Accident risk evaluation method of nuclear power plant, related device and medium

Abstract

The invention provides an accident risk evaluation method for a nuclear power plant, a related device and a medium, and the method comprises the steps: merging all reactor core damage state sequences of the nuclear power plant through employing a connection event tree, and determining a target reactor core damage state sequence; taking the target reactor core damage state sequence as a starting point of a target containment event tree corresponding to the to-be-analyzed accident; determining the position of each node event in the target containment event tree based on the occurrence sequence of the node events in preset risk level accidents corresponding to the to-be-analyzed accident, thereby determining multiple paths of the target containment event tree; analyzing the plurality of node events to quantify the accident occurrence frequency corresponding to each path; and determining an evaluation result of the nuclear power plant under the to-be-analyzed accident based on the accident occurrence frequency corresponding to the plurality of paths in the target containment event tree. The embodiment of the invention can improve the evaluation effect of the accident risk, identify the weak link of the nuclear power plant design, reduce the release risk, and improve the safety of the nuclear power plant.

Description

Technical Field

[0001] The present application relates to the field of nuclear power technology, in particular to accident risk assessment methods, related devices and media for nuclear power plants. Background Art

[0002] A nuclear power plant is a power plant that converts nuclear energy into thermal energy to generate steam for steam turbines, which in turn drive generators to generate electricity. A serious accident in a nuclear power plant may cause a large release of radioactive materials in the reactor and failure of the containment. Therefore, an accident risk assessment is needed to mitigate the occurrence of serious accidents such as failure of the containment and large-scale radioactive release in nuclear power plants.

[0003] In the process of accident risk assessment, relevant technologies use a containment event tree for analysis. However, the containment event tree is relatively simple, and most of the data in it is general data. It cannot accurately and meticulously reflect the development process of serious accidents, and cannot truly reflect the serious accident risk situation of nuclear power plants. The evaluation effect is poor, which may cause more serious consequences. Summary of the invention

[0004] The embodiments of the present application provide a method, related devices and media for evaluating accident risks in nuclear power plants, which can improve the evaluation effect of accident risks in nuclear power plants, identify weak links in the design of nuclear power plants, reduce the risk of large-scale releases, and improve the safety of nuclear power plants.

[0005] In a first aspect, an embodiment of the present application provides a method for analyzing accident risks of a nuclear power plant, including:

[0006] The connection event tree is used to merge the core damage state sequences of nuclear power plants and determine the target core damage state sequence.

[0007] Taking the target core damage state sequence as the starting point of the target containment event tree corresponding to the accident to be analyzed;

[0008] Based on the occurrence order of node events in the accidents with preset risk levels corresponding to the accident to be analyzed, determining the position of each node event in the target containment event tree, thereby determining multiple paths of the target containment event tree;

[0009] Analyzing a plurality of the node events to quantify the frequency of accidents corresponding to each of the paths;

[0010] Based on the accident occurrence frequencies corresponding to the plurality of paths in the target containment event tree, an evaluation result of the nuclear power plant under the accident to be analyzed is determined.

[0011] According to the first aspect of the embodiment of the present application, the accident risk assessment method of the nuclear power plant provided has at least the following beneficial effects: for the risk of a large amount of radioactive release caused by the accident to be analyzed, firstly, the core damage state sequences of the nuclear power plant are merged by using the connection event tree, and the target core damage state sequence corresponding to the accident to be analyzed is determined, and then the starting point of the target containment event tree is determined, so that the starting point of the target containment event tree is adapted to the nuclear power plant. Afterwards, based on the occurrence order of the node events in the preset risk level accident, the position of each node event in the target containment event tree is determined, and then multiple paths of the target containment event tree are determined. Since the node event is an event that occurs in the nuclear power plant during the preset risk level accident corresponding to the accident to be analyzed, the node event of the target containment event tree can reflect the design characteristics of the nuclear power plant reactor type, and each path can reflect the development process of the preset risk level accident, so that the accident frequency obtained is closer to the actual frequency of the nuclear power plant accident, so the evaluation result obtained is more accurate, and the evaluation effect of the accident risk of the nuclear power plant is improved, so that the weak links in the design of the nuclear power plant can be identified, the risk of a large amount of release can be reduced, and the safety of the nuclear power plant can be improved.

[0012] According to some embodiments of the first aspect of the present application, before merging the core damage state sequences of the nuclear power plant using the connection event tree to determine the target core damage state sequence, the method further includes:

[0013] Conducting a first-level probabilistic safety assessment on a nuclear power plant to obtain multiple accident sequences of core damage in the nuclear power plant;

[0014] Determine a list of characteristic quantities according to the reactor type in the nuclear power plant;

[0015] For each of the accident sequences, based on the feature quantity corresponding to each accident in the accident sequence, determining the feature attribute corresponding to the accident in the feature quantity list;

[0016] Based on a plurality of the characteristic attributes corresponding to each of the accident sequences, the accident sequences are merged into a core damage state sequence.

[0017] According to some embodiments of the first aspect of the present application, the method of merging the core damage state sequences of the nuclear power plant using the connection event tree to determine the target core damage state sequence includes:

[0018] For each of the core damage state sequences, a connection event tree is used to determine a plurality of characteristic attributes corresponding to the core damage state sequence;

[0019] If the plurality of characteristic attributes include a target characteristic attribute corresponding to the accident to be analyzed, the core damage state sequence is determined as a target core damage state sequence.

[0020] According to some embodiments of the first aspect of the present application, if the plurality of characteristic attributes include a target characteristic attribute corresponding to the accident to be analyzed, after determining the core damage state sequence as a target core damage state sequence, the method further includes:

[0021] For the target core damage state sequence, determining a first characteristic attribute having the highest correlation with the target core damage state sequence from among the plurality of characteristic attributes;

[0022] Based on the first characteristic attribute, determining a functional header of the target core damage state sequence in the connection event tree;

[0023] The target core damage state sequence is simulated through the functional header, and when the simulation result does not meet the preset condition, the first characteristic attribute is used as the boundary condition of the target containment event tree.

[0024] According to some embodiments of the first aspect of the present application, the analyzing the plurality of node events to quantify the frequency of accidents corresponding to each of the paths includes:

[0025] For each of the paths in the target containment event tree, determining a plurality of node events;

[0026] Analyze each of the node events using determinism and probability theory to obtain a first probability density function and a second probability density function corresponding to the node event, wherein the first probability density function is used to represent a mapping relationship between a containment load and a first probability density, and the second probability density function is used to represent a mapping relationship between a containment load capacity and a second probability density;

[0027] Calculating, based on the first probability density function and the second probability density function, a failure probability that the containment load is greater than the containment bearing capacity;

[0028] Based on the failure probabilities of the plurality of node events and the occurrence frequency of the target core damage state sequence, the accident occurrence frequency corresponding to the path is determined.

