A method of coupling a vver unit system program with a neutron kinetics program

Abstract

The present application relates to the field of nuclear power unit accident analysis, and particularly relates to a coupling method of VVER unit system program and neutron dynamics program. The method is as follows: a reactor system program is compiled into a dynamic link library by adding an interface function in a three-dimensional neutron dynamics program of a reactor core, and the dynamic link library is accessed and called by the reactor system program in a plug-in mode; a neutron physics model of a specific loading scheme of a VVER type reactor is established; a simulation model of an accident analysis system of the VVER type pressurized water reactor is established; control logic of each main system and device of the primary loop and the secondary loop of the VVER type reactor is established; control logic and conservative condition settings for specific accidents are established; coupling calculation of the reactor system program and the three-dimensional neutron dynamics of the reactor core is performed; and key safety parameters of the VVER type reactor in specific accidents are obtained based on the results of the coupling calculation. The present application can simulate the changes of physical thermal parameters and the action responses of main systems in the primary loop and the secondary loop of the reactor after accidents, and is convenient for efficient and reliable accident analysis of the VVER type nuclear power unit.

Description

Technical Field

[0001] This invention relates to the field of nuclear power unit accident analysis, and in particular to a coupling method between VVER unit system program and neutron dynamics program. Background Technology

[0002] Commercial pressurized water reactors require refueling every 12 or 18 months. During refueling, irradiated fuel assemblies are removed from the reactor core and replaced with unirradiated new fuel assemblies. Irradiated fuel assemblies that have reached a certain burnup level and the new fuel assemblies are rearranged within the core to form a new core loading scheme. The safety of this new core loading scheme needs to be confirmed through nuclear design calculations and evaluation of critical neutronics parameters. When a critical neutronics parameter exceeds the limits in the final safety analysis report of the nuclear power plant, a reanalysis of the accident related to that critical neutronics parameter is required to verify that the consequences of the accident still meet the relevant acceptance criteria.

[0003] Therefore, nuclear power plant accident analysis is a crucial part of the nuclear energy field, involving a comprehensive and in-depth assessment of various anomalies and potential accidents that nuclear power plants may encounter during operation. Especially in the reactor core, the core area of ​​a nuclear reactor, physical phenomena such as flow, heat transfer, neutronics, and mechanical actions interact to form an extremely complex system. The simulation and analysis of these phenomena and processes require the collaborative work of multiple specialized procedures to ensure the accuracy and reliability of the analysis.

[0004] In nuclear power unit accident analysis, the coupling relationship between physical and thermal phenomena within the reactor core is particularly close, often exhibiting dramatic dynamic changes. For example, in control rod ejection accident analysis, as the control rods rapidly eject from the core after the drive mechanism fails, it first causes rapid changes in the total core power and power spatial distribution in terms of neutron physics. These changes in power and its distribution then cause changes in the coolant temperature and density fields within the core in terms of thermal and hydraulic aspects. These changes, in turn, affect the temporal dynamic behavior of the core power and its distribution through feedback. Therefore, in reactivity-related accident analysis, physical-thermal-engineering coupled calculations are particularly important, requiring coupling programs capable of accurately simulating and predicting these complex physical-thermal-engineering phenomena.

[0005] Furthermore, since the time response of neutron physics parameters is much faster than that of thermal hydraulic parameters, the coupling method between the neutron physics program and the thermal hydraulic program is also very important in order to improve the computational efficiency of the coupled program.

[0006] In the development and application of nuclear power unit accident analysis tools, there are relatively more tools available for prismatic pressurized water reactors (PWRs). However, for VVER-type PWRs, tools that directly couple system programs with core 3D neutron dynamics programs to complete specific accident analyses remain relatively scarce. Therefore, developing a physical-thermal-engineering coupling program capable of specifically simulating transient accidents in VVER-type units to enhance the assessment of the physical-thermal-engineering characteristics of VVER-type PWRs and their response capabilities under accident conditions is of great significance for the safe operation of VVER-type PWRs. Summary of the Invention

[0007] The technical problem to be solved by the present invention is to provide a coupling method between the VVER unit system program and the neutron dynamics program, which can realize the simulation of the changes in physical and thermal parameters of the primary and secondary loop systems and the action response of the main systems after an accident, so as to facilitate efficient and reliable accident analysis of VVER type nuclear power units.

