Reactor core neutron detector sleeve fracture accident numerical analysis method

Abstract

The invention discloses a numerical analysis method for a reactor core neutron detector casing pipe fracture accident. The method mainly comprises the following steps: 1, dividing a casing pipe into different areas according to the heat exchange condition of the outer wall surface of the casing pipe, and respectively modeling; 2, the models of all the parts are connected to form a neutron detector system casing fracture accident analysis model, and meanwhile geometric parameters, material physical properties, outer wall surface heat exchange conditions and thermal hydraulic initial values of all the parts are set; 3, calculating steady-state working condition operation parameters; 4, correcting the initial transient calculation parameters of the casing into the calculated steady-state parameters; 5, setting an accident action sequence, compiling an accident logic trigger signal, and setting conditions such as the area, the position and the energy loss coefficient of a crevasse; and 6, calculating the transient working condition of the casing fracture accident to obtain the temperature field change of the downstream flow channel, and carrying out sensitivity analysis on the position and the area of the fracture to obtain the 95% confidence interval of the temperature of the key position. According to the method, simulation data can be quickly generated, and sensitivity analysis and calculation are facilitated.

Description

Technical Field

[0001] The invention relates to the field of casing rupture accident analysis of a pressurized water reactor nuclear power plant, and in particular to a numerical analysis method for casing rupture accidents of a core neutron detector. Background Art

[0002] The neutron detector sleeve is an important part of the pressurized water reactor core neutron flux measurement system. Its structure is an L-shaped ring tube, one end of which directly extends into the pressurized water reactor fuel assembly, and the other end is directly connected to the outside atmosphere. There are three sets of isolation valves downstream of the sleeve, namely: manual valve, ball check valve and electric valve. Generally speaking, the manual valve and electric valve are in a normally closed state. During the operation of the reactor, the operator can open the downstream isolation valve and extend the measuring hose with a mobile micro-fission chamber on the top along the sleeve into the reactor fuel assembly.

[0003] During normal operation of a pressurized water reactor nuclear power plant, the coolant in the core flow channel may vibrate laterally due to disturbances caused by fixed equipment such as grids. This type of vibration will cause wear of the neutron detector's casing and surrounding grids in the reactor. After long-term wear, the casing wall may be worn through, causing the primary coolant to flow into it. Generally, during actual operation, once significant wear of the casing is detected, the casing will be blocked to prevent the primary coolant from leaking into the casing. After a casing rupture accident, the downstream valve assumes the role of isolating the primary coolant from the external atmospheric environment. However, since the internal sealing material of the valve is made of high polymer materials such as polytetrafluoroethylene, when the temperature field inside the casing changes beyond the tolerance threshold of non-metallic materials, the valve will lose its sealing performance, causing the primary coolant to leak out.

[0004] At present, in-depth research has been conducted at home and abroad on the distribution law of casing wear, and the casing wear mechanism, wear causes and distribution law have been analyzed, and optimization suggestions have been put forward for the casing wear treatment criteria. However, in the actual operation monitoring process of the reactor, there is no means to effectively check the casing wear during operation except for stopping the reactor for inspection. Therefore, based on the idea of ​​conservatism, it is necessary to numerically model the casing of the pressurized water reactor and simulate the breach accident. However, there is not much progress in the numerical simulation analysis work carried out in this field at home and abroad. Most of them focus on the mechanism of flow-induced vibration and the analysis of the easy-to-wear area of ​​the casing. The analysis of the temperature change of the flow field inside the casing under accident conditions and the safety assessment of the isolation valve downstream of the casing still need to be further deepened.

[0005] The present invention proposes a method for developing a pressurized water reactor neutron detector casing rupture accident analysis model based on optimal estimation software. The method can establish a two-fluid thermal hydraulic accident analysis model for the entire casing including a flow model, a heat exchange model, and an accident logic trigger signal. Summary of the invention

[0006] In order to numerically simulate the development of the accident after the casing rupture, theoretically calculate the temperature change of the external flow field of the downstream valve after the casing rupture, and verify whether the downstream valve will fail under the accident condition, the present invention provides a numerical analysis method for the casing rupture accident of the core neutron detector by analyzing the geometric structure, heat exchange boundary and rupture accident process of the casing. The present invention is of great significance to the maintenance safety of the core measurement system of the pressurized water reactor nuclear power plant during the daily operation.