[0029] According to some embodiments of the first aspect of the present application, the use of determinism and probability theory to analyze each of the node events to obtain a first probability density function and a second probability density function corresponding to the node event includes:

[0030] Performing a qualitative analysis on each of the node events to obtain a plurality of influencing factors corresponding to the node event, and determining a third probability density function of each of the influencing factors;

[0031] Monte Carlo simulation calculation is performed on the plurality of third probability density functions to obtain a first probability density function and a second probability density function.

[0032] According to some embodiments of the first aspect of the present application, based on the accident occurrence frequencies corresponding to the plurality of paths in the target containment event tree, determining the evaluation result of the nuclear power plant under the accident to be analyzed includes:

[0033] Determining a target path among the multiple paths based on the accident occurrence frequencies corresponding to the multiple paths in the target containment event tree;

[0034] Determining a plurality of mitigation node events from a plurality of node events corresponding to the target path;

[0035] Obtaining a mitigation time window corresponding to each of the mitigation node events;

[0036] The plurality of mitigation node events and the mitigation time windows corresponding to the mitigation node events are used as evaluation results of the nuclear power plant under the accident to be analyzed, so as to execute the mitigation node events within the mitigation event window.

[0037] In a second aspect, an embodiment of the present application provides an accident risk assessment device for a nuclear power plant, comprising:

[0038] A sequence merging unit, used to merge various core damage state sequences of the nuclear power plant using a connection event tree to determine a target core damage state sequence;

[0039] A starting point determination unit, configured to use the target core damage state sequence as the starting point of a target containment event tree corresponding to the accident to be analyzed;

[0040] A path determination unit, configured to determine the position of each of the node events in the target containment event tree based on the occurrence order of the node events in the accidents of preset risk levels corresponding to the accident to be analyzed, thereby determining multiple paths of the target containment event tree;

[0041] An analysis and quantification unit, used for analyzing the plurality of node events to quantify the frequency of accidents corresponding to each of the paths;

[0042] The risk assessment unit is used to determine the assessment result of the nuclear power plant under the accident to be analyzed based on the accident occurrence frequency corresponding to the multiple paths in the target containment event tree.

[0043] In a third aspect, an embodiment of the present application provides an electronic device, including a memory and a processor, wherein the memory stores a computer program, and when the processor executes the computer program, the accident risk assessment method for a nuclear power plant as described above is implemented.

[0044] In a fourth aspect, a computer-readable storage medium is provided, wherein the storage medium stores a computer program, and when the computer program is executed by a processor, the accident risk assessment method for a nuclear power plant as described above is implemented.

[0045] Other features and advantages of the present application will be described in the following description, and partly become apparent from the description, or understood by practicing the present application. The purpose and other advantages of the present application can be realized and obtained by the structures specifically pointed out in the description, claims and drawings. BRIEF DESCRIPTION OF THE DRAWINGS

[0046] The accompanying drawings are used to provide further understanding of the technical solution of the present application and constitute a part of the specification. Together with the embodiments of the present application, they are used to explain the technical solution of the present application and do not constitute a limitation on the technical solution of the present application.

[0047] Figure 1 is a system architecture diagram of a method for assessing accident risks in a nuclear power plant according to an embodiment of the present application;

[0048] Figure 2 is an overall flow chart of an accident risk assessment method for a nuclear power plant according to an embodiment of the present application;

[0049] Figure 3 It is a schematic diagram of a high-pressure reactor containment event tree provided in an embodiment of the present application;

[0050] Figure 4 is a flow chart of a sequence for obtaining core damage status provided by an embodiment of the present application;

[0051] Figure 5 is a schematic diagram of a core damage state sequence provided in an embodiment of the present application;

[0052] Figure 6 yes Figure 2 A flow chart of determining a target core damage state sequence in step 210;

[0053] Figure 7 It is a flow chart for setting boundary conditions for a target containment event tree provided by an embodiment of the present application;

[0054] Figure 8 yes Figure 2 A flow chart of determining the accident frequency corresponding to the path in step 240;

[0055] Fig. 9 is a schematic diagram of a first probability density function and a second probability density function provided in an embodiment of the present application;

[0056] Fig.10 is a schematic diagram of a target containment event tree provided in an embodiment of the present application;

[0057] Fig.11 yes Figure 8 Step 820 obtains a flow chart of the first probability density function and the second probability density function;

[0058] Fig.12 yes Figure 2 A schematic diagram of determining the evaluation result in step 250;

[0059] Fig.13 is a module diagram of an accident risk assessment device for a nuclear power plant according to an embodiment of the present application;

[0060] Fig.14 It is a terminal structure diagram for executing the accident risk assessment method of a nuclear power plant according to an embodiment of the present application;

[0061] Fig.15 It is a server structure diagram for executing the accident risk assessment method of a nuclear power plant according to an embodiment of the present application. DETAILED DESCRIPTION

[0062] In order to make the purpose, technical solution and advantages of the present application more clearly understood, the present application is further described in detail below in conjunction with the accompanying drawings and embodiments. It should be understood that the specific embodiments described herein are only used to explain the present application and are not used to limit the present application.

[0063] Before further describing the embodiments of the present application in detail, the nouns and terms involved in the embodiments of the present application are explained. The nouns and terms involved in the embodiments of the present application are subject to the following interpretations:

[0064] A nuclear power plant is a power plant that converts nuclear energy into thermal energy to generate steam for steam turbines, which in turn drive generators to generate electricity. A serious accident in a nuclear power plant may cause a large release of radioactive materials in the reactor and failure of the containment. Therefore, an accident risk assessment is needed to mitigate the occurrence of serious accidents such as failure of the containment and large-scale radioactive release in nuclear power plants.

[0065] In the process of severe accident risk assessment, relevant technologies use a containment event tree for analysis. However, the containment event tree is relatively simple, and most of the data in it is general data. It cannot accurately and meticulously reflect the development process of severe accidents, and cannot truly reflect the severe accident risk situation of nuclear power plants. The evaluation effect is poor, which may cause more serious consequences.

[0066] Based on this, the embodiment of the present application provides a method, related device and medium for assessing the accident risk of a nuclear power plant. The method for assessing the accident risk of a nuclear power plant provided by the embodiment of the present application can improve the assessment effect of the accident risk of a nuclear power plant, identify the weak links in the design of a nuclear power plant, reduce the risk of a large number of releases, and improve the safety of the nuclear power plant process.

[0067] System architecture of the present application embodiment

[0068] Figure 1 This is a system architecture diagram of a nuclear power plant severe accident analysis method according to an embodiment of the present application, which includes an object terminal 110, the Internet 120, a gateway 130, and a server 140.

[0069] The object terminal 110 is a device used to collect the operating status data of each reactor and other devices in the nuclear power plant. The object terminal 110 can be regarded as a device located in the nuclear power plant. The object terminal 110 includes various forms such as desktop computers, laptops, PDAs (personal digital assistants), and mobile phones. In addition, it can be a single device or a collection of multiple devices. For example, multiple devices are connected through a local area network and use a common display device to work together to form a terminal. The terminal can also communicate with the Internet 120 in a wired or wireless manner to exchange data.

