A method for diagnosing and handling loss of heat sink accidents in sodium-cooled fast reactors

Abstract

The present invention relates to a method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident, comprising: (1) diagnosing a sodium-cooled fast reactor loss of heat sink accident, wherein the diagnostic strategy is as follows: among n steam generators, at least m steam generators have an outlet sodium temperature T i Rising above the threshold value θ; (2) determining the loop type of the sodium-cooled fast reactor nuclear power plant, for single-loop design or multi-loop design of the sodium-cooled fast reactor nuclear power plant, respectively using the corresponding process to handle the accident, at the same time, the complex and numerous sodium-cooled fast reactor nuclear power plant loss of heat sink type primary cause events are classified into two categories: partial loss of heat sink and total loss of heat sink, and a matching accident handling strategy is provided, thereby improving the efficiency and reliability of accident handling. The present invention can improve the degree of power plant automation operation, reduce the human influence in the power plant operation and accident diagnosis and processing process, improve the accident response speed, reduce the operator's operating task burden, and ensure the accuracy and timeliness of accident response.

Description

Technical Field

[0001] The invention belongs to nuclear power plant design technology, and in particular relates to a method for diagnosing and processing a sodium-cooled fast reactor loss of heat sink accident. Background Art

[0002] Defense in depth is a safety design requirement that is implemented throughout the safety regulations for nuclear power plant design (HAF). As the third line of defense in the principle of defense in depth, accident procedures play a vital role in limiting the development of accidents and ensuring reactor safety. Once an event or accident condition occurs, the nuclear power plant is considered to have entered an emergency state, requiring operators to use accident procedures to respond quickly, correct deviations from the operating conditions, and take necessary actions to avoid or mitigate possible consequences.

[0003] The heat sink composed of three circuits and steam turbines in sodium-cooled fast reactor nuclear power plants is one of the key functions for the safe operation of nuclear power plants and an important way to achieve core cooling in nuclear power plants. Accidents that lead to failure of the three circuits or steam turbines will cause partial or complete loss of the heat sink function. If no measures are taken, the core cooling capacity will be reduced, threatening the safety of the core and having an adverse impact on the safety of the reactor. Therefore, the heat sink function is directly related to the safe operation of nuclear power plants.

[0004] Due to the characteristics of sodium-cooled fast reactor nuclear power plants, their heat sink design generally adopts a design scheme of multiple loops and multiple steam generator modules, which makes the types of accidents in sodium-cooled fast reactor nuclear power plants complicated. Not only do the failures or accident conditions of the three-circuit and steam turbine directly cause the loss of heat sinks, but the off-site power and compressed air systems can also cause the partial or complete loss of heat sinks. Regardless of whether it is an accident that directly causes the loss of the three-circuit and steam turbine functions, or an accident that indirectly affects the heat sink function of the sodium-cooled fast reactor, the support of the heat sink loss handling strategy is required in the accident handling and the formulation of accident procedures.

[0005] At the same time, fast reactor nuclear power plants have complex systems and many operating parameters. The number of systems and equipment involved in the handling of heat sink loss accidents is large, and the operation sequence, operation timing and mutual cooperation of operators of each device in the accident handling must also meet complex and precise requirements.

[0006] Therefore, given the importance and complexity of the heat sink function of sodium-cooled fast reactor nuclear power plants, it is necessary to establish a set of reasonable and feasible treatment methods and strategies for loss of heat sink accidents, so as to better control the accident process, mitigate the consequences of the accident, and ensure the safety of the unit. Summary of the invention

[0007] The purpose of the present invention is to provide a method for diagnosing and processing a sodium-cooled fast reactor loss of heat sink accident in view of the design characteristics of a sodium-cooled fast reactor nuclear power plant, so as to improve the efficiency and reliability of accident processing and ensure the safety of the unit.