[0008] This invention provides a method for coupling VVER reactor system programs with neutron dynamics programs, comprising the following steps:

[0009] Step 1: After adding interface functions to the core three-dimensional neutron dynamics program, compile it into a dynamic link library and allow it to be accessed and called by the reactor system program as a plug-in;

[0010] Step 2: Establish a neutron physics model for a specific loading scheme of a VVER-type reactor;

[0011] Step 3: Establish a simulation model of the accident analysis system for a VVER-type pressurized water reactor in the reactor system program;

[0012] Step 4: Based on the simulation model of the accident analysis system of the VVER type pressurized water reactor, establish the control logic of each major system and equipment in the primary and secondary loops of the VVER type reactor;

[0013] Step 5: Establish control logic and conservative condition settings for specific accidents;

[0014] Step 6: Perform coupled calculations of the reactor system program and the three-dimensional neutron dynamics of the reactor core;

[0015] Specifically, it includes:

[0016] Step 6-1: Read the reactor model input file and initialize the reactor system program calculation; the model includes a neutron physics model, an accident analysis system simulation model, and control logic;

[0017] Step 6-2: Initialization calculation of the three-dimensional neutron dynamics program for the reactor core;

[0018] Step 6-3: Perform coupled calculations under steady-state conditions. The core state, which matches the thermo-hydraulic parameters and neutronics parameters of each node of the reactor system, is obtained through iterative calculations of the reactor system program and the core three-dimensional neutron dynamics program.

[0019] Step 6-4: Perform coupling calculations under transient conditions. Within the initial time step, the reactor system program calculates the thermal-hydraulic parameters of each loop and node in the reactor system based on the steady-state power distribution fed back by the core three-dimensional neutron dynamics program.

[0020] The thermal-hydraulic parameters of each node are passed to the core three-dimensional neutron dynamics program through interface functions.

[0021] The core three-dimensional neutron dynamics program calculates the core three-dimensional power distribution based on the obtained thermo-hydraulic parameters of each node;

[0022] After obtaining the three-dimensional core power distribution, it is passed to the reactor system program, which determines the new time step based on the parameter status, and then calculates the new parameters based on the three-dimensional core power distribution.

[0023] Repeat the above transient calculation process until the accumulated time reaches the set limit;

[0024] Step 7: Based on the results of the coupled calculations, obtain the key safety parameters of the VVER reactor in a specific accident.

[0025] In one specific embodiment of the present invention, in step 1,

[0026] In the compilation of the three-dimensional neutron dynamics dynamic link library, interface functions are pre-built and compiled together with the code of the core three-dimensional neutron dynamics program into a dynamic link library that can be recognized by the reactor system program, so that the core three-dimensional neutron dynamics program can be accessed and called by the reactor system program in the form of a plug-in;

[0027] The interface function obtains the variable addresses in the reactor system program and reads the required data by using a hash chart, thereby achieving efficient and reliable data transfer between the reactor system program and the core three-dimensional neutron dynamics program.

[0028] In one specific embodiment of the present invention, the interface function includes an initialization function, a driving function, and a feedback function;

[0029] The initialization function is used to initialize the core three-dimensional neutron dynamics program and interface parameters, and to provide the node information of coolant and fuel in the core part of the reactor system model through mapping relationships;

[0030] The driving parameters are used to transmit the thermo-hydraulic parameters of each node along the axial direction of the core three-dimensional neutron dynamics program, and drive the neutron dynamics program to complete the power distribution calculation.

[0031] The feedback function is used to map the three-dimensional power distribution calculated by the core three-dimensional neutron dynamics program to each node of the core in the reactor system program, and to feed the data back to the reactor system program.

[0032] In one specific embodiment of the present invention, the thermal-hydraulic parameters of each node include the temperature and density distribution of the coolant, the fuel temperature, the boron concentration, and the position of the control rod.

[0033] In a specific embodiment of the present invention, step 2 specifically includes:

[0034] Step 2-1: Complete the arrangement of the core loading scheme for the VVER type reactor specific fuel cycle in the core three-dimensional neutronics program;

[0035] Step 2-2: Obtain the pre-generated cross-sectional information of various fuel assemblies based on the core loading scheme;

[0036] Steps 2-3: Use the core three-dimensional neutron dynamics program to perform three-dimensional steady-state neutron diffusion and burnup calculations to obtain a neutron physics model.

[0037] In one specific embodiment of the present invention, in step 4,

[0038] The process signals that define the thermal parameters of the thermal fluid object at a specific location and the overall parameters of the core;

[0039] The control logic signals of the main reactor systems are defined based on the defined process signals.

