspects above.

[0019] The pressure thermal shock risk monitoring method and computer program product for nuclear power plants provided in this application have the following beneficial effects: In the pressure thermal shock risk monitoring method for nuclear power plants provided in this application embodiment, after an accident occurs at the nuclear power plant, the following steps are first taken: the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system. The connection status describes whether the residual heat removal system is connected to the reactor coolant system. If, based on the main pump operating status, it is determined that there are no operating main pumps at the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system, then a first risk monitoring curve is obtained. The first risk monitoring curve is plotted based on the brittle fracture risk of the pressure vessel at the nuclear power plant. Finally, based on the current pressure, the current subcooling, and the first risk monitoring curve, it is determined whether the nuclear power plant faces a pressure thermal shock risk. Since existing technologies can only determine whether a nuclear power plant is at risk of pressure thermal shock when no accident has occurred, this method can determine whether a nuclear power plant is at risk of pressure thermal shock after an accident has occurred, based on the current pressure of the reactor coolant system and the current subcooling of the core outlet coolant. Therefore, it improves the safety of nuclear power plants. Attached Figure Description

[0020] To more clearly illustrate the technical solutions in the embodiments of this application, the drawings used in the description of the embodiments or the prior art will be briefly introduced below. Obviously, the drawings described below are only some embodiments of this application. For those skilled in the art, other drawings can be obtained based on these drawings without creative effort.

[0021] Figure 1 A flowchart illustrating the implementation of a pressure thermal shock risk monitoring method for nuclear power plants, provided in an embodiment of this application; Figure 2 A schematic diagram of a first risk monitoring curve provided for an embodiment of this application; Figure 3 A graph showing the relationship between the allowable pressure and time of a voltage regulator oscillation tube is provided for an embodiment of this application; Figure 4 A schematic diagram of a second risk monitoring curve provided for an embodiment of this application; Figure 5 A schematic diagram of the structure of a pressure thermal shock risk monitoring device for a nuclear power plant provided in this application embodiment; Figure 6 This is a schematic diagram of the structure of an electronic device provided in an embodiment of this application. Detailed Implementation

[0022] It should be noted that the terminology used in the embodiments of this application is only for explaining specific embodiments of this application and is not intended to limit this application. In the description of the embodiments of this application, unless otherwise stated, "multiple" means two or more, "at least one" or "one or more" means one, two or more. The terms "first" and "second" are used for descriptive purposes only and should not be construed as indicating or implying relative importance or implicitly specifying the number of indicated technical features. Thus, a feature defined as "first" or "second" may explicitly or implicitly include one or more of that feature.

[0023] References to "one embodiment" or "some embodiments" as described in this specification mean that one or more embodiments of this application include a specific feature, structure, or characteristic described in connection with that embodiment. Therefore, the phrases "in one embodiment," "in some embodiments," "in other embodiments," "in still other embodiments," etc., appearing in different parts of this specification do not necessarily refer to the same embodiment, but rather mean "one or more, but not all, embodiments," unless otherwise specifically emphasized. The terms "comprising," "including," "having," and variations thereof mean "including but not limited to," unless otherwise specifically emphasized.

[0024] In traditional nuclear power plant operations, existing technologies primarily focus on mitigating pressure shock risks by optimizing equipment configuration. However, methods for monitoring pressure shock risks after an accident have not been fully developed. In particular, existing systems cannot provide real-time risk monitoring information for severe primary circuit undercooling accidents (such as main steam pipe rupture), severe primary circuit overpressure accidents (such as malfunctioning high-pressure safety injection systems), and situations where safety equipment fails to operate or malfunctions. This results in the inability to promptly identify the brittle fracture risk of primary circuit main equipment, thereby affecting the maintenance of the integrity of the primary circuit pressure boundary barrier.

[0025] For example, when a nuclear power plant experiences a main steam pipe rupture accident, the primary coolant temperature drops, and the subcooling at the core outlet increases. Under this condition, the main pumps are shut down, and the residual heat removal system is not connected to the reactor coolant system. Due to the lack of a targeted monitoring mechanism, the current combination of pressure and subcooling is not assessed to determine if it has entered a risk zone, resulting in no warning of the risk of brittle fracture of the pressure vessel, and operators are unable to take preventative measures. Furthermore, the integrity of the primary circuit pressure boundary barrier is potentially threatened, making it difficult to maintain a safe operating state for the system.

[0026] If the above problems are not addressed, the primary circuit main equipment may undergo brittle fracture under accident conditions due to unmonitored pressure thermal shock, causing the failure of the primary circuit pressure boundary barrier. This failure will directly jeopardize the safety barrier function of the nuclear power plant, increase the possibility of radioactive release, and may lead to system-level safety events. Furthermore, the loss of the integrity of the primary circuit pressure boundary barrier will affect the overall safety margin of the nuclear power plant, making the consequences of an accident difficult to control.