[0008] The technical solution of the present invention is as follows: A method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident comprises the following steps:

[0009] (1) To diagnose the loss of heat sink accident in a sodium-cooled fast reactor, the diagnostic strategy is as follows: among n steam generators, at least m steam generators have an outlet sodium temperature T i rises above the threshold value θ,

[0010]

[0011] Where n is the number of steam generator modules designed for actual sodium-cooled fast reactor nuclear power plants;

[0012] (2) Determine the loop type of the sodium-cooled fast reactor nuclear power plant and adopt the corresponding process to handle accidents for sodium-cooled fast reactor nuclear power plants with single-loop design or multi-loop design.

[0013] Furthermore, in the method for diagnosing and handling a loss of heat sink accident in a sodium-cooled fast reactor as described above, in step (1), the threshold value θ should not be higher than the threshold value required for protective shutdown, and m should not be higher than the number of modules whose outlet sodium temperature of the steam generator exceeds the required threshold value required for protective shutdown; and the threshold value θ should not be higher than the threshold value required for the operation of a dedicated safety facility, and m should not be higher than the number of modules whose outlet sodium temperature of the steam generator exceeds the required threshold value required for the operation of a dedicated safety facility.

[0014] Further, in the above-mentioned method for diagnosing and treating the loss of heat sink accident of a sodium-cooled fast reactor, in step (1), if the sodium-cooled fast reactor nuclear power plant is designed with multiple loops, the T that can characterize the loss of heat sink in any loop should be considered. i The combination of , m and θ, the diagnostic strategy is expressed as:

[0015] Σn j =n

[0016] Among them, T ji 、m j and θ j is the outlet sodium temperature, number and threshold of the steam generator in the jth loop; n j is the number of steam generator modules in the j-th loop;

[0017] The definition of total loss of heat sinks should be based on the actual design of sodium-cooled fast reactor nuclear power plants and the most conservative requirements should be planned from the following principles:

[0018] 1. All loops lose heat sinks;

[0019] 2. The sum of the thermal conductivity of the remaining loops that have not lost their heat sinks does not meet the requirements for safe shutdown or the needs of handling the current accident conditions.

[0020] Furthermore, if the sodium-cooled fast reactor nuclear power plant is designed with multiple loops, m j and θ j The value of should not be lower than the requirement of the protection logic for the loss of heat sink in the jth loop.

[0021] Further, in the above-mentioned method for diagnosing and treating the loss of heat sink accident of a sodium-cooled fast reactor, in step (1), if there is a design in which the steam generator modules are not completely independent in the design of the system flow of the secondary or tertiary circuits of the sodium-cooled fast reactor nuclear power plant, then according to the impact of the loss of heat sink of a steam generator module caused by different primary causes on other modules, T is analyzed and reasonably adjusted. i , the combination of m and θ, and the values ​​of m and θ.

[0022] The primary cause of the accident shall at least include:

[0023] Steam turbine shutdown;

[0024] Steam generator feed water loss;

[0025] Loss of off-site power;

[0026] Failure of steam pressure regulating devices resulting in reduced steam flow;

[0027] Condenser vacuum failure;

[0028] Unintentional closure of steam isolation valves;

[0029] Water pipe leakage;

[0030] Steam pipe leaks;

[0031] For each primary cause accident, analyze its impact on the loop and the number of steam generator modules affected.

[0032] Furthermore, in the above-mentioned method for diagnosing and handling the loss of heat sink accident of a sodium-cooled fast reactor, the diagnostic strategy described in step (1) should only respond to the accident condition, and should not respond to abnormal operating events below the accident condition or transients that deviate from normal operation, so as to avoid adverse effects on the operation of the power plant.