[0070] The gateway 130 is also called an internetwork connector or a protocol converter. The gateway 130 realizes network interconnection at the transport layer and is a computer system or device that acts as a converter. The gateway 130 is a translator between two systems that use different communication protocols, data formats or languages, or even completely different architectures. At the same time, the gateway 130 can also provide filtering and security functions. The message sent by the object terminal 110 to the server 140 is sent to the corresponding server 140 through the gateway 130. The message sent by the server 140 to the object terminal 110 is also sent to the corresponding object terminal 110 through the gateway 130.

[0071] The server 140 refers to a computer system that can provide the target terminal 110 with accident risk assessment services for the nuclear power plant. Compared with the target terminal 110, the server 140 has higher requirements in terms of stability, security, performance, etc. The server 140 can be a high-performance computer in a network platform, a cluster of multiple high-performance computers, a part of a high-performance computer (such as a virtual machine), a combination of parts of multiple high-performance computers (such as virtual machines), etc. The server 140 can also communicate with the Internet 120 in a wired or wireless manner to exchange data.

[0072] In addition, the accident risk assessment method for a nuclear power plant provided in the embodiment of the present application mainly analyzes the risks of serious accidents in nuclear power plants, such as containment failure and large-scale radioactive release. It can analyze the risk of a large amount of released substances caused by a high-pressure melt reactor, and can also analyze the risk of a large amount of released substances caused by a low-pressure melt reactor or other accidents.

[0073] General description of the embodiments of the present application

[0074] The accident risk assessment method for nuclear power plants is to construct a target containment event tree for the analyzed accident, and then conduct an assessment based on the target containment event tree. This method can improve the assessment effect of the accident risk of nuclear power plants, identify the weak links in the design of nuclear power plants, reduce the risk of large-scale releases, and improve the safety of nuclear power plants.

[0075] The accident risk assessment method for a nuclear power plant according to the embodiment of the present application may be executed on the object terminal 110 , may be executed on the server 140 , or may be executed partially on the object terminal 110 and the other partially on the server 140 .

[0076] Reference Figure 2 The present application embodiment provides a method for assessing accident risks in a nuclear power plant, including:

[0077] Step 210: merging the core damage state sequences of the nuclear power plant using the connection event tree to determine the target core damage state sequence.

[0078] Step 220: Use the target core damage state sequence as the starting point of the target containment event tree corresponding to the accident to be analyzed.

[0079] Step 230: Based on the occurrence order of the node events in the accidents of preset risk levels corresponding to the accident to be analyzed, the position of each node event in the target containment event tree is determined, thereby determining multiple paths of the target containment event tree.

[0080] Step 240: Analyze multiple node events to quantify the frequency of accidents corresponding to each path.

[0081] Step 250: Determine the evaluation result of the nuclear power plant under the accident to be analyzed based on the accident occurrence frequencies corresponding to the multiple paths in the target containment event tree.

[0082] It should be noted that core damage refers to the damage to the nuclear fuel elements in the nuclear reactor during operation, which may cause damage or melting of the fuel rods in the core. The core damage state refers to a state when the core damage occurs in a nuclear power plant, and the core damage state sequence refers to the sequence of multiple states when the core damage occurs in a nuclear power plant. The core damage state sequence is usually determined based on the primary circuit integrity, primary circuit pressure, secondary circuit integrity, containment status, etc. when the core of the nuclear power plant is damaged.

[0083] The target core damage state sequence refers to the core damage state sequence that may cause the accident to be analyzed in the nuclear power plant. For example, severe accidents in nuclear power plants can be divided into high-pressure melt and low-pressure melt according to the core melting process, and core damage is very likely to cause the occurrence of high-pressure melt or low-pressure melt. If the accident to be analyzed is a high-pressure melt, since the primary circuit of the melt is in a high-pressure state, it may cause the occurrence of a high-pressure melt. In this case, the core damage state sequence with the primary circuit in a high-pressure state is taken as the target core damage state sequence, and then these core damage state sequences are taken as the starting point of the high-pressure melt containment event tree.

[0084] The connection event tree is connected to multiple containment event trees. The connection event tree can analyze and merge each core damage state sequence to determine the containment event tree corresponding to the core damage state sequence, such as the high-pressure melt reactor containment event tree, the low-pressure melt reactor containment event tree, etc.

[0085] Accidents with a preset risk level refer to accidents in which the containment fails and a large amount of radioactivity is released due to the accident to be analyzed. In the process of the accident to be analyzed leading to the accident with a preset risk level, there are many factors that affect the integrity of the containment, such as Figure 3 The containment isolation, primary circuit depressurization, etc. shown in the embodiment of the present application take the time corresponding to these elements affecting the integrity of the containment as node events, and determine the position of each node event in the target containment event tree according to the order of occurrence of the node event in the preset risk level accident. Afterwards, multiple paths of the target containment event tree are constructed according to whether the node event occurs or not.

[0086] Figure 3 A schematic diagram of a high-pressure reactor containment event tree provided in an embodiment of the present application. Figure 3 As shown in Table 1, the starting point of the target containment event tree is the target core damage state sequence, and the first loop of the target core damage state sequence is in the high pressure state. Figure 3 Node events in the medium and high pressure reactor containment event tree. At each node, if the node event question corresponding to the node is answered affirmatively, it means moving forward, which means that the nuclear power plant status is developing in a better direction. If the node event question corresponding to the node is answered negatively, it means moving downward, which means that the nuclear power plant status is developing in a worse direction.

[0087] Table 1

[0088]

[0089] It should be noted that after determining multiple paths of the target containment event tree, multiple node events can be analyzed to quantify the frequency of accidents corresponding to each path.

[0090] It should be noted that the evaluation results of the nuclear power plant under the accident to be analyzed are determined based on the accident occurrence frequencies corresponding to multiple paths in the target containment event tree. According to the accident occurrence frequencies corresponding to multiple paths in the target containment event tree, the preset risk level accident with the highest frequency of occurrence in the nuclear power plant can be determined, and according to the multiple node events corresponding to the accident occurrence, the weak links in the design of the nuclear power plant can be identified, and then corresponding mitigation measures can be taken to reduce the risk of serious accidents.

[0091] It should be noted that after determining the frequency of accidents corresponding to each path, the consequences of the target containment event tree can be determined by merging release classes and calculating source items. Specifically, the consequences of the target containment event tree are various release classes, which are defined by merging the properties of radioactive materials released when a preset risk level accident occurs in the reactor, that is, a severe accident. Then, based on the results of the source item analysis and the quantification of the frequency of accidents corresponding to each path, the release class can give the release amount and release frequency of each type of related nuclides. For example, if the target containment event tree is a high-pressure containment event tree, then the release class is determined based on the containment failure mode, the containment failure time, and the removal mechanism of fission products in the containment, where the containment failure mode includes containment bypass, containment isolation failure, containment rupture, or bottom plate melt-through. In addition, Figure 3 The consequences corresponding to each path are given, where RC101 means that multiple node events did not occur, so the nuclear power plant did not cause a serious accident. Other paths caused a large number of releases (LRF), and the following numbers, such as RC501 and RC502, are the release types corresponding to each path.