[0027] To address the above problems, this application provides a method for monitoring the pressure thermal shock risk of nuclear power plants. The implementing entity of the pressure thermal shock risk monitoring method for nuclear power plants provided in this application can be the electronic equipment of the nuclear power plant.

[0028] Please see Figure 1 , Figure 1 This is a flowchart illustrating the implementation of a pressure thermal shock risk monitoring method for nuclear power plants provided in this application embodiment. The pressure thermal shock risk monitoring method for nuclear power plants provided in this application embodiment may include steps S101 to S103, as detailed below: In S101, after an accident occurs at a nuclear power plant, the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system are obtained. The connection status describes whether the residual heat removal system is connected to the reactor coolant system.

[0029] In the embodiments of this application, the pressure-induced thermal shock (PTS) risk of nuclear power plants refers to the risk that, during the operation of a nuclear power plant, the primary loop main equipment such as the reactor pressure vessel is subjected to high pressure while simultaneously being impacted by cryogenic coolant, causing a sharp drop in the temperature of the equipment materials, which may lead to brittle fracture. This risk typically occurs under accident conditions and poses a potential threat to the safe operation of the nuclear power plant.

[0030] The reactor coolant system (RCS) is the core component of the primary loop in a nuclear power plant. Its main function is to remove the heat generated by the reactor core and transfer it to the secondary loop via a steam generator. This system typically consists of a reactor pressure vessel, main coolant pumps, a steam generator, and connecting piping, and circulates a high-temperature, high-pressure coolant within it.

[0031] The subcooling of the coolant at the reactor core outlet refers to the difference between the actual temperature of the coolant at the core outlet and its saturation temperature at the current pressure. Subcooling is an important parameter for measuring how close the coolant is to saturation and is of great significance for evaluating the heat transfer capacity of the coolant and the thermal-hydraulic state of the system.

[0032] The main pump is a key component of the reactor coolant system, responsible for driving the coolant circulation in the loop to ensure effective cooling of the reactor core. The operating status of the main pump directly affects the coolant flow rate and heat transfer efficiency.

[0033] The Residual Heat Removal System (RHR) is one of the safety systems in a nuclear power plant. Its primary function is to remove the decay heat from the reactor core after reactor shutdown, allowing the reactor coolant system to cool and depressurize to a safe state. This system typically connects to the reactor coolant system and introduces cryogenic cooling water for heat exchange.

[0034] Following a nuclear power plant accident, a series of critical operating parameters need to be obtained. These parameters include the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system. The connection status describes whether the residual heat removal system is connected to the reactor coolant system. These parameters can be acquired in real time through the nuclear power plant's instrumentation and control systems. For example, pressure sensors can measure the current pressure of the reactor coolant system; temperature and pressure sensors combined can calculate the current subcooling of the core outlet coolant; the operating status of the main pumps can be monitored through their power supply or control signals; and the connection status of the residual heat removal system can be determined through valve position sensors or system flow diagrams. Obtaining these parameters forms the basis for subsequent risk assessments.

[0035] In S102, if it is determined that there is no operating main pump in the nuclear power plant based on the main pump's operating status, and it is determined that the residual heat removal system is not connected to the reactor coolant system based on the connection status, then the first risk monitoring curve is obtained; the first risk monitoring curve is pre-drawn based on the brittle fracture risk of the nuclear power plant's pressure vessel.

[0036] In this embodiment, the first risk monitoring curve is plotted based on the brittle fracture risk characteristics of the pressure vessel in a nuclear power plant. This curve typically represents the critical pressure required for the pressure vessel to undergo brittle fracture at different coolant temperatures or subcooling degrees. By comparing the actual operating point of the nuclear power plant with this curve, the pressure-bearing thermal shock risk of the pressure vessel can be assessed.

[0037] After obtaining the operating status of the main pumps and the connection status of the waste heat removal system of the nuclear power plant, the operating mode of the nuclear power plant can be determined based on the obtained operating status of the main pumps and the connection status of the waste heat removal system.

[0038] If, based on the main pump's operating status, it is determined that there is no operating main pump at the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system, then the electronic equipment can obtain the first risk monitoring curve.

[0039] In one possible implementation, the first risk monitoring curve can be plotted through steps a to g, as detailed below: In step a, at least two preset cooling temperatures of the cold pipe section are obtained; the cold pipe section is connected to the pressure vessel.

[0040] In this implementation, the cold pipe section refers to a section of pipe in the reactor coolant system, which is usually connected to the inlet of the reactor pressure vessel and is responsible for delivering coolant to the pressure vessel.