[0033] Further, in the above-mentioned method for diagnosing and handling a loss of heat sink accident in a sodium-cooled fast reactor, in step (2), for a sodium-cooled fast reactor nuclear power plant with a multi-loop design having a total loss of heat sink accident, or a sodium-cooled fast reactor nuclear power plant with a single-loop design having a loss of heat sink accident, the accident handling process is as follows:

[0034] (I) If the total loss of heat sink occurs, check and confirm that the dedicated safety facilities for emergency waste heat removal have been put into operation and that available waste heat removal pathways have been established; otherwise, skip this step;

[0035] (II) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include isolation of feedwater and steam, and check whether other normally open pipes connected to the steam generator have been isolated;

[0036] (III) Isolation of the loop where the accident occurred shall include at least water isolation and steam isolation;

[0037] (IV) If the power plant has a secondary or tertiary system flow design in which the steam generator modules are not completely independent, it has been checked that the connecting pipes between the steam generator module where the accident occurred and the affected steam generator module have been isolated.

[0038] Further, in the above-mentioned method for diagnosing and handling a loss of heat sink accident in a sodium-cooled fast reactor, in step (2), for a partial loss of heat sink accident in a sodium-cooled fast reactor nuclear power plant with a multi-loop design, the accident handling process is as follows:

[0039] (a) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include isolation of feedwater and steam, and check whether other normally open pipes connected to the steam generator have been isolated;

[0040] (b) Isolation of the loop where the accident occurs shall include at least water supply isolation and steam isolation;

[0041] (c) If the power plant has a secondary or tertiary system flow design in which the steam generator modules are not completely independent, it has been checked that the connecting pipes between the steam generator module where the accident occurred and the affected steam generator module have been isolated.

[0042] (d) Check and confirm that the water level in the deaerator is within the normal range. If the water level is insufficient, add water to the normal range;

[0043] (e) Determine the feedwater flow demand based on the current power plant operation requirements, and then determine the number of feedwater pumps that need to be operated, and configure the corresponding feedwater pumps for operation;

[0044] (f) Check that the pressure of the water supply system is within the normal range; if not, check whether the upstream system is operating normally;

[0045] (g) According to the power plant design, the feed water flow rate shall be controlled manually or automatically to restore the steam generator outlet temperature to the normal range.

[0046] The beneficial effects of the present invention are as follows:

[0047] 1. The present invention combines the design characteristics of fast reactor nuclear power plants and proposes a method and strategy for handling the loss of heat sink accident in sodium-cooled fast reactors, which effectively improves the operational safety of fast reactor nuclear power plants;

[0048] 2. The present invention has good versatility. The diagnostic strategy and processing strategy given by the method take into account the design combination of multiple steam generator modules and multiple loops of the three-loop fast reactor power plant. The specific fast reactor power plant design can give corresponding specific strategies according to the method, so that the loss of heat sink accident processing strategy of the sodium-cooled fast reactor power plant applicable to various heat sink system design schemes can be developed through the method;

[0049] 3. The present invention solves the problem of accident diagnosis and accident handling of heat sink loss accidents in sodium-cooled fast reactor nuclear power plants. The designed strategy can improve the handling measures of the unit under accident conditions and ensure the accuracy and timeliness of accident response;

[0050] 4. The present invention classifies the complex and numerous primary cause events of loss of heat sink in sodium-cooled fast reactor nuclear power plants into two categories: partial loss of heat sink and total loss of heat sink, and provides a matching accident handling strategy, which is conducive to simplifying the accident diagnosis and handling process and improving the efficiency and reliability of accident handling;

[0051] 5. The present invention can be applied to the digital operation design of the power plant in combination with the specific protection system design, primary / secondary / tertiary circuit system design, constant value design, and instrumentation design of the power plant, thereby improving the degree of automated operation of the power plant, reducing human influence during the operation of the power plant and the diagnosis and processing of accidents, improving the accident response speed, and reducing the burden of operating tasks on operators. BRIEF DESCRIPTION OF THE DRAWINGS

[0052] Figure 1 A schematic diagram of a diagnostic strategy for a loss of heat sink accident in a sodium-cooled fast reactor according to the present invention;

[0053] Figure 2 It is a schematic diagram of the treatment strategy for the loss of heat sink accident of the sodium-cooled fast reactor according to the present invention. DETAILED DESCRIPTION

[0054] In order to make the purpose, technical solution and advantages of the present invention more clearly understood, the present invention 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 invention and are not used to limit the present invention.