[0092] It should be noted that, for the risk of a large amount of radioactive release caused by the accident to be analyzed, the embodiment of the above steps 210 to 250 first uses the connection event tree to merge the core damage state sequences of the nuclear power plant, determines the target core damage state sequence corresponding to the accident to be analyzed, and then determines the starting point of the target containment event tree, so that the starting point of the target containment event tree is adapted to the nuclear power plant. Afterwards, based on the occurrence order of the node events in the preset risk level accident, the position of each node event in the target containment event tree is determined, and then multiple paths of the target containment event tree are determined. Since the node event is an event that occurs in the nuclear power plant during the preset risk level accident corresponding to the accident to be analyzed, the node event of the target containment event tree can reflect the design characteristics of the nuclear power plant reactor type, and each path can reflect the development process of the preset risk level accident, so that the accident frequency obtained is closer to the actual frequency of the nuclear power plant accident, so the evaluation result obtained is more accurate, and the evaluation effect of the accident risk of the nuclear power plant is improved, so that the weak links in the design of the nuclear power plant can be identified, the risk of a large amount of release is reduced, and the safety of the nuclear power plant is improved.

[0093] It is understandable that, referring to Figure 4 Before step 210, the accident risk assessment method for a nuclear power plant provided in the embodiment of the present application further includes:

[0094] Step 410: Perform a first-level probabilistic safety assessment on the nuclear power plant to obtain multiple accident sequences of core damage in the nuclear power plant.

[0095] Step 420: Determine a list of characteristic quantities according to the reactor type of the nuclear power plant.

[0096] Step 430: For each accident sequence, based on the feature quantity corresponding to each accident in the accident sequence, determine the feature attribute corresponding to the accident in the feature quantity list.

[0097] Step 440: Based on a plurality of characteristic attributes corresponding to each accident sequence, the accident sequences are merged into a core damage state sequence.

[0098] It should be noted that the first-level probabilistic safety assessment (PSA), also often called probabilistic risk assessment, is a risk quantification assessment technology based on probability theory. It links the failure probability of the entire system with the failure probability of its subsystems, components and external conditions at all levels through structural logical reasoning, so as to find out the frequency of various accidents and conduct safety assessment. According to the first-level probabilistic safety assessment results of nuclear power plants, multiple accident sequences of core damage in nuclear power plants can be obtained. The accident sequence is a sequence of multiple events when core damage occurs in nuclear power plants.

[0099] It should be noted that the feature quantity list is determined according to the core of the reactor in the nuclear power plant. The feature quantity list includes feature quantities and feature attributes corresponding to the feature quantities. If the reactor is a pressurized water reactor, the determined feature quantity list is shown in Table 2. The feature quantities include primary circuit integrity, primary circuit pressure, steam generator status, secondary circuit integrity and AC power supply status, and each feature quantity has a corresponding feature attribute. For example, the feature attributes corresponding to the primary circuit integrity include bypass, steam generator heat transfer tube rupture, large and medium rupture, small rupture and integrity.

[0100] Table 2

[0101]

[0102] The accident sequence of core damage in nuclear power plants contains multiple accidents. For each accident sequence, according to the logical order of core damage, the characteristic attributes corresponding to the accident are determined in the characteristic quantity list. Figure 5 The characteristic attribute corresponding to the primary circuit integrity of sequence 6 is steam generator heat transfer tube rupture (SGTR). This accident occurs first, and the corresponding primary circuit pressure, steam generator status, etc. are not applicable. When determining the primary circuit pressure, the characteristic attribute of sequence 4-AT is ultra-high pressure.

[0103] It should be noted that after determining the multiple characteristic attributes corresponding to the accident sequence, the accident sequence can be merged into a core damage state sequence. Figure 5 The tree diagram groups the accident sequences.

[0104] It should be noted that the above-mentioned embodiments of step 410 to step 440 are aimed at each accident sequence of core damage in a nuclear power plant, and the characteristic attributes corresponding to multiple accidents are merged, and the characteristic attributes of the accident are determined based on the characteristic quantity list corresponding to the reactor type in the nuclear power plant, which reflects the design characteristics of the reactor type in the nuclear power plant, and the evaluation results obtained are more accurate. In addition, the accident sequence of core damage in a nuclear power plant is processed by this merging method, so that the core damage state sequence is clearer and more concise, and the efficiency of determining the target core damage state sequence is improved.

[0105] It is understandable that, referring to Figure 6 , step 210 comprises:

[0106] Step 610: For each core damage state sequence, a connection event tree is used to determine a plurality of characteristic attributes corresponding to the core damage state sequence.

[0107] Step 620: If the plurality of characteristic attributes include a target characteristic attribute corresponding to the accident to be analyzed, the core damage state sequence is determined as a target core damage state sequence.

[0108] It should be noted that the target characteristic attribute corresponds to the accident to be analyzed, and is the characteristic attribute corresponding to the occurrence of the accident to be analyzed. If the multiple characteristic attributes corresponding to the core damage state sequence include the target characteristic attribute, the core damage state sequence may eventually lead to an accident of the preset risk level corresponding to the accident to be analyzed.

[0109] Assuming that the accident to be analyzed is a high-pressure melt, the target characteristic attributes are primary high pressure and primary ultra-high pressure, and then the core damage state sequence including primary high pressure or primary ultra-high pressure is determined as the target core damage state sequence. Figure 5 The primary circuit pressure of the core damage state sequence AT is ultra-high pressure, while the primary circuit pressure of the core damage state sequences SG, SL, TP, TR, VS1 and VS2 is high pressure. Therefore, the core damage state sequences AT, SG, SL, TP, TR, VS1 and VS2 are taken as the target core damage state sequences.

[0110] It should be noted that the embodiments of the above-mentioned steps 610 and 620 connect the event tree through multiple characteristic attributes of the core damage state sequence and the target characteristic attributes corresponding to the accident to be analyzed, and determine the target core damage state sequence among multiple core damage state sequences, thereby improving the accuracy of the starting point of the target containment event tree, and further improving the accuracy of the accident risk assessment.

[0111] It is understandable that, referring to Figure 7 After step 620, the method further includes:

[0112] Step 710: for a target core damage state sequence, determine a first characteristic attribute having the highest correlation with the target core damage state sequence from among a plurality of characteristic attributes.

[0113] Step 720: Based on the first characteristic attribute, determine the function header of the target core damage state sequence in the connection event tree.

[0114] Step 730: simulate the target core damage state sequence through the function header, and when the simulation result does not meet the preset condition, use the first characteristic attribute as the boundary condition of the target containment event tree.

[0115] It should be noted that the first characteristic attribute is the characteristic attribute with the highest correlation with the target core damage state sequence and is also the main cause of core damage.

[0116] The functional header is located in the connection event tree, which is used to analyze the functional events required to mitigate severe accidents corresponding to the target core damage state.