[0040] Based on the development process of a specific accident, a complete VVER-type reactor system control logic was established.

[0041] In one specific embodiment of the present invention, step 5 specifically includes:

[0042] For the reactor system program, the settings for disabled systems, enabled systems, and control signals during the accident process are determined based on the conservative assumptions made in the final safety analysis report regarding the accident.

[0043] For the core three-dimensional neutron dynamics program, the uncertainties to be considered are set in the configuration file according to the conservative conditions of the neutronics parameters under the accident as stated in the final safety analysis report.

[0044] In one specific embodiment of the present invention, step 6-3 specifically includes:

[0045] Step 6-3-1: Perform steady-state calculations of the reactor system program based on the read initialization state, and obtain the equilibrium state of the core, primary loop and secondary loop matching according to the input power distribution and system equipment boundary conditions;

[0046] Step 6-3-2: Organize the thermal-hydraulic parameters of each node obtained by the reactor system program under steady-state conditions, map the thermal-hydraulic parameters of each node based on the core grid in the reactor system program to the distribution of the core grid in the core three-dimensional neutron dynamics program, and pass them to the core three-dimensional neutron dynamics program through the interface function;

[0047] Step 6-3-3: Based on the thermal-hydraulic parameter distribution of each node, the core three-dimensional neutron dynamics program independently performs steady-state calculations of three-dimensional core diffusion to obtain a core three-dimensional power distribution that is adapted to the thermal-hydraulic parameters of each node.

[0048] The obtained three-dimensional power distribution is compared with the three-dimensional power distribution of the previous iteration step to obtain the relative deviation of the power. If the maximum value of the relative deviation is less than 1‰ or the number of iteration steps exceeds 15, then steady-state convergence is achieved, and coupling calculation under transient conditions continues; otherwise, the core three-dimensional power distribution needs to be sorted out and mapped to the core mesh of the system program, and fed back to the reactor system program, repeating steps 6-3-1 to 6-3-3.

[0049] In one specific embodiment of the present invention, step 6-4 specifically includes:

[0050] Step 6-4-1: After achieving steady-state convergence, organize the thermo-hydraulic parameters of each node in the reactor system program and transfer them to the core three-dimensional neutron dynamics program.

[0051] Step 6-4-2: Based on the thermal-hydraulic parameters of each node, call the transient calculation module of the core three-dimensional neutron dynamics program to perform three-dimensional core diffusion calculations separately, and adjust the neutron parameters used for subsequent accident analysis according to the uncertainties that need to be considered for each neutron parameter;

[0052] Step 6-4-3: Transfer the three-dimensional power distribution of the reactor core calculated by the three-dimensional core neutron dynamics program to the reactor system program;

[0053] Step 6-4-4: After obtaining the core power distribution, the reactor system program performs transient calculations;

[0054] During the transient calculations performed by the reactor system program,

[0055] The time step of the reactor system program is different from the time step of the core three-dimensional neutron dynamics program.

[0056] In the reactor system program, a maximum limit is set for the time step of the core three-dimensional neutron dynamics program. If the time step of the reactor system program is greater than the maximum limit, then the maximum limit is used as the time step of the core three-dimensional neutron dynamics program, and the core three-dimensional neutron dynamics program performs core three-dimensional neutron dynamics calculations independently until the time is accumulated to the time step of the reactor system program, and then data exchange between the programs occurs. If the time step of the reactor system program is less than or equal to the maximum limit, then the core three-dimensional neutron dynamics program performs core three-dimensional neutron dynamics calculations until the time is accumulated to the time step of the reactor system program, and then data exchange between the programs occurs.

[0057] In one specific embodiment of the present invention, the data interaction between the programs includes: the reactor system program feeding back the thermo-hydraulic parameters of each node to the core three-dimensional neutron dynamics program, and the core three-dimensional neutron dynamics program feeding back the core three-dimensional power distribution to the reactor system program.