[0007] In order to achieve the above object, the present invention is implemented by adopting the following technical solutions:

[0008] A numerical analysis method for a core neutron detector casing rupture accident, the steps are as follows:

[0009] Step 1: According to the different heat exchange conditions of the outer wall of the neutron detector sleeve of the pressurized water reactor core, the neutron detector sleeve is divided into three parts and modeled separately. The three parts are: the inner part of the pressure vessel, the front part of the pressure vessel outer sealing assembly, and the rear part of the pressure vessel outer sealing assembly;

[0010] The casing inside the pressure vessel is simulated by pipe components, and the pipe wall is simulated by thermal components; the external core flow channel is directly given according to the average channel assumption: the entire core is regarded as a coolant channel, the length of the channel is equal to the actual reactor height, and the flow area is equal to the actual coolant flow channel area in the PWR core at this height. The PWR pressure boundary and mass flow rate are given by using time-dependent control bodies and time-dependent nozzles, and a point pile model with six groups of delayed neutrons is used to simulate the PWR fuel rods in the form of thermal components to give the correct power distribution in the core flow channel;

[0011] The air inside the casing of the front part of the pressure vessel outer sealing assembly conducts natural convection heat exchange with the pipe wall, the primary circuit working fluid conducts natural convection heat exchange with the outer wall of the casing and the inner wall of the guide ring pipe, the outer wall of the guide ring pipe in the penetration is in contact with the concrete shielding wall, and the outer wall of the guide ring pipe in the non-penetration is in natural convection heat exchange with the outside air; this part of the casing is simulated by pipeline components, the pipe wall is simulated by thermal components, the guide ring pipe is simulated by annular pipeline components, the guide ring pipe wall is simulated by thermal components, and the atmospheric environment in the containment is simulated by pipeline components; the concrete shielding wall is simulated by the pipe wall thermal component of the guide ring pipe in the penetration, and the different material properties and thicknesses of the pipe wall and concrete are distinguished by the radial spacing and composition of the thermal component network;

[0012] The casing of the rear part of the outer sealing assembly of the pressure vessel is simulated by a pipeline component, the pipe wall is simulated by a thermal component, and the atmospheric environment is also simulated by a pipeline component. A ball valve is provided in the middle of this part of the casing, and an electric valve is provided at the end.

[0013] Step 2: After the establishment of each part model, connect the parts together to form the neutron detector system casing rupture accident analysis model. At the same time, set the geometric parameters, material properties, outer wall heat transfer conditions and thermal hydraulic initial values ​​of each part to provide initial conditions and boundary conditions for the development of accident conditions.

[0014] The working condition solution is based on the thermal hydraulic calculation method of nuclear reactors. Its field equations consist of three conservation equations of two phases, and eight primary dependent variables need to be solved: cavitation fraction α g , the velocity v of the vapor and liquid phases g and v f , the specific internal energy of vapor phase and liquid phase U g and U f , pressure P, boron density ρ b and non-condensable gas ratio X n ;

[0015] It is necessary to set the control body flow area A and control body length x of the hydraulic components in the neutron detector system casing rupture accident analysis model as geometric parameters, and the thermal conductivity λ and heat capacity C of the thermal components need to be given. P , and initial calculation parameters such as liquidus temperature T f , gas phase temperature T g , pressure P, liquid phase velocity v f , gas phase velocity v g , wall temperature T w ;

[0016] Step 3: Based on the established neutron detector system casing rupture accident analysis model, steady-state operation is performed, and the operating parameters of the neutron detector system casing rupture accident analysis model when it reaches the steady-state condition are calculated. The accuracy of the neutron detector system casing rupture accident analysis model is verified by comparing it with the actual design operating parameters;

[0017] Step 4: Use the steady-state operation results of the previous step as the initial conditions for the next transient calculation. During the transient calculation process, it is necessary to trigger the accident on the basis of keeping the model parameters stable, which can significantly reduce the time required for the model to reach a steady state.

[0018] Step 5: Use the trigger valve component to simulate the breach. Set multiple trigger valves at the connection with the grid according to the location where the casing is prone to wear. One end of the trigger valve is connected to the casing and the other end is connected to the core coolant channel to simulate the breach that may occur at different locations. At the same time, set the breach at the highest coolant temperature at the core outlet for conservative estimation. Add a congestion model to the trigger valve component and give the breach flow energy loss coefficient.