[0117] The boundary condition is the boundary condition of the target containment event tree. The boundary condition refers to the state of the target containment event tree to be constructed, and clarifies where the target containment event tree is constructed. In the embodiment of the present application, the boundary condition corresponds to the target core damage state sequence, which can reflect the difference between different target core damage state sequences.

[0118] Reference Figure 5 , the target core damage state sequence TP corresponds to two accident sequences, one of which is that the first circuit is not depressurized, the first circuit has a breach, and the AC power is unavailable, and the other is that the first circuit is depressurized and the AC power is unavailable. It can be determined that the target core damage state sequence TP corresponds to the core damage state sequence caused by the power outage of the nuclear power plant, and the first characteristic attribute is that the AC power is unavailable. After the power outage of the nuclear power plant, the AC power needs to be restored to provide power for subsequent core cooling and containment depressurization operations to alleviate the deterioration of the accident. Based on this, the function header "POW_REC" is set to simulate the restoration of AC power before the battery is exhausted. If the simulation result does not meet the preset conditions, that is, the AC power is not restored, the unavailable AC power is used as the boundary condition of the target containment event tree. If the AC power is restored, there is no need to set the boundary condition.

[0119] Reference Figure 5 , the target core damage state sequence VS1 corresponds to the primary circuit without overpressure and the breach outside the containment. It can be determined that the target core damage state sequence VS1 corresponds to the melt sequence caused by the breach inside the secondary containment. The target core damage state sequence VS1 has the risk of early overpressure of the containment after deterministic analysis. Typical sequences include the breach inside the containment of the main steam pipe or the main feed water pipe, the failure of direct injection of the safety spray, and the breach inside the containment of the main steam pipe or the main feed water pipe, and the failure of isolation of the main steam or main feed water. Based on this, the function header "CF_P" is set to containment without early overpressure failure. If the containment pressure sensor detects that the containment pressure reaches the containment spray setting value, the safety spray is automatically triggered or the safety spray can be manually started successfully within 1 hour, then the containment will not have early overpressure. Considering conservatively, it is transferred to the high-pressure melt reactor containment event tree, and the boundary condition "break outside the containment" is set to select the target core damage state sequence VS1. If the safety spray fails to be automatically triggered or the manual start of the safety spray fails within 1 hour, the containment will fail to be overpressured and an early large-scale release will occur.

[0120] It should be noted that the embodiments of the above steps 710 to 730 determine the functional header of the target core damage state sequence in the connection event tree based on the first characteristic attribute with the highest correlation with the target core damage state sequence. Afterwards, the target core damage state sequence is simulated through the functional header, and when the simulation result does not meet the preset conditions, the first characteristic attribute is used as the boundary condition of the target containment event tree. This method sets the boundary conditions based on the first characteristic attribute of the target core damage state sequence, which can reflect the differences between different target core damage state sequences, thereby improving the accuracy of accident risk assessment.

[0121] It is understandable that, referring to Figure 8 , step 240 comprises:

[0122] Step 810: For each path in the target containment event tree, determine multiple node events.

[0123] Step 820: Analyze each node event using determinism and probability theory to obtain a first probability density function and a second probability density function corresponding to the node event. The first probability density function is used to represent the mapping relationship between the containment load and the first probability density, and the second probability density function is used to represent the mapping relationship between the containment load capacity and the second probability density.

[0124] Step 830: Calculate the failure probability that the containment load is greater than the containment bearing capacity based on the first probability density function and the second probability density function.

[0125] Step 840: Determine the accident occurrence frequency corresponding to the path based on the failure probabilities of multiple node events and the occurrence frequency of the target core damage state sequence.

[0126] It should be noted that the embodiment of the present application combines determinism and probability theory to analyze the probability of occurrence of each node event, and obtains a first probability density function and a second probability density function corresponding to the node event. Among them, the first probability density function is used to represent the mapping relationship between the containment load and the first probability density, and the second probability density function is used to represent the mapping relationship between the containment load capacity and the second probability density.

[0127] If a node event causes containment failure, then the containment load corresponding to the node event is greater than the containment bearing capacity. Therefore, the failure probability of the node event can be determined by calculation based on the first probability density function and the second probability density function.

[0128] Reference Fig. 9 , Pr(Pc) is the first probability density function, and Pr(Pf) is the second probability density function. If Pr(Pc)>Pr(Pf), the containment fails. Therefore, we can calculate Fig. 9The shaded area in determines the containment failure probability. The failure probability of a node event can be expressed as:

[0129]

[0130] It should be noted that, when determining the failure probabilities of multiple node events corresponding to the path, the frequency of accidents corresponding to the path can be determined based on the failure probabilities of multiple events and the frequency of occurrence of the target core damage state sequence. The frequency of occurrence of the target core damage state sequence can be determined based on the ratio of the number of target core damage state sequences to the total number of multiple core damage state sequences, and its specific value can be determined through the first-level probabilistic safety assessment.

[0131] Reference Fig.10 , the starting point of the target containment event tree is CDSi, and the node events include A, B, C and D. Among them, if the node event question corresponding to the node is answered affirmatively, then forward, it means that the state of the nuclear power plant is developing in a better direction, if the node event question corresponding to the node is answered negatively, then downward, it means that the state of the nuclear power plant is developing in a worse direction. Sequence se l 1 is that the question corresponding to node event A is answered affirmatively, the question corresponding to node event C is answered affirmatively, and the question corresponding to node event D is answered negatively. Then the accident frequency of sequence se l is:

[0132]

[0133] in, represents the frequency of node event occurrence, node event C occurrence, and node event D non-occurrence, f(CDSi) is the occurrence frequency of the target core damage state sequence CDSi, which can be determined by the first-level probabilistic safety assessment, Q(A / CDSi) is the probability of node event A occurrence based on the occurrence of the target core damage state sequence, that is, the difference between 1 and the failure probability of node event A, Q(C / CDSi∪A) is the probability of node event C occurrence based on the occurrence of the target core damage state sequence or A occurrence, that is, the difference between 1 and the failure probability of node event C, is the probability that node event D does not occur based on the occurrence of the target core damage state sequence, or A occurs, or C occurs, that is, the failure probability of node event D. Based on this, the accident occurrence frequency corresponding to multiple paths in the target containment event tree can be determined.

[0134] It should be noted that the embodiments of the above steps 810 to 840 use determinism and probability theory to analyze each node event, determine the probability of containment failure of each node event, and then determine the accident frequency corresponding to each path. The accident frequency is closer to the actual frequency of nuclear power plant accidents, so the evaluation result is more accurate, which improves the evaluation effect of nuclear power plant accident risks, and then can identify the weak links in the design of nuclear power plants, reduce large-scale release risks, and improve the safety of nuclear power plants.

[0135] It is understandable that, referring to Fig.11 , step 820 includes:

[0136] Step 1110: Perform a qualitative analysis on each node event to obtain multiple impact factors corresponding to the node event, and determine a third probability density function of each impact factor.

[0137] Step 1120: Perform Monte Carlo simulation calculation on the multiple third probability density functions to obtain a first probability density function and a second probability density function.