[0055] The invention provides a method for diagnosing and treating a loss of heat sink accident in a sodium-cooled fast reactor nuclear power plant, which includes a diagnostic method and strategy and a treatment method and strategy. The former is used for accident diagnosis, and the latter is used for accident treatment after diagnosis.

[0056] (1) Diagnostic methods and strategies

[0057] Sodium-cooled fast reactor nuclear power plants generally adopt a design with multiple loops and multiple steam generator modules. The heat generated by the core is transferred to the secondary loop, and then to the tertiary loop through the various modules of the steam generator, converting the feed water of the tertiary loop into steam and sending it to the steam turbine. The sign of the loss of heat sink of sodium-cooled fast reactors is directly manifested as an increase in the outlet sodium temperature of the steam generator module.

[0058] The diagnostic strategy for the loss of heat sink accident in a sodium-cooled fast reactor is: among n steam generators, at least m steam generators have an outlet sodium temperature T i The rise exceeds the threshold value θ, that is:

[0059]

[0060] Where n is the number of steam generator modules designed for an actual sodium-cooled fast reactor nuclear power plant.

[0061] The values ​​of m and θ are based on the following principles:

[0062] Principle A1: It should not be lower than the emergency shutdown requirement. That is, according to the protection logic design of the nuclear power plant, θ should not be higher than the threshold required for the protection shutdown and m should not be more than the number of modules whose steam generator outlet sodium temperature exceeds the required threshold required for the protection shutdown.

[0063] Principle A2: It should not be lower than the requirements for the commissioning of dedicated safety facilities. That is, according to the protection logic design of the nuclear power plant, θ should not be higher than the threshold required for the commissioning of dedicated safety facilities and m should not be more than the number of modules whose steam generator outlet sodium temperature exceeds the required threshold required for the commissioning of dedicated safety facilities.

[0064] Principle A3: If a nuclear power plant is designed with multiple loops, the T i The combination of , m and θ.

[0065] For example, in the form of loop grouping, the jth loop loses the heat sink:

[0066] ∑n j =n

[0067] Among them, T ji 、m j and θ j is the outlet sodium temperature, number and threshold of the j-th loop steam generator.

[0068] According to the actual power plant design, the m of each group j and θ j They can take different values ​​or the same value.

[0069] At this time, the definition of total loss of heat sinks should be based on the actual design of sodium-cooled fast reactor nuclear power plants and the most conservative requirements should be planned from the following principles:

[0070] Principle A31: All loops lose heat sinks;

[0071] Principle A32: The sum of the thermal conductivity of the remaining loops that have not lost their heat sinks does not meet the requirements for safe shutdown or the needs of handling the current accident conditions.

[0072] Principle A4: If the power plant is designed with multiple loops, m j and θ j The value of should not be lower than the requirement of the protection logic for the loss of heat sink in the jth loop.

[0073] Principle A5: If the power plant has a non-completely independent design between steam generator modules in the design of the secondary or tertiary system flow, the T should be analyzed and reasonably adjusted according to the impact of the loss of heat sink of a steam generator module caused by different primary causes on other modules. i , m and θ, and the values ​​of m and θ (including single-loop design and multi-loop design). For example, there is a design case with 8 steam generator modules in one loop, but due to process design factors, the two groups of 4 modules are independent of each other. In this case, one loop needs to be treated as two loops, and the diagnostic strategy formula for one loop is also split into two.

[0074] The impact of the above primary cause of the accident is analyzed according to the following method:

[0075] The initial causes of accidents should at least cover the accident types in the following table, and be supplemented and improved according to the actual power plant design requirements.

[0076] 1 Steam turbine shutdown 2 Steam generator feed water loss 3 Loss of external power 4 Failure of the steam pressure regulating device resulting in a reduction in steam flow 5 Condenser vacuum failure 6 Unintentional closure of steam isolation valve 7 Water pipe leakage 8 Steam pipe leak

[0077] For each primary cause accident, determine the specific impact according to the following table.