[0019] In addition, it is also necessary to set the action sequence of the PWR in the accident process, use the trigger signal to control the trigger valve to open, and after the PWR reaches a steady state for a period of time, the trigger valve receives the trigger signal and opens, indicating that a casing rupture accident has occurred; other non-rupture valves are open under normal conditions, and corresponding actions are performed after the accident occurs. When the water level of the control body at the external leakage probe position is greater than 0, the electric valve at the end of the casing is immediately closed, and after a period of time, the manual isolation valve between the concrete wall and the sealing component begins to be closed. The trigger logic of the ball valve behind the sealing component is to automatically close when the water level of the upstream control body is higher than the threshold, or the corresponding valve is not triggered to simulate the process of superimposed valve failure under accident conditions;

[0020] Step 6: Based on the established neutron detector system casing rupture accident analysis model, perform transient numerical simulation calculations of the accident. After the calculation reaches a steady state, simulate the occurrence of the accident by opening the rupture valve, and perform a sensitivity analysis on the rupture position and rupture area.

[0021] Compared with the prior art, the method for developing a pressurized water reactor neutron detector casing rupture accident analysis model based on the best estimation software of the present invention has the following advantages:

[0022] 1. Taking into account the different heat exchange conditions of the casing at different positions, the temperature distribution is more accurate. The modeling area is divided according to the different heat exchange conditions of the outer wall of the casing, which effectively reduces the difficulty of modeling. In addition, the influence of the outer wall of the core part on the temperature field of the downstream flow channel due to the coolant heat exchange in the case of casing rupture accident is considered;

[0023] 2. The average channel concept was used to establish a model of heat exchange between the core flow channel and the casing in the primary loop, which better reflects the core power distribution and improves the calculation efficiency;

[0024] 3. Set up breach models with different positions and sizes to conduct sensitivity analysis on breach position and breach area to improve the credibility of the results;

[0025] In summary, this method can analyze casing rupture accidents more comprehensively, effectively and efficiently. BRIEF DESCRIPTION OF THE DRAWINGS

[0026] Figure 1 The present invention is a flow chart of a method for developing a pressurized water reactor neutron detector casing rupture accident analysis model based on optimal estimation software.

[0027] Figure 2 It is a schematic diagram of a pressurized water reactor neutron detector casing rupture accident analysis model based on the best estimation software of the present invention. DETAILED DESCRIPTION

[0028] The present invention is further described in detail below in conjunction with the accompanying drawings and specific embodiments.

[0029] The present invention provides a numerical analysis method for a core neutron detector casing rupture accident, the process is as follows: Figure 1 As shown, the steps are as follows:

[0030] Step 1: According to the different heat exchange conditions of the outer wall of the neutron detector casing of the pressurized water reactor core, the neutron detector casing is divided into three parts and modeled separately. The three parts are: the inner part of the pressure vessel, the front part of the pressure vessel outer sealing assembly, and the rear part of the pressure vessel outer sealing assembly.

[0031] The outer wall of the casing inside the pressure vessel is in direct contact with the primary working fluid, which is equivalent to the primary pressure boundary. The tube wall is directly heated by the forced convection of the primary high-temperature working fluid on the outside. This part of the casing is simulated by a pipe component, and the tube wall is simulated by a thermal component. The external core flow channel is directly given according to the average channel assumption: the entire core is regarded as a coolant channel, the length of which is equal to the actual reactor height, and the flow area is equal to the actual coolant flow channel area in the PWR core at this height. The PWR pressure boundary and mass flow rate are given using a time-dependent control body and a time-dependent nozzle, and a point pile model with six groups of delayed neutrons is used to simulate the PWR fuel rods in the form of thermal components to give the correct power distribution in the core flow channel.

[0032] The outer wall of the casing in the front part of the pressure vessel outer sealing assembly is surrounded by a guide ring tube, the inside of the guide ring tube is directly connected to the primary circuit, and the outer wall directly exchanges heat with the air environment inside the shielding wall by natural convection. Since the inside of the guide ring tube is a dead water area, the coolant temperature therein will be significantly different from the temperature inside the pressure vessel. At the same time, this part of the casing also has a special heat exchange structure: the guide ring tube and the casing in the penetration. In this special heat exchange structure, if the working fluid inside the casing wants to transfer heat to the external atmospheric environment, the thermal resistances that need to be passed are: the casing wall, the working fluid of the primary circuit in the guide ring tube, the guide ring tube wall and the concrete shielding wall. In short, the air inside the casing in the front part of the pressure vessel outer sealing assembly exchanges heat with the tube wall by natural convection, the working fluid of the primary circuit exchanges heat with the outer wall of the casing and the inner wall of the guide ring tube by natural convection, the outer wall of the guide ring tube in the penetration is in contact with the concrete shielding wall for heat conduction, and the outer wall of the guide ring tube in the non-penetration is in natural convection heat exchange with the outside air. This part of the casing is simulated by pipe components, the pipe wall is simulated by thermal components, the guide ring is simulated by annular pipe components, the guide ring pipe wall is simulated by thermal components, and the atmosphere in the containment is simulated by pipe components. The concrete shielding wall is simulated by the pipe wall thermal component of the guide ring pipe in the penetration, and the different material properties and thicknesses of the pipe wall and concrete are distinguished by the radial spacing and composition of the thermal component network.