[0041] Preset cooling temperatures refer to key temperature points artificially set to simulate the cooling process during the analysis of brittle fracture risk in pressure vessels. These preset cooling temperatures serve as input parameters for calculating brittle fracture risk and are used to simulate the stress conditions of the pressure vessel under different coolant temperatures. Representative temperature points can be selected based on nuclear power plant design specifications, operational experience, or safety analysis reports. For example, multiple discrete temperature values ​​within the range of hot shutdown temperature to cold shutdown temperature can be selected, or key temperature points that may occur under accident conditions can be chosen.

[0042] For example, the preset cooling temperature can be (60℃, 80℃, 100℃, 120℃, 150℃, 180℃, etc.).

[0043] In step b, for each preset cooling temperature, assuming that the temperature of the cold pipe section decreases from the hot shutdown temperature to the preset cooling degree at a preset rate, the critical pressure that causes brittle fracture of the pressure vessel is determined.

[0044] In this implementation, for each preset cooling temperature, assuming the temperature of the cold pipe section decreases from the hot shutdown temperature to that preset cooling level at a preset rate, the critical pressure causing brittle fracture of the pressure vessel is determined. This step aims to establish the critical conditions for brittle fracture of the pressure vessel at a specific cooling temperature and cooling rate. Through fracture mechanics analysis, combined with the pressure vessel's material properties (such as ductile-brittle transition temperature and fracture toughness curve), geometry, defect distribution, and thermal stress distribution, the minimum internal pressure causing brittle fracture of the pressure vessel at a given cooling temperature and cooling rate is calculated.

[0045] The preset rate can be greater than the preset rate threshold. That is, for each preset cooling degree, the critical pressure that causes the pressure vessel to fracture can be determined under the assumption that the temperature of the cold tube section drops from the hot shutdown temperature at a very high rate (greater than the preset rate threshold).

[0046] Please refer to Table 1, which shows the relationship between various preset cooling temperatures and critical pressures provided in the embodiments of this application.

[0047] Table 1 Preset Cooling Temperature and Critical Pressure

[0048] In step c, a first relationship is obtained between the temperature of the cold tube section and the temperature at the core outlet of the nuclear power plant.

[0049] In this implementation, since step b obtains the relationship between the preset cooling temperature and critical pressure of the cold pipe section, while the subcooling measurement of the core outlet coolant in a nuclear power plant is based on the core outlet temperature, it is necessary to obtain the first relationship between the temperature of the cold pipe section and the temperature of the core outlet of the nuclear power plant.

[0050] The following is an example of obtaining the first relation: Under natural circulation conditions with the main pump shut down, the fluid in the cold pipe section flows along the descending annulus to the reactor core inlet, driven by the density difference between the hot and cold sections. After flowing through the core, it reaches the core outlet. Due to the significant decay heat in the core, the fluid experiences a significant temperature rise as it flows through it. This temperature rise can be approximated as the temperature difference between the cold pipe section temperature and the core outlet temperature of the nuclear power plant. In practical applications, it has been found that the temperature rise of the fluid flowing through the core typically does not exceed 20°C. Therefore, a conservative initial relationship between the cold pipe section temperature and the core outlet temperature of the nuclear power plant can be defined as the cold pipe section temperature being 20°C lower than the core outlet temperature.

[0051] In step d, the critical pressure corresponding to each core outlet temperature is determined based on the critical pressure corresponding to each preset cooling temperature and the first relationship.

[0052] In this implementation, for example, based on the critical pressure corresponding to each preset cooling temperature shown in Table 1, and combined with the first relationship determined above, the critical pressure corresponding to each core outlet temperature corresponding to each preset cooling temperature can be determined. The critical pressure corresponding to each core outlet temperature can be found in Table 2.

[0053] Table 2 Core Exit Temperature and Critical Pressure

[0054] In step e, a second relationship is obtained between the core outlet temperature and the subcooling of the core outlet coolant.

[0055] In this implementation, since the critical pressure corresponding to each core outlet temperature is obtained in step d, and plotting the first risk monitoring curve requires obtaining the critical pressure corresponding to each of the second preset cooling degrees of the core outlet coolant, it is necessary to first obtain the second relationship between the core outlet temperature and the subcooling degree of the core outlet coolant through a preset method. The specific method of obtaining this relationship is not limited here.

[0056] In step f, based on the critical pressure corresponding to each core outlet temperature and the second relationship, the critical pressure corresponding to each second preset cooling degree of the core outlet coolant corresponding to the core outlet temperature is determined.

[0057] In this implementation, for example, based on the critical pressures corresponding to each core outlet temperature shown in Table 2, and combined with the second relationship obtained in step e, the critical pressures corresponding to each second preset cooling degree of the core outlet coolant corresponding to the core outlet temperature can be determined. The critical pressures corresponding to each second preset cooling degree can be found in Table 3.

[0058] Table 3 Second Preset Cooling Degree and Critical Pressure

[0059] In step g, a first risk monitoring curve is determined based on the critical pressure corresponding to each second preset cooling degree; the horizontal axis of the first risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0060] In this implementation, after determining the critical pressure corresponding to each of the second preset cooling degrees, the first risk monitoring curve can be determined based on the critical pressure corresponding to each of the second preset cooling degrees.