[0058] Compared with existing technologies, the present invention provides a method for coupling a system program and a core three-dimensional neutron dynamics program for VVER-type pressurized water reactor accident analysis. The three-dimensional neutron dynamics program is compiled into a dynamic link library and coupled to the system program as a plug-in, thus achieving data transfer between the system program and the three-dimensional neutron dynamics program through internal coupling. In the coupled calculation, the system program acts as the leading program, driving the neutron dynamics program. Specifically, within the current time step, the system program first calculates the core state parameters and control parameters and feeds them back to the three-dimensional neutron dynamics program. Then, the three-dimensional neutron dynamics program updates the neutronics parameters based on the core state parameters, completes the three-dimensional core power distribution calculation, and provides it to the reactor system program through an interface. The three-dimensional neutron dynamics program can use a smaller time step than the system program, and the system program coordinates the synchronization during data transfer, thereby improving the simulation accuracy of the three-dimensional neutron dynamics. This coupling method can simulate various reactive accidents in VVER-type pressurized water reactors, thus establishing the capability for VVER-type pressurized water reactor accident analysis. Attached Figure Description

[0059] Figure 1 A flowchart illustrating the coupling method between the VVER reactor system program and the neutron dynamics program;

[0060] Figure 2 Interface functions representing the coupling interface;

[0061] Figure 3 A coupling model and data transfer diagram showing the reactor system program and the three-dimensional neutron dynamics program;

[0062] Figure 4 This diagram illustrates a reactor system control model using a boron dilution accident as an example.

[0063] Figure 5 This is a flowchart illustrating the coupled iterative calculation process between the reactor system program and the three-dimensional neutron dynamics program.

[0064] Figure 6 This diagram illustrates the interaction strategy between the reactor system program and the three-dimensional neutron dynamics program data. Detailed Implementation

[0065] To further understand the present invention, embodiments of the present invention are described below in conjunction with examples. However, it should be understood that these descriptions are only for further illustrating the features and advantages of the present invention, and not for limiting the present invention.

[0066] Embodiments of the present invention disclose a coupling method between a VVER reactor system program and a neutron dynamics program, such as... Figure 1 As shown, it includes the following steps:

[0067] Step 1: After adding interface functions to the core three-dimensional neutron dynamics program, compile it into a dynamic link library and allow it to be accessed and called by the reactor system program as a plug-in;

[0068] In the compilation of the three-dimensional neutron dynamics dynamic link library, interface functions are pre-built and compiled together with the code of the core three-dimensional neutron dynamics program into a dynamic link library that can be recognized by the reactor system program, so that the core three-dimensional neutron dynamics program can be accessed and called by the reactor system program in the form of a plug-in;

[0069] like Figure 2 As shown, the interface functions include an initialization function, a driver function, and a feedback function;

[0070] The initialization function is used to initialize the core three-dimensional neutron dynamics program and the mapping relationship for data transfer;

[0071] The driving parameters are used to transmit the thermal-hydraulic parameters of each node along the axis of the core three-dimensional neutron dynamics program, and drive the neutron dynamics program to complete the power distribution calculation; the thermal-hydraulic parameters of each node include the temperature and density distribution of the coolant, the fuel temperature, the boron concentration, and the position of the control rod.

[0072] The feedback function is used to map the three-dimensional power distribution calculated by the core three-dimensional neutron dynamics program to each node of the core in the reactor system program, and to feed the data back to the reactor system program.

[0073] More specifically, the interface function obtains the variable addresses in the reactor system program and reads the required data by using a hash chart, thereby achieving efficient and reliable data transfer between the reactor system program and the core three-dimensional neutron dynamics program.

[0074] Step 2: Establish a neutron physics model for a specific loading scheme of a VVER-type reactor;

[0075] The neutron physics model includes: neutron cross-sectional parameters of fuel assemblies for three-dimensional core neutron physics calculations; initial core state before accident analysis, such as three-dimensional core power distribution and burnup distribution, critical boron concentration, and control rod positions; the role of the neutron physics model is to provide the necessary input parameters for three-dimensional core neutron diffusion calculations based on the given initial core neutron physics information before accident analysis.

[0076] Step 2 specifically includes:

[0077] Step 2-1: Complete the arrangement of the core loading scheme for the VVER type reactor specific fuel cycle in the core three-dimensional neutronics program;

[0078] Step 2-2: Obtain the pre-generated cross-sectional information of various fuel assemblies based on the core loading scheme;

[0079] Steps 2-3: Use the core three-dimensional neutron dynamics program to perform three-dimensional steady-state neutron diffusion and burnup calculations to obtain a neutron physics model.

[0080] Step 3: Establish a simulation model of the accident analysis system for a VVER-type pressurized water reactor;

[0081] This accident analysis simulation model simulates the important systems and equipment in the primary and secondary loops of a VVER-type reactor, including the geometric topology and initialization parameters of each system. The purpose of the VVER-type pressurized water reactor accident analysis system simulation model is to simulate the operating characteristics of each system and equipment, thereby reflecting the response changes of coolant thermal fluid parameters within each system and equipment during the accident analysis process.