[0138] It should be noted that qualitative analysis is an analysis method that infers the nature and development trend of things based on the subjective judgment and analysis ability of the forecaster. In the process of qualitative analysis, engineering experience and expert judgment can be combined to determine multiple influencing factors corresponding to node events.

[0139] Table 3 is the analysis result of the steam explosion in the pressure vessel of the node event. The original reason for the failure of the containment due to the steam explosion in the pressure vessel of the node event is that the hot molten material falls into the volatile coolant in the pressure vessel during the melting process and produces a violent reaction. The corresponding multiple influencing factors include the load formed by the steam explosion, the proportion of the core mass involved in the premixing of the steam explosion, the load threshold of the upper head of the pressure vessel, etc. Then, the load formed by the steam explosion is determined as:

[0140] E L =m c ×f p ×E c ×C tm ×C ms

[0141] Among them, E L is the load caused by steam explosion, m c is the total mass of the core, f p is the core mass proportion participating in steam explosion premixing, E c is the core heat storage per unit mass, C tm is the thermal energy-mechanical energy conversion rate, C ms is the mechanical energy conversion rate. Except for the total mass of the core, all other values ​​are assigned in the form of probability distribution, that is, the form of the third probability density function.

[0142] In addition, the failure probability corresponding to the node event steam explosion in the pressure vessel is:

[0143] Q F =Q INEX ×Q(E L >E LC )

[0144] Among them, Q F is the probability of pressure vessel failure caused by steam explosion, Q INEX is the probability of steam explosion, E L is the load caused by steam explosion, E LC It is the load threshold of the upper head of the pressure vessel.

[0145] Afterwards, Monte Carlo simulation calculations are performed on multiple third probability density functions to determine the load distribution of the upper and lower heads due to steam explosion under high pressure scenarios, as well as the load-bearing capacity distribution of the upper and lower heads of the pressure vessel, i.e., the first probability density function and the second probability density function of the upper and lower heads. Afterwards, the failure probability of the upper and lower heads caused by steam explosion in the pressure vessel under high pressure scenarios is calculated, and assuming that the containment will also fail when the upper head fails, the failure probability of the containment can be determined.

[0146] Table 3

[0147]

[0148] It should be noted that the independent variable of the third probability density function is the value related to the third influencing factor, and the dependent variable is the probability.

[0149] It should be noted that the embodiments of the above steps 1110 and 1120 perform qualitative analysis on each node event, obtain multiple influencing factors corresponding to the node event, and determine the third probability density function of each influencing factor. Then, a Monte Carlo simulation calculation is performed on the multiple third probability density functions to obtain the first probability density function and the second probability density function, thereby improving the accuracy of the failure probability corresponding to the node event and the accuracy of the accident risk assessment.

[0150] It is understandable that, referring to Fig.12 , step 250 comprises:

[0151] Step 1210: Determine a target path from among the multiple paths based on the accident occurrence frequencies corresponding to the multiple paths in the target containment event tree.

[0152] Step 1220: Determine multiple mitigation node events from multiple node events corresponding to the target path.

[0153] Step 1230: Obtain the mitigation time window corresponding to each mitigation node event.

[0154] Step 1240: Use multiple mitigation node events and mitigation time windows corresponding to the mitigation node events as evaluation results of the nuclear power plant under the accident to be analyzed, so as to execute the mitigation node events within the mitigation event window.

[0155] It should be noted that multiple target paths refer to multiple paths with the highest accident frequency in the target containment event tree.

[0156] Mitigation node events refer to node events that can mitigate severe accidents, that is, accidents with preset risk levels. In the high-pressure reactor containment event tree analysis, the mitigation measures considered are: primary circuit depressurization, water supply to steam generators, containment isolation, low-pressure injection and temporary primary circuit water injection, containment spraying and temporary spraying, and containment filtration and discharge. The mitigation event window refers to the time window in which the mitigation node event is executed. Severe accidents can only be mitigated if the mitigation node event is executed within the mitigation time window.

[0157] It should be noted that, refer to Figure 3 The frequency of large-scale release caused by high-pressure melting reactor is 1.59E-06 / reactor-year, and there are 9 event sequences with a frequency higher than 1E-08 / reactor-year. The top 3 event sequences, i.e., the event sequences corresponding to the target path, can be determined as shown in Table 4. The top 3 dominant cut sets, i.e., the minimum equipment failure sequence that causes containment failure, are shown in Figure 5 shown.

[0158] Table 4

[0159]

[0160] Table 5

[0161]

[0162]

[0163] It should be noted that, according to Table 4, it can be determined that for high-pressure melt reactor accidents, core reflooding failure, core melt-concrete interaction (MCCI) phenomenon in the containment, and bottom plate melt penetration are the main contributing factors. The core reflooding failure is due to the failure of primary water injection, and the temporary water injection of the primary water injection was not put into operation in time. The main reason for the failure of primary water injection is that when the primary pressure relief fails, the low-pressure injection cannot be put into operation, or the core temperature monitoring signal cannot be correctly sent out due to the non-self-detectable fault of the digital instrumentation and control system (DCS) processing unit, and the primary water injection cannot be manually implemented. Secondly, the containment isolation failure and the failure to start spraying when it fails also contribute to the large-scale release. Bottom plate melt penetration is an inevitable phenomenon in existing CPR1000 nuclear power plants.

[0164] Therefore, reducing the risk of large-scale radioactive release from high-pressure melt reactors requires focusing on core reflooding and primary circuit depressurization. During the process of core reflooding, successful primary circuit water injection can take away the core heat release, cool the core, restore the core reflooding state, avoid pressure vessel rupture, and thus maintain the integrity of the containment. The premise for the implementation of this mitigation measure is the successful primary circuit depressurization. Optimizing the design and accident handling strategy, providing reliable water injection means under severe accidents, and timely injection can effectively increase the success probability of primary circuit water injection. If the primary circuit depressurization is successfully carried out, the risk of high-pressure melt reactors can be transformed into the risk of low-pressure melt reactors, avoiding the consequences of high-pressure melt reactors. In addition, the successful primary circuit depressurization is the premise for the subsequent mitigation measures such as core reflooding. Therefore, improving the effectiveness of primary circuit depressurization plays an important role in mitigating the risk of large-scale release caused by high-pressure melt reactors.

[0165] It is to be understood that, although the steps in the above-mentioned flowcharts are sequentially displayed according to the characterization of arrows, these steps are not necessarily executed in sequence according to the order of arrow characterization. Unless there is a clear description in the present embodiment, the execution of these steps does not have a strict order restriction, and these steps can be executed in other orders. Moreover, at least a portion of the steps in the above-mentioned flowcharts can include multiple steps or multiple stages, and these steps or stages are not necessarily executed at the same time, but can be executed at different times, and the execution order of these steps or stages is not necessarily to be carried out in sequence, but can be executed in turn or alternately with other steps or at least a portion of the steps or stages in other steps.