[0061] Please see Figure 2 , Figure 2 This is a schematic diagram of a first risk monitoring curve provided for an embodiment of this application. Figure 2 In this context, ΔTsat represents the degree of cooling, and PRCP represents the pressure.

[0062] In S103, based on the current pressure, current subcooling, and the first risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

[0063] In this embodiment, after obtaining the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, and the first risk monitoring curve, a second operating point corresponding to the current pressure and current subcooling can be determined. Then, based on the relative position of the second operating point and the first risk monitoring curve, it is determined whether the nuclear power plant faces a risk of pressure thermal shock.

[0064] Specifically, "determining whether a nuclear power plant is at risk of pressure thermal shock based on the relative position of the second operating point and the first risk monitoring curve" can be defined as follows: if the second operating point is above the first risk monitoring curve, then the nuclear power plant is at risk of pressure thermal shock; if the second operating point is below the first risk monitoring curve, then the nuclear power plant is at risk of pressure thermal shock.

[0065] As can be seen from the above, in the nuclear power plant pressure thermal shock risk monitoring method provided in the embodiments of this application, after an accident occurs at the nuclear power plant, the current pressure of the reactor coolant system, the current subcooling of the reactor core outlet coolant, the operating status of the main pump, and the connection status of the residual heat removal system are first obtained. The connection status describes whether the residual heat removal system is connected to the reactor coolant system. If, based on the main pump operating status, it is determined that there is no operating main pump at the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system, then a first risk monitoring curve is obtained. The first risk monitoring curve is plotted based on the brittle fracture risk of the pressure vessel of the nuclear power plant. Finally, based on the current pressure, the current subcooling, and the first risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock. Since existing technologies can only determine whether a nuclear power plant is at risk of pressure thermal shock when no accident has occurred, this method can determine whether a nuclear power plant is at risk of pressure thermal shock after an accident has occurred, based on the current pressure of the reactor coolant system and the current subcooling of the core outlet coolant. Therefore, it improves the safety of nuclear power plants.

[0066] The embodiments described above provide how to determine whether a nuclear power plant is at risk of pressure thermal shock when there is no operating main pump and the waste heat removal system is not connected to the reactor coolant system.

[0067] The following provides guidance on how to determine the risk of pressure thermal shock at a nuclear power plant in other circumstances.

[0068] In one possible implementation, if it is determined that there is at least one operating main pump in the nuclear power plant based on the main pump's operating status, then a second risk monitoring curve is obtained; the second risk monitoring curve is pre-plotted based on the brittle fracture risk of the pressurizer surge tube in the nuclear power plant. Based on the current pressure, current subcooling, and the second risk monitoring curve, determine whether the nuclear power plant faces the risk of pressure thermal shock.

[0069] In this implementation, the second risk monitoring curve can be plotted through steps h to j, as detailed below: In step h, at least two first preset cooling degrees of the core outlet coolant are obtained.

[0070] For example, at least two first preset cooling temperatures of the core outlet coolant may include: 180°C, 210°C, 240°C, 270°C, and 300°C.

[0071] In step i, for each first preset cooling degree, assuming that the water level in the regulator connected to the regulator oscillation tube fluctuates abnormally, causing all the subcooled water in the heat pipe section connected to the regulator oscillation tube to flow into the regulator oscillation tube, the critical pressure that causes the regulator oscillation tube to undergo brittle fracture under the condition of the first preset cooling degree is calculated.

[0072] In this implementation, for each first preset cooling degree, assuming that the water level in the regulator connected to the regulator oscillation tube fluctuates abnormally, causing all the subcooled water in the heat pipe section connected to the regulator oscillation tube to flow into the regulator oscillation tube, the critical pressure that causes the regulator oscillation tube to undergo brittle fracture under the first preset cooling degree is calculated, which is the core of constructing the second risk monitoring curve.

[0073] The scenario described above, "assuming abnormal fluctuations in the water level within the voltage regulator connected to the voltage regulator's oscillator, causing all the subcooled water in the heat pipe section connected to the voltage regulator's oscillator to flow into the voltage regulator's oscillator," depicts a specific accident scenario that could lead to significant thermal shock to the voltage regulator's oscillator. In this scenario, the voltage regulator's oscillator would experience a rapid temperature drop, resulting in significant thermal stress.

[0074] The phrase "calculating the critical pressure that causes the voltage regulator's oscillating tube to undergo brittle fracture under the first preset cooling temperature" refers to determining the critical pressure required to cause the voltage regulator's oscillating tube to undergo brittle fracture through fracture mechanics analysis under the aforementioned assumed operating conditions and specific cooling temperature.