[0082] Specifically, the primary loop system of a reactor includes the reactor active zone, reflector, upper chamber, lower chamber, pressurizer, cold legs and hot legs of the four loops, primary side of the horizontal steam generator and main pumps, as well as the core emergency cooling system; the secondary loop system of the reactor includes the secondary side of the horizontal steam generator, bypass valves, atmospheric release valves and turbines, etc. Figure 3 As shown;

[0083] The accident analysis system simulation model for VVER-type pressurized water reactors is established within the reactor system program. Only after establishing this simulation model can thermal-hydraulic calculations and simulations of accidents be performed. The established VVER-type pressurized water reactor accident analysis system simulation model and the neutron physics model of the core three-dimensional neutron dynamics program need to exchange and transfer data at the core. Therefore, the modeling methods and mapping relationships of the core active zone components need to be considered in both programs. Specifically, in the reactor system program, the core active zone is divided into several channels based on fuel assemblies. In the three-dimensional neutron dynamics program, the mesh is also divided radially based on fuel assemblies, while in the height direction of the active zone, it is divided into 10–30 layers.

[0084] In this embodiment, a mapping relationship is established between the flow channels of the core components in the VVER type reactor system program and the core component distribution in the three-dimensional neutron dynamics program through the coupling interface between the reactor system program and the core system program.

[0085] In some embodiments, the three-dimensional neutron dynamics program may divide the core radially and axially into more or fewer grids;

[0086] Step 4: Based on the simulation model of the accident analysis system of the VVER type pressurized water reactor, establish the control logic of each major system and equipment in the primary and secondary loops of the VVER type reactor;

[0087] Specifically, process signals that define the thermal parameters of the thermal fluid object at a specific location and the overall core parameters, such as... Figure 4 As shown; the thermal fluid object is a pipe or valve, and the thermal parameters are pressure, temperature or enthalpy; the overall core parameter is the total core power;

[0088] The control logic signals for the main reactor systems are defined based on the defined process signals; such as control signals for triggering emergency shutdown and core boron concentration control signals, etc. Figure 4 ;

[0089] Based on the development process of a specific accident, a complete VVER-type reactor system control logic is established, namely the triggering of system action signals during the accident and the subsequent action response of the main systems in the reactor.

[0090] Step 5: Establish control logic and conservative condition settings for specific accidents;

[0091] When analyzing a specific accident in a VVER reactor, the process signals and control signals that trigger the accident are defined based on the accident development process in the final safety analysis report of the VVER reactor.

[0092] Step 5 specifically includes:

[0093] For the reactor system program, the settings for disabled systems, enabled systems, and control signals during the accident process are determined based on the conservative assumptions made in the final safety analysis report regarding the accident.

[0094] For the core three-dimensional neutron dynamics program, the uncertainties to be considered are set in the configuration file according to the conservative conditions of the neutronics parameters under the accident as stated in the final safety analysis report.

[0095] Step 6: Perform coupled calculations of the reactor system program and the three-dimensional neutron dynamics of the reactor core, such as... Figure 5 As shown;

[0096] Specifically, it includes:

[0097] Step 6-1: Read the reactor model input file and initialize the reactor system program calculation; the model includes a neutron physics model, an accident analysis system simulation model, and control logic;

[0098] Step 6-2: Initialization calculation of the three-dimensional neutron dynamics program for the reactor core;

[0099] Step 6-3: Perform coupled calculations under steady-state conditions. The reactor system parameters and neutronics parameters of each node are matched by iterative calculations of the reactor system program and the core three-dimensional neutron dynamics program.

[0100] Step 6-3 specifically includes:

[0101] Step 6-3-1: Perform steady-state calculations of the reactor system program based on the read initialization state. According to the input power distribution and system equipment boundary conditions and other parameters, obtain the equilibrium state of the core, primary loop and secondary loop matching.

[0102] Step 6-3-2: Organize the thermal-hydraulic parameters of each node obtained by the reactor system program under steady-state conditions, map the thermal-hydraulic parameters of each node based on the core grid in the reactor system program to the core grid distribution in the core 3D neutron dynamics program, and transfer them to the core 3D neutron dynamics program through interface functions; the thermal-hydraulic parameters of each node include: coolant temperature and density distribution, fuel temperature, boron concentration, and control rod position;

[0103] Step 6-3-3: Based on the thermal-hydraulic parameter distribution of each node, the core three-dimensional neutron dynamics program independently performs steady-state calculations of three-dimensional core diffusion to obtain a core three-dimensional power distribution that is adapted to the thermal-hydraulic parameters of each node.