[0033] The outer wall of the casing behind the outer sealing assembly of the pressure vessel directly exchanges heat with the atmosphere by natural convection, and the air inside it exchanges heat with the inner side of the pipe wall by natural convection. This part of the casing is simulated by pipeline components, the pipe wall is simulated by thermal components, and the atmospheric environment is also simulated by pipeline components. A ball valve is set in the middle of this part of the casing, and an electric valve is set at the end.

[0034] Step 2: After each part of the model is built, connect the parts together to form the following Figure 2 The neutron detector system casing rupture accident analysis model shown in the figure sets the geometric parameters, material properties, outer wall heat transfer conditions and thermal hydraulic initial values ​​of each part to provide initial conditions and boundary conditions for the development of accident conditions.

[0035] The working condition solution is based on the thermal hydraulic calculation method of nuclear reactors. Its field equations are mainly composed of three conservation equations (mass, momentum and energy conservation equations) of two phases (liquid phase and gas phase). Eight primary dependent variables need to be solved: cavitation fraction α g , the velocity v of the vapor and liquid phases g and v f , the specific internal energy of vapor phase and liquid phase U g and U f , pressure P, boron density ρ b and non-condensable gas ratio X n . Its internal field equations are as follows:

[0036] (1) The mass conservation equation is as follows:

[0037] The gas phase is:

[0038]

[0039] The liquid phase is:

[0040]

[0041] Where:

[0042] A——circulation area / m 2 ;

[0043] ρ g ——Gas phase density / kg·m -3 ;

[0044] ρ f ——Liquid phase density / kg·m -3 ;

[0045] Γ g ——Gas phase generation rate / kg·m -3 ;

[0046] Γ f——Liquid phase generation rate / kg·m -3 ;

[0047] α f ——Liquid phase volume fraction;

[0048] t——running time / s;

[0049] x——control body length / m.

[0050] (2) The momentum conservation equation is as follows:

[0051] The gas phase is:

[0052]

[0053] The liquid phase is:

[0054]

[0055] Where:

[0056] B x ——Volume force in the x direction / N·kg -1 ;

[0057] v gi ——Gas phase reference velocity / m·s -1 ;

[0058] v fi ——Liquid phase reference velocity / m·s -1 ;

[0059] ρ m ——Mixed density / kg·m -3 ;

[0060] C——true mass coefficient;

[0061] FIG——Gas phase interface resistance coefficient / s -1 ;

[0062] FIF——Liquid interface resistance coefficient / s -1 ;

[0063] FWG——Gas phase wall resistance coefficient / s -1 ;

[0064] FWF——Liquid phase wall resistance coefficient / s -1 . (3) The energy conservation equation is as follows:

[0065] The gas phase is:

[0066]

[0067] The liquid phase is:

[0068]

[0069] Where:

[0070] ——Compression work / J;

[0071] ——Expansion work / J;

[0072] ——rate of change of internal energy related to the liquid phase;

[0073] ——rate of change of internal energy related to the vapor phase;

[0074] Q wf ——Heat exchange rate of liquid phase on unit volume of wall surface / W·m -3 ;

[0075] Q wg ——Heat exchange rate of vapor phase on unit volume of wall surface / W·m -3 ;

[0076] Q if ——Heat exchange rate on liquid interface / W·m -3 ;

[0077] Q ig ——Heat exchange rate on the vapor phase interface / W·m -3 ;

[0078] ——Energy change caused by liquid mass migration in the interface / J;

[0079] ——Energy change caused by vapor mass migration in the interface / J;

[0080] Γ wf h f ——Energy change caused by liquid mass migration in the wall / J;

[0081] Γ wg h g ——Energy change caused by vapor mass migration within the wall / J;

[0082] DISS g ——Vapor phase energy dissipation function / W·m -3 ;

[0083] DISS f ——Liquid phase energy dissipation function / W·m -3 , which is expressed as follows:

[0084]

[0085]

[0086] Therefore, it is necessary to set the control body flow area A and control body length x of the hydraulic components in the neutron detector system casing rupture accident analysis model as geometric parameters, and it is also necessary to give the thermal conductivity λ and heat capacity C of the thermal components. P , and initial calculation parameters such as liquidus temperature T f , gas phase temperature T g , pressure P, liquid phase velocity v f , gas phase velocity v g , wall temperature T w .