[0078]

[0079]

[0080] Principle A6: The values ​​of m and θ should not be too conservative. That is, the actual effect of the diagnostic strategy should only respond to accident conditions; it should not respond to abnormal operating events below accident conditions or transients that deviate from normal operation to avoid adverse effects on power plant operation. This principle needs to be completed in the accident strategy design of actual sodium-cooled fast reactor power plants by conducting verification and confirmation of the loss of heat sink diagnostic strategy.

[0081] (2) Treatment methods and strategies

[0082] The strategies for dealing with the loss of heat sink accident are divided into two parts: Strategy 1 and Strategy 2, which are applicable to the corresponding working conditions according to the following principles:

[0083] Principle B1: For single-loop design power plants, Strategy 2 should be used to deal with loss of heat sink accidents.

[0084] Principle B2: For power plants with multi-loop designs, the current operating condition should be determined based on diagnostic methods and strategies to determine whether it is a partial loss of heat sink or a complete loss of heat sink: If a partial loss of heat sink occurs, Strategy 1 should be used based on Strategy 2 and combined with the operating condition processing requirements; if a complete loss of heat sink occurs, Strategy 2 should be used.

[0085] Strategy 1 - Handle the incident according to the following process:

[0086] I) Check and confirm that the water level in the deaerator is within the normal range. If the water level is insufficient, add water to the normal range;

[0087] II) According to the current power plant operation requirements, determine the feedwater flow requirements, and then determine the number of feedwater pumps that need to be operated, and configure the corresponding feedwater pumps for operation;

[0088] III) Check that the pressure of the water supply system is within the normal range; otherwise, check whether the upstream system is operating normally;

[0089] IV) According to the power plant design, the feed water flow rate is controlled manually or automatically to restore the steam generator outlet temperature to the normal range.

[0090] Strategy 2 - Handle the incident according to the following process:

[0091] a) If the total loss of heat sink occurs, check and confirm that the dedicated safety facilities for emergency waste heat removal have been put into operation and that available waste heat removal paths have been established; otherwise, skip this step;

[0092] b) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include feedwater isolation and steam isolation, and check whether other normally open pipes connected to the steam generator have been isolated;

[0093] c) The loop where the accident is isolated should at least include water supply isolation and steam isolation;

[0094] d) If the power plant has a design in which the steam generator modules are not completely independent in the secondary or tertiary system flow, it has been checked that the connecting pipes between the steam generator modules where the accident occurred and the affected modules have been isolated.

[0095] Example

[0096] (1) Diagnostic strategy

[0097] Take a sodium-cooled fast reactor nuclear power plant with two loops and 16 steam generator modules as an example.

[0098] Diagnostic strategy:

[0099]

[0100] According to principle A1, based on the protection logic design of the nuclear power plant's protection shutdown, θ≤300℃, m≤3.

[0101] According to principle A2, based on the operational protection logic design of the nuclear power plant's dedicated safety facilities, θ≤300℃, m≤6.

[0102] According to principle A3: considering the two-loop design, the setting value of the loss of heat sink in each loop is the same as the protection trip setting value, and the form of the diagnostic strategy is

[0103] The first loop loses the heat sink

[0104]

[0105] The second loop loses the heat sink

[0106]

[0107] The definition of total loss of heat sinks is "all loops have lost heat sinks".

[0108] According to Principle A4, the power plant is a multi-loop design. The values ​​of m and θ in the form given by Principle A3 are not lower than the requirements of the protection logic for the loss of heat sink in any loop, which meets this principle.

[0109] According to Principle A5: In the design of the system flow of the secondary or tertiary circuits of the power plant, there is a design in which the steam generator modules are not completely independent. According to the primary cause accident table, the impact of the primary cause accident is analyzed, and after evaluation, it is considered that the form given in Principle A3 is applicable.

[0110]

[0111] According to Principle A6, the form given in Principle A3 is verified and confirmed, and there is no problem in the assessment.