[0104] The obtained three-dimensional power distribution is compared with the three-dimensional power distribution of the previous iteration step to obtain the relative deviation of the power. If the maximum value of the relative deviation is less than 1‰ or the number of iteration steps exceeds 15, then steady-state convergence is achieved, and coupling calculation under transient conditions continues; otherwise, the core three-dimensional power distribution needs to be sorted out and mapped to the core mesh of the system program, and fed back to the reactor system program, repeating steps 6-3-1 to 6-3-3.

[0105] Step 6-4: Perform coupling calculations under transient conditions. Within the initial time step, the reactor system program calculates the thermal-hydraulic parameters of each loop and node in the reactor system based on the steady-state power distribution fed back by the core three-dimensional neutron dynamics program.

[0106] The thermal-hydraulic parameters of each node are passed to the core three-dimensional neutron dynamics program through interface functions.

[0107] The core three-dimensional neutron dynamics program calculates the core three-dimensional power distribution based on the obtained thermo-hydraulic parameters of each node;

[0108] After obtaining the three-dimensional core power distribution, it is transmitted to the reactor system program, which then determines the new time step based on the thermal-hydraulic parameter status and calculates the new thermal-hydraulic parameters based on the three-dimensional core power distribution.

[0109] Repeat the above transient calculation process until the accumulated time reaches the set limit;

[0110] Step 6-4 specifically includes:

[0111] Step 6-4-1: After achieving steady-state convergence, organize the thermal-hydraulic parameters of each node in the reactor system program and transfer them to the core three-dimensional neutron dynamics program.

[0112] Step 6-4-2: Based on the transmitted node parameters, call the transient calculation module of the core three-dimensional neutron dynamics program to perform three-dimensional core diffusion calculations separately, and adjust the neutron parameters used for subsequent accident analysis according to the uncertainties that need to be considered for each neutron parameter;

[0113] Step 6-4-3: Transfer the three-dimensional power distribution of the reactor core calculated by the three-dimensional core neutron dynamics program to the reactor system program;

[0114] Step 6-4-4: After obtaining the core power distribution, the reactor system program performs transient calculations;

[0115] During transient calculations performed by the reactor system program, such as Figure 6 As shown,

[0116] The time step of the reactor system program is different from the time step of the core three-dimensional neutron dynamics program.

[0117] In the reactor system program, a maximum limit is set for the time step of the three-dimensional neutron dynamics program for the reactor core.

[0118] If the time step of the reactor system program is greater than the maximum limit, then the maximum limit is used as the time step of the core 3D neutron dynamics program. The core 3D neutron dynamics program performs core 3D neutron dynamics calculations independently until the time is accumulated to the time step of the reactor system program, and then data exchange between the programs occurs. If the time step of the reactor system program is less than or equal to the maximum limit, then the core 3D neutron dynamics program performs core 3D neutron dynamics calculations until the time is accumulated to the time step of the reactor system program, and then data exchange between the programs occurs.

[0119] The data interaction between the programs includes: the reactor system program feeding back the thermo-hydraulic parameters of each node to the core three-dimensional neutron dynamics program, and the core three-dimensional neutron dynamics program feeding back the core three-dimensional power distribution to the reactor system program.

[0120] Step 7: Based on the results of the coupled calculations, obtain the key safety parameters of the VVER reactor in a specific accident.

[0121] Specifically, the fluid pressure and temperature of the cold and hot pipe sections of the VVER pressurized water reactor during the accident process are obtained from the coupled calculation result file, as well as the pressure and level of the pressurizer, the pressure and level of the steam generator, the flow rate and temperature of the outlet steam, the flow rate of the secondary loop main feedwater, and the flow rates of the atmospheric release valve and bypass valve.

[0122] More importantly, it is necessary to obtain the highest center temperature of the fuel rods, the highest temperature of the cladding surface, the minimum deviation of the cladding surface from the nucleus boiling ratio, and the highest system pressure in the primary and secondary loops under the limiting power distribution conditions of the reactor core thermal channels.

[0123] Key safety parameters for VVER-type pressurized water reactors typically include the highest center temperature of the fuel rods, the highest temperature of the cladding surface, the minimum deviation from the nucleus boiling ratio of the cladding surface, and the highest system pressure in the primary and secondary loops.