[0087] Step 3: Based on the established neutron detector system casing rupture accident analysis model, steady-state operation is performed, and the operating parameters of the neutron detector system casing rupture accident analysis model when it reaches the steady-state condition are calculated. The accuracy of the neutron detector system casing rupture accident analysis model is verified by comparing it with the actual design operating parameters.

[0088] Step 4: Use the steady-state operating results of the previous step as the initial conditions for the next transient calculation. During the transient calculation process, it is necessary to trigger the accident on the basis of keeping the model parameters stable. This step can significantly reduce the time required for the model to reach a steady state.

[0089] Step 5: Use the trigger valve component to simulate the breach. Set multiple trigger valves at the connection with the grid according to the location where the casing is prone to wear. One end of the trigger valve is connected to the casing and the other end is connected to the core coolant channel to simulate the breach that may occur at different locations. At the same time, set the breach at the highest coolant temperature at the core outlet for conservative estimation. Add a congestion model to the trigger valve component and give the breach flow energy loss coefficient.

[0090] In addition, it is also necessary to set the action sequence of the pressurized water reactor in the accident process, use the trigger signal to control the trigger valve to open, and the trigger valve receives the trigger signal and opens after the system reaches a steady state for a period of time, indicating that a casing rupture accident has occurred. Other non-rupture valves are open under normal conditions, and corresponding actions are performed after the accident occurs. When the water level of the control body at the external leakage probe position is greater than 0, the electric valve at the end of the casing is immediately closed. After a period of time, the manual isolation valve between the concrete wall and the sealing component begins to close. The trigger logic of the ball valve behind the sealing component is to automatically close when the water level of the upstream control body is higher than the threshold. The process of superimposed valve failure under accident conditions can also be simulated by not triggering the corresponding valve.

[0091] Step 6: Based on the established neutron detector system casing rupture accident analysis model, the accident transient numerical simulation calculation is performed. After the calculation reaches the steady state, the occurrence of the accident is simulated by opening the trigger valve, and the sensitivity analysis of the rupture position and rupture area is performed. Taking the rupture area as an example, simple random sampling is performed within the value range of the rupture area, and 100 groups of rupture areas are selected to set the initial parameters. After the calculation is completed, the peak temperature corresponding to 100 groups of rupture areas is obtained. The calculated values ​​are statistically analyzed to obtain the 95% confidence interval and the best estimate of the reactor peak temperature after the core neutron detector casing rupture accident applicable to the geometric parameters of the target power plant. Among them, the 95% confidence interval result expresses that the initial conditions of the accident calculation change arbitrarily under actual circumstances, and the peak temperature of the reactor after the accident occurs has a 95% confidence level falling within this interval. The best estimate expresses the maximum probability value of the reactor peak temperature caused by the arbitrarily changing initial conditions of the accident calculation. The two statistical results obtained can be used to quantify the uncertainty in the accident calculation and improve the credibility of the results.

[0092] In summary, the casing model of the pressurized water reactor neutron detector system was initially established through step 1; the operating parameters of the model were set through step 2; the parameters of the casing during steady-state operation were determined by repeating steps 3 and 4, and used as the initial parameters for transient analysis; the breach model was established through step 5, and the accident logic signal triggering conditions were set; and the downstream temperature field changes of the casing accident conditions under different breach positions and areas were finally calculated through step 6. The failure probability of the non-metallic sealing components inside the downstream valve can be evaluated through the numerical simulation results.