[0112] According to the design of principles A1-A6, the diagnostic strategy is:

[0113] The first loop loses the heat sink:

[0114]

[0115] The second loop loses the heat sink:

[0116]

[0117] Total loss of heat sink: The first loop loses the heat sink, and the second loop loses the heat sink.

[0118] (2) Treatment strategy

[0119] The nuclear power plant is a 2-loop design, adopting the principle B2 design treatment strategy.

[0120] Strategies for dealing with total loss of heat sinks:

[0121] 1) If the total loss of heat sink occurs, check and confirm that the dedicated safety facilities have been put into operation and that available waste heat removal paths have been established;

[0122] 2) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include feedwater isolation and steam isolation. Check whether other normally open pipes connected to the steam generator have been isolated;

[0123] 3) The loop where the accident is isolated should at least include water supply isolation and steam isolation;

[0124] 4) In the design of the secondary or tertiary system flow of the power plant, there is an incompletely independent design between the steam generator modules, isolating the connecting pipelines of the steam generator modules where the accident is located and those affected.

[0125] Strategies for dealing with partial loss of heat sink:

[0126] 1) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include feedwater isolation and steam isolation. Check whether other normally open pipes connected to the steam generator have been isolated;

[0127] 2) The loop where the accident is isolated should at least include water supply isolation and steam isolation;

[0128] 3) In the design of the secondary or tertiary system flow of the power plant, there is an incompletely independent design between the steam generator modules, isolating the connecting pipelines of the steam generator modules where the accident is located and those affected;

[0129] 4) Check and confirm that the water level in the deaerator is within the normal range. If the water level is insufficient, add water to the normal range;

[0130] 5) According to the current power plant operation requirements, determine the water supply flow requirements, and then determine the number of water supply pumps that need to be operated, and configure the corresponding water supply pumps for operation;

[0131] 6) Check that the pressure of the water supply system is within the normal range; otherwise, check whether the upstream system is operating normally;

[0132] 7) According to the power plant design, use manual or automatic methods to control the feed water flow rate to restore the steam generator outlet temperature to the normal range.

[0133] It will be apparent to those skilled in the art that the invention is not limited to the details of the exemplary embodiments described above and that the invention can be implemented in other specific forms without departing from the spirit or essential features of the invention. Therefore, the embodiments should be considered exemplary and non-limiting in all respects, and the scope of the invention is defined by the appended claims rather than the foregoing description, and it is intended that all variations falling within the meaning and scope of the equivalent elements of the claims be included in the invention. Any reference numeral in a claim should not be considered as limiting the claim to which it relates.

[0134] In addition, it should be understood that although the present specification is described according to implementation modes, not every implementation mode contains only one independent technical solution. This description of the specification is only for the sake of clarity. Those skilled in the art should regard the specification as a whole. The technical solutions in each embodiment may also be appropriately combined to form other implementation modes that can be understood by those skilled in the art.

Claims

1. A method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident, comprising the following steps: (1) Diagnose the loss of heat sink accident of sodium-cooled fast reactor. The diagnostic strategy is as follows: among n steam generators, at least m steam generators have an outlet sodium temperature T i rises above the threshold value θ, in, n is the number of steam generator modules designed for an actual sodium-cooled fast reactor nuclear power plant; The values ​​of m and θ are based on the following principles: Principle A1: The threshold value θ should not be higher than the threshold value required for the protection shutdown, and m should not be more than the number of modules whose outlet sodium temperature of the steam generator exceeds the required threshold value required for the protection shutdown; Principle A2: The threshold value θ should not be higher than the threshold value required for commissioning of dedicated safety facilities, and m should not be more than the number of modules whose outlet sodium temperature of steam generators exceeds the required threshold value required for commissioning of dedicated safety facilities; Principle A3: If a sodium-cooled fast reactor nuclear power plant is designed with multiple loops, the T that can characterize the loss of heat sink in any loop should be considered. i The combination of , m and θ, the diagnostic strategy is expressed as: Among them, T ji 、m j and θ j is the outlet sodium temperature, number and threshold of the steam generator in the jth loop; n j is the number of steam generator modules in the j-th loop; At this time, the definition of total loss of heat sinks should be based on the actual design of sodium-cooled fast reactor nuclear power plants and the most conservative requirements should be planned from the following principles: 1) All loops lose heat sinks; 2) The sum of the thermal conductivity of the remaining loops that have not lost their heat sinks does not meet the requirements for safe shutdown or the needs of handling the current accident conditions; Principle A4: If a sodium-cooled fast reactor nuclear power plant is designed with multiple loops, m j and θ j The value of should not be lower than the requirement of the protection logic for the loss of heat sink in the jth loop; Principle A5: If the steam generator modules are not completely independent in the design of the secondary or tertiary system flow of a sodium-cooled fast reactor nuclear power plant, the T should be analyzed and reasonably adjusted according to the impact of the loss of heat sink of a steam generator module caused by different primary causes on other modules. i , the combination of m and θ, and the values ​​of m and θ; (2) Determine the loop type of the sodium-cooled fast reactor nuclear power plant and adopt the corresponding process to handle accidents for sodium-cooled fast reactor nuclear power plants with single-loop design or multi-loop design.