[0166] It should be noted that in each specific implementation of the present application, when it comes to the need to perform relevant processing based on data related to the characteristics of the target object such as the target object attribute information or attribute information set, the permission or consent of the target object will be obtained first, and the collection, use and processing of these data will comply with relevant laws, regulations and standards. In addition, when the embodiment of the present application needs to obtain the attribute information of the target object, the separate permission or separate consent of the target object will be obtained through a pop-up window or jump to a confirmation page. After clearly obtaining the separate permission or separate consent of the target object, the necessary target object-related data used to enable the normal operation of the embodiment of the present application is obtained.

[0167] Description of the apparatus and device of the present application

[0168] Fig.13 The schematic diagram of the structure of the accident risk assessment device of a nuclear power plant provided in the embodiment of the present application, the accident risk assessment device 1300 of the nuclear power plant includes:

[0169] The sequence merging unit 1310 is used to merge the core damage state sequences of the nuclear power plant by using the connection event tree to determine the target core damage state sequence.

[0170] The starting point determination unit 1320 is used to use the target core damage state sequence as the starting point of the target containment event tree corresponding to the accident to be analyzed.

[0171] The path determination unit 1330 is used to determine the position of each node event in the target containment event tree based on the occurrence order of the node events in the preset risk level accidents corresponding to the accident to be analyzed, thereby determining multiple paths of the target containment event tree.

[0172] The analysis and quantification unit 1340 is used to analyze multiple node events to quantify the frequency of accidents corresponding to each path.

[0173] The risk assessment unit 1350 is used to determine the assessment result of the nuclear power plant under the accident to be analyzed based on the accident occurrence frequencies corresponding to multiple paths in the target containment event tree.

[0174] Reference Fig.14 , Fig.14The structural block diagram of the terminal part of the accident risk assessment method of the nuclear power plant in the embodiment of the present application is as follows: the terminal includes: a radio frequency (RF) circuit 1410, a memory 1415, an input unit 1430, a display unit 1440, a sensor 1450, an audio circuit 1460, a wireless fidelity (WiFi) module 1470, a processor 1480, and a power supply 1490. Those skilled in the art can understand that Fig.14 The terminal structure shown does not constitute a limitation on the mobile phone or computer, and may include more or fewer components than shown in the figure, or combine certain components, or arrange the components differently.

[0175] The RF circuit 1410 may be used for receiving and sending signals during information transmission or communication or during a call. In particular, after receiving downlink information from a base station, the information is sent to the processor 1480 for processing. In addition, the uplink data is sent to the base station.

[0176] The memory 1415 may be used to store software programs and modules. The processor 1480 executes various functional applications and data processing of the content terminal by running the software programs and modules stored in the memory 1415 .

[0177] The input unit 1430 may be used to receive input digital or character information and generate key signal input related to the setting and function control of the content terminal. Specifically, the input unit 1430 may include a touch panel 1431 and other input devices 1432 .

[0178] The display unit 1440 may be used to display input information or provided information and various menus of the content terminal. The display unit 1440 may include a display panel 1441.

[0179] The audio circuit 1460 , the speaker 1461 , and the microphone 1462 may provide an audio interface.

[0180] In this embodiment, the processor 1480 included in the terminal can execute the accident risk assessment method of the nuclear power plant in the previous embodiment.

[0181] The terminals of the embodiments of the present application include but are not limited to mobile phones, computers, intelligent voice interaction devices, smart home appliances, vehicle terminals, aircraft, etc. The embodiments of the present invention can be applied to various scenarios, including but not limited to data collection, data export, data analysis, etc.

[0182] Fig.15A block diagram of the structure of a portion of a server 140 for implementing the accident risk assessment method for a nuclear power plant according to an embodiment of the present application. The server 140 may have relatively large differences due to different configurations or performances, and may include one or more central processing units (CPUs) 1522 (e.g., one or more processors) and a memory 1532, and one or more storage media 1530 (e.g., one or more mass storage devices) storing application programs 1542 or data 1544. Among them, the memory 1532 and the storage medium 1530 may be temporary storage or permanent storage. The program stored in the storage medium 1530 may include one or more modules (not shown in the figure), and each module may include a series of instruction operations on the server 140. Furthermore, the central processing unit 1522 may be configured to communicate with the storage medium 1530 and execute a series of instruction operations in the storage medium 1530 on the server 140.

[0183] The server 140 may also include one or more power supplies 1526, one or more wired or wireless network interfaces 1550, one or more input and output interfaces 1558, and / or one or more operating systems 1541, such as Wi ndows Server™, Mac OS X™, Unix™, Linux™, FreeBSD™, etc.

[0184] The central processor 1522 in the server 140 can be used to execute the accident risk assessment method for a nuclear power plant according to the embodiment of the present application.

[0185] An embodiment of the present application also provides a computer-readable storage medium, which is used to store program code, and the program code is used to execute the accident risk assessment method of a nuclear power plant in each of the aforementioned embodiments.

[0186] The embodiment of the present application also provides a computer program product, which includes a computer program. The processor of the computer device reads and executes the computer program, so that the computer device executes the above-mentioned accident risk assessment method for a nuclear power plant.

[0187] The terms "first", "second", "third", "fourth", etc. (if any) in the specification of the present application and the above-mentioned drawings are used to distinguish similar contents, and are not necessarily used to describe a specific order or sequential order. It should be understood that the data used in this way can be interchangeable where appropriate, so that the embodiments of the present application described herein can be implemented in an order other than those illustrated or described herein, for example. In addition, the terms "comprise" and "comprising" and any of their variations are intended to cover non-exclusive inclusions, for example, a process, method, system, product or device comprising a series of steps or units is not necessarily limited to those steps or units clearly listed, but may include other steps or units that are not clearly listed or inherent to these processes, methods, products or devices.

[0188] It should be understood that in the present application, "at least one (item)" means one or more, and "plurality" means two or more. "And / or" is used to describe the association relationship of associated content, indicating that three relationships may exist. For example, "A and / or B" can mean: only A exists, only B exists, and A and B exist at the same time, where A and B can be singular or plural. The character " / " generally indicates that the associated content before and after is in an "or" relationship. "At least one of the following" or similar expressions refers to any combination of these items, including any combination of single or plural items. For example, at least one of a, b or c can mean: a, b, c, "a and b", "a and c", "b and c", or "a and b and c", where a, b, c can be single or multiple.

[0189] It should be understood that in the description of the embodiments of the present application, the meaning of multiple (or multiple items) is more than two, greater than, less than, exceed, etc. are understood to not include the number, and above, below, within, etc. are understood to include the number.

[0190] In the several embodiments provided in the present application, it should be understood that the disclosed systems, devices and methods can be implemented in other ways. For example, the device embodiments described above are only schematic, for example, the division of units is only a logical function division, and there may be other division methods in actual implementation, such as multiple units or components can be combined or integrated into another system, or some features can be ignored or not executed. In addition, the mutual coupling or direct coupling or communication connection shown or discussed can be an indirect coupling or communication connection through some interfaces, devices or units, which can be electrical, mechanical or other forms.