[0124] The above description of the embodiments is only for the purpose of helping to understand the method and core ideas of the present invention. It should be noted that those skilled in the art can make several improvements and modifications to the present invention without departing from the principles of the present invention, and these improvements and modifications also fall within the protection scope of the claims of the present invention.

[0125] The above description of the disclosed embodiments enables those skilled in the art to make or use the invention. Various modifications to these embodiments will be readily apparent to those skilled in the art, and the general principles defined herein may be implemented in other embodiments without departing from the spirit or scope of the invention. Therefore, the invention is not to be limited to the embodiments shown herein, but is to be accorded the widest scope consistent with the principles and novel features disclosed herein.

Claims

1. A method of coupling a VVER reactor system program with a neutron kinetics program, characterized in that, The method comprises the following steps: Step 1: compiling the interface function into a dynamic link library after adding the interface function in the three-dimensional neutron dynamics program of the reactor core, and accessing and calling the interface function in the reactor system program in the form of a plug-in; Step 2: establishing a neutron physics model of a specific loading scheme of a VVER-type reactor; Specifically, Step 2-1: arranging the fuel cycle core loading scheme of the VVER-type reactor in the three-dimensional neutron program of the reactor core; Step 2-2: obtaining the cross-section table information of various types of fuel assemblies generated in advance according to the core loading scheme; Step 2-3: performing three-dimensional steady-state neutron diffusion and burnup calculation by using the three-dimensional neutron dynamics program of the reactor core to obtain the neutron physics model; Step 3: establishing an accident analysis system simulation model of the VVER-type pressurized water reactor in the reactor system program; Step 4: establishing the control logic of the main systems and devices of the primary loop and the secondary loop of the VVER-type reactor based on the accident analysis system simulation model of the VVER-type pressurized water reactor; Step 5: establishing the control logic and conservative condition setting for a specific accident; Step 6: performing coupled calculation of the reactor system program and the three-dimensional neutron dynamics of the reactor core; Specifically, Step 6-1: reading the reactor model input file, and initializing the calculation of the reactor system program; the model comprises the neutron physics model, the accident analysis system simulation model and the control logic; Step 6-2: initializing the calculation of the three-dimensional neutron dynamics program of the reactor core; Step 6-3: performing coupled calculation under steady-state conditions, and obtaining the core state in which the thermal-hydraulic parameters and the neutron parameters of each node of the reactor system are matched by iterative calculation of the reactor system program and the three-dimensional neutron dynamics program of the reactor core; Step 6-4: performing coupled calculation under transient conditions, and calculating the thermal-hydraulic parameters of each node in each loop and system of the reactor according to the steady-state power distribution fed back by the three-dimensional neutron dynamics program of the reactor core in the initial time step; passing the thermal-hydraulic parameters of each node to the three-dimensional neutron dynamics program of the reactor core through the interface function; calculating the three-dimensional power distribution of the reactor core by the three-dimensional neutron dynamics program of the reactor core according to the thermal-hydraulic parameters of each node obtained; after obtaining the three-dimensional power distribution of the reactor core, passing the three-dimensional power distribution to the reactor system program, and determining a new time step according to the parameter state by the reactor system program, then calculating new parameters according to the three-dimensional power distribution of the reactor core; repeating the above transient calculation process until the accumulated time reaches the set limit value; Step 7: obtaining the key safety parameters of the VVER-type reactor in a specific accident based on the results of the coupled calculation; the key safety parameters of the specific accident refer to the maximum center temperature of the fuel rod, the maximum surface temperature of the cladding, the minimum deviation from the nucleate boiling ratio of the cladding surface, and the maximum system pressure in the primary loop and the secondary loop. In step 1, 2. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that, the interface function is pre-prepared in the three-dimensional neutron dynamics dynamic link library, and is compiled into a dynamic link library that can be recognized by the reactor system program together with the code of the three-dimensional neutron dynamics program of the reactor core, so that the three-dimensional neutron dynamics program of the reactor core is accessed and called in the form of a plug-in by the reactor system program; ​ The interface function includes an initialization function, a driving function and a feedback function.

3. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 2, characterized in that, The initialization function is used for initializing the core three-dimensional neutron kinetics program and the interface parameters, and giving the node information of the coolant and fuel in the core part of the reactor system model through the mapping relationship. The driving parameter is used for transmitting the thermal-hydraulic parameters of each node in the axial direction of the core three-dimensional neutron kinetics program, and driving the neutron kinetics program to complete the power distribution calculation. The feedback function is used for mapping the three-dimensional power distribution calculated by the core three-dimensional neutron kinetics program to each node of the core of the reactor system program, and feeding back the data to the reactor system program. The thermal-hydraulic parameters of each node include the temperature and density distribution of the coolant, the fuel temperature, the boron concentration and the control rod position.

4. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 3, characterized in that, In step 4, 5. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that, Define the thermal-hydraulic parameters of the thermal-hydraulic object at a specific location and the process signal of the core overall parameter; Define the control logic signal of the main system of the reactor based on the defined process signal; Based on the development process of the specific accident, a complete control logic of the VVER type reactor system is established. The step 5 specifically includes:

6. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that, For the reactor system program, the setting values of the disabled system, the enabled system and the control signal in the accident process are set according to the conservative assumption conditions about the accident in the final safety analysis report; For the core three-dimensional neutron kinetics program, the uncertainties required to be considered are set in the configuration file according to the conservative conditions of the neutron parameters in the final safety analysis report. The step 6-3 specifically includes:

7. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that, Step 6-3-1: based on the read initialization state, the reactor system program is calculated in a steady state, and the equilibrium state of the core, the primary loop and the secondary loop is obtained according to the input power distribution and the system device boundary condition; Step 6-3-2: the thermal-hydraulic parameters of each node obtained by the reactor system program in the steady state are sorted out, the thermal-hydraulic parameters of each node based on the core grid in the reactor system program are mapped to the distribution of the core grid in the core three-dimensional neutron kinetics program, and are transmitted to the core three-dimensional neutron kinetics program through the interface function; Step 6-3-3: based on the distribution of the thermal-hydraulic parameters of each node transmitted, the core three-dimensional neutron kinetics program independently performs steady-state calculation of three-dimensional core diffusion to obtain the three-dimensional power distribution of the core suitable for the thermal-hydraulic parameters of each node; The relative deviation of the obtained three-dimensional power distribution and the three-dimensional power distribution of the previous iteration step is compared to obtain the relative deviation of the power, and if the maximum relative deviation is less than 1 ‰ or the iteration step number exceeds 15 steps, the steady state converges, and the coupled calculation under the transient condition is continued; otherwise, the core three-dimensional power distribution needs to be sorted out and mapped to the core grid of the system program, and is fed back to the reactor system program, and steps 6-3-1 to 6-3-3 are repeated. The step 6-4 specifically includes:

8. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 7, characterized in that, ​ Step 6-4-1: After reaching steady-state convergence, the thermal-hydraulic parameters of each node of the reactor system program are sorted and transferred to the three-dimensional core neutron kinetics program; Step 6-4-2: Based on the transferred thermal-hydraulic parameters of each node, the three-dimensional core diffusion calculation of the three-dimensional core neutron kinetics program transient calculation module is separately performed, and the neutron parameters for subsequent accident analysis are adjusted according to the uncertainty to be considered for each neutron parameter; Step 6-4-3: The three-dimensional core power distribution calculated by the three-dimensional core neutron kinetics program is transferred to the reactor system program; Step 6-4-4: After obtaining the core power distribution, the reactor system program performs transient calculation; During the transient calculation process of the reactor system program, The time step of the reactor system program is different from the time step of the three-dimensional core neutron kinetics program; The reactor system program sets a maximum limit value for the time step of the three-dimensional core neutron kinetics program, If the time step of the reactor system program is greater than the maximum limit value, the maximum limit value is used as the time step of the three-dimensional core neutron kinetics program, the three-dimensional core neutron kinetics calculation of the three-dimensional core neutron kinetics program is performed until the time is accumulated to the time step of the reactor system program, and then the data interaction between the programs is performed; if the time step of the reactor system program is less than or equal to the maximum limit value, the three-dimensional core neutron kinetics program performs three-dimensional core neutron kinetics calculation until the time is accumulated to the time step of the reactor system program, and then the data interaction between the programs is performed.

9. The method of coupling a VVER reactor system program with a neutron kinetics program according to claim 8, characterized in that, The data interaction between the programs includes: the reactor system program feeding back the thermal-hydraulic parameters of each node to the three-dimensional core neutron kinetics program, and the three-dimensional core neutron kinetics program feeding back the three-dimensional core power distribution to the reactor system program.

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