2. The method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident according to claim 1, characterized in that: The primary cause of the accident shall at least include: Steam turbine shutdown; Steam generator feed water loss; Loss of off-site power; Failure of steam pressure regulating devices resulting in reduced steam flow; Condenser vacuum failure; Unintentional closure of steam isolation valves; Water pipe leakage; Steam pipe leaks; For each primary cause accident, analyze its impact on the loop and the number of steam generator modules affected.

3. The method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident according to claim 1, characterized in that: The diagnostic strategy described in step (1) should only respond to accident conditions and should not respond to abnormal operating events below the accident conditions or transients that deviate from normal operation, so as to avoid adverse effects on power plant operation.

4. The method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident according to claim 1, characterized in that: In step (2), if a total loss of heat sink accident occurs in a sodium-cooled fast reactor nuclear power plant with a multi-loop design, or a loss of heat sink accident occurs in a sodium-cooled fast reactor nuclear power plant with a single-loop design, the accident handling process is as follows: (I) If the total loss of heat sink occurs, check and confirm that the dedicated safety facilities for emergency waste heat removal have been put into operation and that available waste heat removal pathways have been established; otherwise, skip this step; (II) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include isolation of feedwater and steam, and check whether other normally open pipes connected to the steam generator have been isolated; (III) Isolation of the loop where the accident occurred shall include at least water supply isolation and steam isolation; (IV) If the power plant has a design in which the steam generator modules are not completely independent in the secondary or tertiary system flow, it has been checked that the connecting pipes between the steam generator modules where the accident occurred and the affected modules have been isolated.

5. The method for diagnosing and treating a sodium-cooled fast reactor loss of heat sink accident according to claim 1, characterized in that: In step (2), for a partial loss of heat sink accident in a sodium-cooled fast reactor nuclear power plant with a multi-loop design, the accident handling process is as follows: (a) Check and confirm that the steam generator module where the accident occurred and the affected module have been isolated, which should at least include isolation of feedwater and steam, and check whether other normally open pipes connected to the steam generator have been isolated; (b) Isolation of the loop where the accident occurs shall include at least water supply isolation and steam isolation; (c) If the power plant has a secondary or tertiary system flow design in which the steam generator modules are not completely independent, it has been checked that the connecting pipes between the steam generator module where the accident occurred and the affected steam generator module have been isolated. (d) Check and confirm that the water level in the deaerator is within the normal range. If the water level is insufficient, add water to the normal range; (e) Determine the feedwater flow demand based on the current power plant operation requirements, and then determine the number of feedwater pumps that need to be operated, and configure the corresponding feedwater pumps for operation; (f) Check that the pressure of the water supply system is within the normal range; if not, check whether the upstream system is operating normally; (g) According to the power plant design, the feed water flow rate shall be controlled manually or automatically to restore the steam generator outlet temperature to the normal range.

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