[0191] The units described as separate components may or may not be physically separated, and the components shown as units may or may not be physical units, that is, they may be located in one place or distributed on multiple network units. Some or all of the units may be selected according to actual needs to achieve the purpose of the solution of this embodiment.

[0192] In addition, each functional unit in each embodiment of the present application may be integrated into one processing unit, or each unit may exist physically separately, or two or more units may be integrated into one unit. The above-mentioned integrated unit may be implemented in the form of hardware or in the form of software functional units.

[0193] If the integrated unit is implemented in the form of a software functional unit and sold or used as an independent product, it can be stored in a computer-readable storage medium. Based on this understanding, the technical solution of the present application is essentially or the part that contributes to the relevant technology or all or part of the technical solution can be embodied in the form of a software product, and the computer software product is stored in a storage medium, including a number of instructions to enable a computer device (which can be a personal computer, server 140, or network device, etc.) to perform all or part of the steps of the various embodiments of the present application. The aforementioned storage medium includes: U disk, mobile hard disk, read-only memory (Read-Only Memory, referred to as ROM), random access memory (Random Access Memory, referred to as RAM), disk or optical disk and other media that can store program codes.

[0194] It should also be understood that the various implementations provided in the embodiments of the present application can be combined arbitrarily to achieve different technical effects.

[0195] The above is a specific description of the implementation method of the present application, but the present application is not limited to the above-mentioned implementation method. Technical personnel familiar with the field can also make various equivalent modifications or substitutions without violating the spirit of the present application. These equivalent modifications or substitutions are all included in the scope defined by the claims of the present application.

Claims

1. A method for assessing accident risk in a nuclear power plant, characterized in that: include The connection event tree is used to merge the core damage state sequences of nuclear power plants and determine the target core damage state sequence. Taking the target core damage state sequence as the starting point of the target containment event tree corresponding to the accident to be analyzed; Based on the occurrence order of node events in the accidents with preset risk levels corresponding to the accident to be analyzed, determining the position of each node event in the target containment event tree, thereby determining multiple paths of the target containment event tree; Analyzing a plurality of the node events to quantify the frequency of accidents corresponding to each of the paths; Based on the accident occurrence frequencies corresponding to the plurality of paths in the target containment event tree, an evaluation result of the nuclear power plant under the accident to be analyzed is determined.

2. The method according to claim 1, characterized in that Before merging the core damage state sequences of the nuclear power plant using the connection event tree to determine the target core damage state sequence, the method further includes: Conducting a first-level probabilistic safety assessment on a nuclear power plant to obtain multiple accident sequences of core damage in the nuclear power plant; Determine a list of characteristic quantities according to the reactor type in the nuclear power plant; For each of the accident sequences, based on the feature quantity corresponding to each accident in the accident sequence, determining a feature attribute corresponding to the accident in the feature quantity list; Based on a plurality of the characteristic attributes corresponding to each of the accident sequences, the accident sequences are merged into a core damage state sequence.

3. The method according to claim 1, characterized in that The method of merging the core damage state sequences of the nuclear power plant by using the connection event tree to determine the target core damage state sequence includes: For each of the core damage state sequences, a connection event tree is used to determine a plurality of characteristic attributes corresponding to the core damage state sequence; If the plurality of characteristic attributes include a target characteristic attribute corresponding to the accident to be analyzed, the core damage state sequence is determined as a target core damage state sequence.

4. The method according to claim 3, characterized in that If the plurality of characteristic attributes include a target characteristic attribute corresponding to the accident to be analyzed, after determining the core damage state sequence as a target core damage state sequence, the method further includes: For the target core damage state sequence, determining a first characteristic attribute having the highest correlation with the target core damage state sequence from among the plurality of characteristic attributes; Based on the first characteristic attribute, determining a functional header of the target core damage state sequence in the connection event tree; The target core damage state sequence is simulated through the functional header, and when the simulation result does not meet the preset condition, the first characteristic attribute is used as the boundary condition of the target containment event tree.

5. The method according to claim 1, characterized in that The analyzing of the plurality of node events to quantify the frequency of accidents corresponding to each path includes: For each of the paths in the target containment event tree, determining a plurality of node events; Analyze each of the node events using determinism and probability theory to obtain a first probability density function and a second probability density function corresponding to the node event, wherein the first probability density function is used to represent a mapping relationship between a containment load and a first probability density, and the second probability density function is used to represent a mapping relationship between a containment load capacity and a second probability density; Calculating, based on the first probability density function and the second probability density function, a failure probability that the containment load is greater than the containment bearing capacity; Based on the failure probabilities of the plurality of node events and the occurrence frequency of the target core damage state sequence, the accident occurrence frequency corresponding to the path is determined.

6. The method according to claim 5, characterized in that The method of analyzing each of the node events by using determinism and probability theory to obtain a first probability density function and a second probability density function corresponding to the node event includes: Performing a qualitative analysis on each of the node events to obtain a plurality of influencing factors corresponding to the node event, and determining a third probability density function of each of the influencing factors; Monte Carlo simulation calculation is performed on the plurality of third probability density functions to obtain a first probability density function and a second probability density function.

7. The method according to claim 1, characterized in that The determining, based on the accident occurrence frequencies corresponding to the plurality of paths in the target containment event tree, an evaluation result of the nuclear power plant under the accident to be analyzed, comprises: Determining a target path among the multiple paths based on the accident occurrence frequencies corresponding to the multiple paths in the target containment event tree; Determining a plurality of mitigation node events from a plurality of node events corresponding to the target path; Obtaining a mitigation time window corresponding to each of the mitigation node events; The plurality of mitigation node events and the mitigation time windows corresponding to the mitigation node events are used as evaluation results of the nuclear power plant under the accident to be analyzed, so as to execute the mitigation node events within the mitigation event window.

8. An accident risk assessment device for a nuclear power plant, characterized in that: include: A sequence merging unit, used to merge various core damage state sequences of the nuclear power plant using a connection event tree to determine a target core damage state sequence; A starting point determination unit, configured to use the target core damage state sequence as the starting point of a target containment event tree corresponding to the accident to be analyzed; A path determination unit, configured to determine the position of each of the node events in the target containment event tree based on the occurrence order of the node events in the accidents of preset risk levels corresponding to the accident to be analyzed, thereby determining multiple paths of the target containment event tree; An analysis and quantification unit, used for analyzing the plurality of node events to quantify the frequency of accidents corresponding to each of the paths; The risk assessment unit is used to determine the assessment result of the nuclear power plant under the accident to be analyzed based on the accident occurrence frequency corresponding to the multiple paths in the target containment event tree.

9. An electronic device comprising a memory and a processor, wherein the memory stores a computer program, wherein: When the processor executes the computer program, the accident risk assessment method for a nuclear power plant according to any one of claims 1 to 7 is implemented.

10. A computer-readable storage medium storing a computer program, characterized in that: When the computer program is executed by a processor, the accident risk assessment method for a nuclear power plant according to any one of claims 1 to 7 is implemented.

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