Claims

1. A numerical analysis method for a core neutron detector casing rupture accident, characterized in that: Here are the steps: Step 1: According to the different heat exchange conditions of the outer wall of the neutron detector sleeve of the pressurized water reactor core, the neutron detector sleeve is divided into three parts and modeled separately. The three parts are: the inner part of the pressure vessel, the front part of the pressure vessel outer sealing assembly, and the rear part of the pressure vessel outer sealing assembly; The casing inside the pressure vessel is simulated by pipe components, and the pipe wall is simulated by thermal components; the external core flow channel is directly given according to the average channel assumption: the entire core is regarded as a coolant channel, the length of the channel is equal to the actual reactor height, and the flow area is equal to the actual coolant flow channel area in the PWR core at this height. The PWR pressure boundary and mass flow rate are given by using time-dependent control bodies and time-dependent nozzles, and a point pile model with six groups of delayed neutrons is used to simulate the PWR fuel rods in the form of thermal components to give the correct power distribution in the core flow channel; The air inside the casing of the front part of the pressure vessel outer sealing assembly conducts natural convection heat exchange with the pipe wall, the primary circuit working fluid conducts natural convection heat exchange with the outer wall of the casing and the inner wall of the guide ring pipe, the outer wall of the guide ring pipe in the penetration is in contact with the concrete shielding wall, and the outer wall of the guide ring pipe in the non-penetration is in natural convection heat exchange with the outside air; this part of the casing is simulated by pipeline components, the pipe wall is simulated by thermal components, the guide ring pipe is simulated by annular pipeline components, the guide ring pipe wall is simulated by thermal components, and the atmospheric environment in the containment is simulated by pipeline components; the concrete shielding wall is simulated by the pipe wall thermal component of the guide ring pipe in the penetration, and the different material properties and thicknesses of the pipe wall and concrete are distinguished by the radial spacing and composition of the thermal component network; The casing of the rear part of the outer sealing assembly of the pressure vessel is simulated by a pipeline component, the pipe wall is simulated by a thermal component, and the atmospheric environment is also simulated by a pipeline component. A ball valve is provided in the middle of this part of the casing, and an electric valve is provided at the end. Step 2: After the establishment of each part model, connect the parts together to form the neutron detector system casing rupture accident analysis model. At the same time, set the geometric parameters, material properties, outer wall heat transfer conditions and thermal hydraulic initial values ​​of each part to provide initial conditions and boundary conditions for the development of accident conditions. The working condition solution is based on the thermal hydraulic calculation method of nuclear reactors. Its field equations consist of three conservation equations of two phases, and eight primary dependent variables need to be solved: cavitation fraction α g , the velocity v of the vapor and liquid phases g and v f , the specific internal energy of vapor phase and liquid phase U g and U f , pressure P, boron density ρ b and non-condensable gas ratio X n ; It is necessary to set the control body flow area A and control body length x of the hydraulic components in the neutron detector system casing rupture accident analysis model as geometric parameters, and the thermal conductivity λ and heat capacity C of the thermal components need to be given. P , and initial calculation parameters such as liquidus temperature T f , gas phase temperature T g , pressure P, liquid phase velocity v f , gas phase velocity v g , wall temperature T w ; Step 3: Based on the established neutron detector system casing rupture accident analysis model, steady-state operation is performed, and the operating parameters of the neutron detector system casing rupture accident analysis model when it reaches the steady-state condition are calculated. The accuracy of the neutron detector system casing rupture accident analysis model is verified by comparing it with the actual design operating parameters; Step 4: Use the steady-state operation results of the previous step as the initial conditions for the next transient calculation. During the transient calculation process, it is necessary to trigger the accident on the basis of keeping the model parameters stable, which can significantly reduce the time required for the model to reach a steady state. Step 5: Use the trigger valve component to simulate the breach. Set multiple trigger valves at the connection with the grid according to the location where the casing is prone to wear. One end of the trigger valve is connected to the casing and the other end is connected to the core coolant channel to simulate the breach that may occur at different locations. At the same time, set the breach at the highest coolant temperature at the core outlet for conservative estimation. Add a congestion model to the trigger valve component and give the breach flow energy loss coefficient. In addition, it is also necessary to set the action sequence of the PWR in the accident process, use the trigger signal to control the trigger valve to open, and after the PWR reaches a steady state for a period of time, the trigger valve receives the trigger signal and opens, indicating that a casing rupture accident has occurred; other non-rupture valves are open under normal conditions, and corresponding actions are performed after the accident occurs. When the water level of the control body at the external leakage probe position is greater than 0, the electric valve at the end of the casing is immediately closed, and after a period of time, the manual isolation valve between the concrete wall and the sealing component begins to be closed. The trigger logic of the ball valve behind the sealing component is to automatically close when the water level of the upstream control body is higher than the threshold, or the corresponding valve is not triggered to simulate the process of superimposed valve failure under accident conditions; Step 6: Based on the established neutron detector system casing rupture accident analysis model, perform transient numerical simulation calculations of the accident. After the calculation reaches a steady state, simulate the occurrence of the accident by opening the rupture valve, and perform a sensitivity analysis on the rupture position and rupture area.

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