[0075] This calculation typically involves the following aspects: First, based on the geometry of the voltage regulator's oscillator tube, material properties (such as fracture toughness, yield strength, coefficient of thermal expansion, etc.), and a preset cooling degree, a temperature and stress field model of the oscillator tube is established. Second, considering potential defects (such as cracks) within the oscillator tube, the stress intensity factor or J-integral at the defect tip is calculated under different pressures using linear elastic fracture mechanics or elastoplastic fracture mechanics theories. Finally, the calculated stress intensity factor or J-integral is compared with the material's fracture toughness to determine the critical pressure leading to brittle fracture at a given cooling degree.

[0076] In practical applications, because the temperature gradient on the waveguide changes over time, the allowable pressure also changes accordingly. Here, the critical pressure is taken as the lowest allowable pressure during the entire transient period. For example... Figure 3 As shown, Figure 3 This application provides a diagram illustrating the relationship between the allowable pressure and time of a voltage regulator's oscillation tube. This can be used... Figure 3 The minimum allowable pressure shown is determined to be the corresponding critical pressure.

[0077] In step j, a second risk monitoring curve is determined based on the critical pressure corresponding to each first preset cooling degree; the horizontal axis of the second risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0078] In this implementation, after obtaining at least two first preset cooling degrees and their corresponding critical pressures, a second risk monitoring curve can be plotted by data fitting, interpolation, or connecting these data points.

[0079] Please see Figure 4 , Figure 4 This is a schematic diagram of a second risk monitoring curve provided in an embodiment of this application. Figure 4 In this context, ΔTsat represents the degree of cooling, and PRCP represents the pressure.

[0080] After obtaining the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, and the second risk monitoring curve, the first operating point corresponding to the current pressure and current subcooling can be determined. Then, based on the relative position of the first operating point and the second risk monitoring curve, it can be determined whether the nuclear power plant faces the risk of pressure thermal shock.

[0081] Specifically, "determining whether a nuclear power plant is at risk of pressure thermal shock based on the relative position of the first operating point and the second risk monitoring curve" can be defined as follows: if the first operating point is above the second risk monitoring curve, then the nuclear power plant is at risk of pressure thermal shock; if the first operating point is below the second risk monitoring curve, then the nuclear power plant is at risk of pressure thermal shock.

[0082] In another possible implementation, if it is determined that there is no operating main pump in the nuclear power plant based on the main pump's operating status, and it is determined that the residual heat removal system is connected to the reactor coolant system based on the connection status, then the pressure threshold for the pressure shock risk is obtained. Based on the current pressure and the pressure threshold for pressure-induced thermal shock risk, determine whether the nuclear power plant faces a risk of pressure-induced thermal shock.

[0083] In this implementation, with the residual heat removal system connected, the core coolant temperature is relatively low. Under these conditions, the main threat to the pressure vessel comes from the compressive stress generated by the pressure difference between the inside and outside of the pressure vessel. If the pressure difference is too high, the significant compressive stress poses a risk of pressure thermal shock. Since the overpressure protection of the primary loop is ensured by the safety valves on the residual heat removal system pipeline during reactor operation with the residual heat removal system connected, the opening setting value of the safety valves in the residual heat removal system can be determined as the pressure threshold for the risk of pressure thermal shock.

[0084] After obtaining the current pressure of the reactor coolant system and the pressure threshold for pressure shock risk, the following methods can be used to determine whether a nuclear power plant is at risk of pressure shock: If the current pressure is greater than or equal to the pressure threshold for pressure-induced thermal shock risk, then the nuclear power plant is determined to be at risk of pressure-induced thermal shock; if the current pressure is less than the pressure threshold for pressure-induced thermal shock risk, then the nuclear power plant is determined not to be at risk of pressure-induced thermal shock.

[0085] When the current pressure of the reactor coolant system in a nuclear power plant reaches or exceeds the preset pressure threshold for pressure shock risk, it indicates that the system is in or has entered a high-risk state, requiring corresponding risk response measures. This judgment condition can be implemented through a comparator circuit or software logic module, monitoring the current pressure value in real time and comparing it with the stored risk pressure threshold. Alternatively, an alarm system can be set up to trigger a risk alarm and automatically record the risk status when the current pressure reaches or exceeds the threshold. The technical feature that "if the current pressure is less than the pressure threshold for pressure shock risk, then the nuclear power plant is determined to have no pressure shock risk" defines the condition for determining that the nuclear power plant has no pressure shock risk. When the current pressure of the reactor coolant system in a nuclear power plant is lower than the preset pressure threshold for pressure shock risk, it indicates that the system is in a relatively safe state with a low pressure shock risk. This judgment condition can also be implemented through a comparator circuit or software logic module; when the current pressure is lower than the threshold, the system maintains normal operation or does not trigger a risk alarm. Alternatively, a status indicator can display a safe status, or a no-risk message can be displayed on the control interface.

[0086] Through the above technical solutions, under different operating conditions after a nuclear power plant accident, the current pressure of the reactor coolant system and the current subcooling of the reactor core outlet coolant can be used to determine whether the nuclear power plant is at risk of pressure thermal shock, thus improving the safety of the nuclear power plant.

[0087] Based on the pressure thermal shock risk monitoring method for nuclear power plants provided in the above embodiments, this application further provides a pressure thermal shock risk monitoring device for nuclear power plants that implements the above method embodiments. Please refer to... Figure 5 , Figure 5 This is a schematic diagram of a pressure thermal shock risk monitoring device for a nuclear power plant, provided as an embodiment of this application. Figure 5 As shown, the pressure thermal shock risk monitoring device 50 for nuclear power plants may include: a first acquisition unit 51, a second acquisition unit 52, and a determination unit 53. Wherein: The first acquisition unit 51 is used to acquire, after an accident occurs at the nuclear power plant, the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system. The connection status is used to describe whether the residual heat removal system is connected to the reactor coolant system.

[0088] The second acquisition unit 52 is used to acquire a first risk monitoring curve if it is determined, based on the main pump operating status, that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system; the first risk monitoring curve is pre-drawn based on the brittle fracture risk of the pressure vessel of the nuclear power plant.

[0089] The determination unit 53 is used to determine whether the nuclear power plant is at risk of pressure thermal shock based on the current pressure, current subcooling and the first risk monitoring curve.

[0090] Optionally, the determining unit 53 is also used for: If, based on the operating status of the main pump, it is determined that at least one main pump in the nuclear power plant is in operation, then a second risk monitoring curve is obtained; the second risk monitoring curve is pre-plotted based on the brittle fracture risk of the pressurizer surge tube in the nuclear power plant. Based on the current pressure, current subcooling, and the second risk monitoring curve, determine whether the nuclear power plant faces the risk of pressure thermal shock.

[0091] Optionally, the determining unit 53 is specifically used for: Obtain at least two first preset cooling degrees for the core outlet coolant; For each first preset cooling degree, assuming that the water level in the regulator connected to the regulator oscillation tube fluctuates abnormally, causing all the subcooled water in the heat pipe section connected to the regulator oscillation tube to flow into the regulator oscillation tube, calculate the critical pressure that causes the regulator oscillation tube to undergo brittle fracture under the condition of the first preset cooling degree. A second risk monitoring curve is determined based on the critical pressure corresponding to each first preset cooling degree; the horizontal axis of the second risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0092] Optionally, the determining unit 53 is specifically used for: Determine the first operating point corresponding to the current pressure and current subcooling; Based on the relative positions of the first operating point and the second risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

[0093] Optionally, the determining unit 53 is also used for: If, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is connected to the reactor coolant system, then the pressure threshold for the risk of pressure thermal shock is obtained. Based on the current pressure and the pressure threshold for pressure-induced thermal shock risk, determine whether the nuclear power plant faces a risk of pressure-induced thermal shock.

[0094] Optionally, the determining unit 53 is specifically used for: The opening setting value of the safety valve of the waste heat discharge system is determined as the pressure threshold for the risk of thermal shock under pressure.

[0095] Optionally, the determining unit 53 is specifically used for: If the current pressure is greater than or equal to the pressure threshold for pressure-bearing thermal shock risk, then the nuclear power plant is determined to be at risk of pressure-bearing thermal shock. If the current pressure is less than the pressure threshold for pressure-induced thermal shock risk, then the nuclear power plant is determined to be free from pressure-induced thermal shock risk.

[0096] Optionally, the second acquisition unit 52 is specifically used for: Obtain at least two preset cooling temperatures for the cold pipe section; the cold pipe section is connected to the pressure vessel; For each preset cooling temperature, under the assumption that the temperature of the cold pipe section drops from the hot shutdown temperature to the preset cooling degree at a preset rate, the critical pressure that causes brittle fracture of the pressure vessel is determined. To obtain the primary relationship between the temperature of the cold tube section and the temperature at the core outlet of the nuclear power plant; Based on the critical pressure corresponding to each preset cooling temperature and the first relationship, determine the critical pressure corresponding to each core outlet temperature corresponding to each preset cooling temperature. To obtain a second relationship between the core exit temperature and the subcooling of the core exit coolant; Based on the critical pressure corresponding to each core outlet temperature and the second relationship, determine the critical pressure corresponding to each second preset cooling degree of the core outlet coolant corresponding to the core outlet temperature. A first risk monitoring curve is determined based on the critical pressure corresponding to each second preset cooling degree; the horizontal axis of the first risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

[0097] Optionally, the determining unit 53 is specifically used for: Determine the second operating point corresponding to the current pressure and current subcooling; Based on the relative position of the second operating point and the first risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

[0098] It should be noted that the information interaction and execution process between the above-mentioned units are based on the same concept as the method embodiments of this application. Their specific functions and technical effects can be referred to the method embodiments section, and will not be repeated here.

[0099] Please see Figure 6 , Figure 6 This is a schematic diagram of the structure of an electronic device provided in an embodiment of this application. Figure 6 As shown, the electronic device 6 provided in this embodiment may include: a processor 60, a memory 61, and a computer program 62 stored in the memory 61 and executable on the processor 60, such as a program corresponding to a pressure thermal shock risk monitoring method for nuclear power plants. When the processor 60 executes the computer program 62, it implements the steps described in the embodiment of the pressure thermal shock risk monitoring method for nuclear power plants, for example... Figure 1 S101~S103 are shown. Alternatively, when processor 60 executes computer program 62, it implements the functions of each module / unit in the above-described embodiment of the pressure thermal shock risk monitoring device for nuclear power plants, for example... Figure 5 The functions of units 51-53 shown.

[0100] For example, computer program 62 can be divided into one or more modules / units, one or more of which are stored in memory 61 and executed by processor 60 to complete this application. One or more modules / units can be a series of computer program instruction segments capable of performing specific functions, which describe the execution process of computer program 62 in electronic device 6. For example, computer program 62 can be divided into a first acquisition unit 51, a second acquisition unit 52, and a determination unit 53. For the specific functions of each unit, please refer to... Figure 5 The relevant descriptions in the corresponding embodiments are not repeated here.

[0101] Those skilled in the art will understand that Figure 6 This is merely an example of electronic device 6 and does not constitute a limitation on electronic device 6. It may include more or fewer components than shown, or combine certain components, or use different components.

[0102] The processor 60 can be a central processing unit (CPU), or other general-purpose processors, digital signal processors (DSPs), application-specific integrated circuits (ASICs), field-programmable gate arrays (FPGAs), or other programmable logic devices, discrete gate or transistor logic devices, discrete hardware components, etc. The general-purpose processor can be a microprocessor or any conventional processor.

[0103] The memory 61 can be an internal storage unit of the electronic device 6, such as a hard disk or RAM. The memory 61 can also be an external storage device of the electronic device 6, such as a plug-in hard disk, smart media card (SMC), secure digital (SD) card, or flash card. Furthermore, the memory 61 can include both internal and external storage units of the electronic device 6. The memory 61 is used to store computer programs and other programs and data required by the electronic device. The memory 61 can also be used to temporarily store data that has been output or will be output.

[0104] Those skilled in the art will clearly understand that, for the sake of convenience and brevity, the above-described division of functional units is merely an example. In practical applications, the above functions can be assigned to different functional units as needed. For instance, the internal structure of a nuclear power plant's pressure thermal shock risk monitoring device can be divided into different functional units to complete all or part of the functions described above. The functional units in the embodiments can be integrated into one processing unit, or each unit can exist physically separately, or two or more units can be integrated into one unit. The integrated unit can be implemented in hardware or as a software functional unit. Furthermore, the specific names of the functional units are only for easy differentiation and are not intended to limit the scope of protection of this application. The specific working process of the units in the above system can be referred to the corresponding process in the foregoing method embodiments, and will not be repeated here.

[0105] This application also provides a computer-readable storage medium storing a computer program that, when executed by a processor, can implement the steps in the various method embodiments described above.

[0106] This application provides a computer program product that, when run on a terminal device, enables the terminal device to implement the steps described in the various method embodiments above.

[0107] In the above embodiments, the descriptions of each embodiment have different focuses. For parts that are not described in detail or recorded in a certain embodiment, refer to the relevant descriptions of other embodiments.

[0108] Those skilled in the art will recognize that the units and algorithm steps of the various examples described in conjunction with the embodiments disclosed herein can be implemented in electronic hardware, or a combination of computer software and electronic hardware. Whether these functions are implemented in hardware or software depends on the specific application and design constraints of the technical solution. Those skilled in the art can use different methods to implement the described functions for each specific application, but such implementation should not be considered beyond the scope of this application.

[0109] The above-described embodiments are only used to illustrate the technical solutions of this application, and are not intended to limit them. Although this application has been described in detail with reference to the foregoing embodiments, those skilled in the art should understand that modifications can still be made to the technical solutions described in the foregoing embodiments, or equivalent substitutions can be made to some of the technical features. Such modifications or substitutions do not cause the essence of the corresponding technical solutions to deviate from the spirit and scope of the technical solutions of the embodiments of this application, and should all be included within the protection scope of this application.

Claims

1. A method for monitoring the pressure thermal shock risk of a nuclear power plant, characterized in that, include: After an accident at a nuclear power plant, the following information is obtained: the current pressure of the reactor coolant system, the current subcooling of the core outlet coolant, the operating status of the main pumps, and the connection status of the residual heat removal system. The connection status describes whether the residual heat removal system is connected to the reactor coolant system. If, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is not connected to the reactor coolant system, then a first risk monitoring curve is obtained. The first risk monitoring curve was pre-plotted based on the brittle fracture risk of the pressure vessel of the nuclear power plant; Based on the current pressure, the current subcooling, and the first risk monitoring curve, determine whether the nuclear power plant faces a risk of pressure thermal shock.

2. The method according to claim 1, characterized in that, Also includes: If, based on the operating status of the main pump, it is determined that at least one main pump in the nuclear power plant is in operation, then a second risk monitoring curve is obtained. The second risk monitoring curve was pre-plotted based on the risk of brittle fracture of the pressurizer surge tube in the nuclear power plant; Based on the current pressure, the current subcooling, and the second risk monitoring curve, determine whether the nuclear power plant is at risk of pressure thermal shock.

3. The method according to claim 2, characterized in that, The second risk monitoring curve is plotted in the following way: Obtain at least two first preset cooling degrees of the coolant at the reactor core outlet; For each of the first preset cooling degrees, assuming that the water level in the regulator connected to the regulator oscillation tube fluctuates abnormally, causing all the subcooled water in the heat pipe section connected to the regulator oscillation tube to flow into the regulator oscillation tube, calculate the critical pressure that causes the regulator oscillation tube to undergo brittle fracture under the condition of the first preset cooling degree. The second risk monitoring curve is determined based on the critical pressure corresponding to each of the first preset cooling degrees; the horizontal axis of the second risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

4. The method according to claim 2, characterized in that, The step of determining whether the nuclear power plant faces a risk of pressure thermal shock based on the current pressure, the current subcooling, and the second risk monitoring curve includes: Determine the first operating point corresponding to the current pressure and the current subcooling; Based on the relative position of the first operating point and the second risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

5. The method according to claim 1, characterized in that, Also includes: If, based on the operating status of the main pump, it is determined that there is no operating main pump in the nuclear power plant, and based on the connection status, it is determined that the residual heat removal system is connected to the reactor coolant system, then the pressure threshold for pressure shock risk is obtained. Based on the current pressure and the pressure threshold for pressure-induced thermal shock risk, determine whether the nuclear power plant is at risk of pressure-induced thermal shock.

6. The method according to claim 5, characterized in that, The process of obtaining the pressure threshold for thermal shock risk includes: The opening setting value of the safety valve of the waste heat discharge system is determined as the pressure threshold for the pressure-bearing thermal shock risk.

7. The method according to claim 5, characterized in that, The step of determining whether the nuclear power plant faces a pressure shock risk based on the current pressure and the pressure threshold for pressure shock risk includes: If the current pressure is greater than or equal to the pressure threshold for pressure-bearing thermal shock risk, then the nuclear power plant is determined to be at risk of pressure-bearing thermal shock. If the current pressure is less than the pressure threshold for pressure-bearing thermal shock risk, then it is determined that the nuclear power plant does not have a pressure-bearing thermal shock risk.

8. The method according to claim 1, characterized in that, The first risk monitoring curve is plotted in the following way: At least two preset cooling temperatures are obtained for the cold pipe section; the cold pipe section is connected to the pressure vessel; For each of the preset cooling temperatures, assuming that the temperature of the cold pipe section decreases from the hot shutdown temperature to the preset cooling degree at a preset rate, the critical pressure that causes the pressure vessel to fracture brittlely is determined. Obtain a first relationship between the temperature of the cold pipe section and the temperature at the core outlet of the nuclear power plant; Based on the critical pressure corresponding to each preset cooling temperature and the first relationship, determine the critical pressure corresponding to each core outlet temperature corresponding to each preset cooling temperature. Obtain a second relationship between the temperature at the core outlet and the subcooling of the coolant at the core outlet; Based on the critical pressure corresponding to each of the core outlet temperatures and the second relationship, the critical pressure corresponding to each of the second preset cooling degrees of the core outlet coolant corresponding to the core outlet temperature is determined. The first risk monitoring curve is determined based on the critical pressure corresponding to each of the second preset cooling degrees; the horizontal axis of the first risk monitoring curve is the cooling degree, and the vertical axis is the pressure.

9. The method according to any one of claims 1 to 8, characterized in that, The step of determining whether the nuclear power plant faces a risk of pressure thermal shock based on the current pressure, the current subcooling, and the first risk monitoring curve includes: Determine the second operating point corresponding to the current pressure and the current subcooling; Based on the relative position of the second operating point and the first risk monitoring curve, it is determined whether the nuclear power plant is at risk of pressure thermal shock.

10. A computer program product, characterized in that, When the computer program product is executed by a processor, it implements the steps of the pressure thermal shock risk monitoring method for nuclear power plants as described in any one of claims 1 to 7.

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