Coupling method of VVER unit system program and neutron dynamics program

Abstract

The invention relates to the field of nuclear power unit accident analysis, in particular to a coupling method of a VVER unit system program and a neutron dynamics program. The method comprises the following steps: adding an interface function in a reactor core three-dimensional neutron dynamics program, compiling the reactor core three-dimensional neutron dynamics program into a dynamic link library, and accessing and calling the dynamic link library by a reactor system program in a plug-in manner; establishing a neutronics physical model of a specific loading scheme of the VVER type reactor; establishing an accident analysis system simulation model of the VVER type pressurized water reactor; establishing control logic of main systems and equipment of a primary loop and a secondary loop in the VVER type reactor; establishing control logic and conservative condition settings for specific accidents; executing coupling calculation of a reactor system program and reactor core three-dimensional neutron dynamics; and obtaining key safety parameters of the VVER type reactor in a specific accident based on a coupling calculation result. According to the invention, simulation of changes of physical and thermal parameters in a primary loop system and a secondary loop system and action response of a main system after a reactor has an accident can be realized, and efficient and reliable accident analysis can be conveniently carried out on a VVER type nuclear power unit.

Description

Technical Field

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

[0002] Commercial pressurized water reactors need to be shut down for refueling every 12 or 18 months. When the reactor is refueled, some of the irradiated fuel assemblies will be removed from the reactor core and replaced with new, unirradiated fuel assemblies. The irradiated fuel assemblies that have reached a certain burnup and the new fuel assemblies that have not been removed from the core will be rearranged in the core to form a new core loading plan. The safety of the new core loading plan needs to be confirmed through nuclear design calculations and evaluation of key neutronics parameters. When a key neutronics parameter exceeds the limit in the final safety analysis report of the nuclear power plant, it is necessary to reanalyze the accident related to the key neutronics parameter to verify that the consequences of the accident can still meet the relevant acceptance criteria.

[0003] Therefore, nuclear power unit accident analysis is a crucial part of the nuclear energy field. It involves a comprehensive and in-depth assessment of various abnormal situations and potential accidents that nuclear power units may encounter during operation. Especially in the core area of ​​the nuclear reactor, physical phenomena such as flow, heat transfer, neutronics, and mechanical action interact with each other to form an extremely complex system. The simulation analysis of these phenomena and processes requires the collaborative work of multiple professional programs to ensure the accuracy and reliability of the analysis.

[0004] In the accident analysis of nuclear power units, the coupling relationship between the physical and thermal phenomena inside the core is particularly close, and often shows drastic dynamic changes. For example, in the analysis of control rod ejection accidents, as the control rods are quickly ejected from the core after the drive mechanism fails, it will first cause rapid changes in the total power and power spatial distribution of the core in terms of neutron physics. The changes in power and its distribution will cause changes in the coolant temperature field and density field in the core in terms of thermal hydraulics. The changes in the coolant temperature field and density field will affect the temporal dynamic behavior of the core power and its distribution in a feedback manner. Therefore, in the analysis of accidents related to reactivity, physical and thermal coupling calculations are particularly important, which requires the coupling program to accurately simulate and predict these complex physical and thermal phenomena.

[0005] In addition, 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 to improve the computational efficiency of the coupled program.

[0006] In the development and application of accident analysis tools for nuclear power units, there are relatively more analysis tools for square pressurized water reactors. However, for VVER pressurized water reactors, there is still a relative lack of tools that directly couple system programs with core three-dimensional neutron dynamics programs to complete specific accident analysis. Therefore, the development of a physical thermal coupling program that can specifically simulate transient accidents of VVER units to enhance the evaluation of the physical thermal characteristics of VVER pressurized water reactors and their response capabilities under accident conditions is of great significance to the safe operation of VVER pressurized water reactors. Summary of the invention

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

[0008] The present invention provides a method for coupling a VVER reactor system program with a neutron kinetics program, 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 access and call it in the reactor system program in a plug-in manner;

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

[0011] Step 3: Establishing a simulation model of the accident analysis system of 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, the control logic of the main systems and equipment of the primary and secondary loops in the VVER type reactor is established;

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

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

[0015] Specifically include:

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

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

[0018] Step 6-3: Perform coupled calculation under steady-state conditions, and obtain the core state in which the thermal-hydraulic parameters of each node of the reactor system match the neutronics parameters through iterative calculation of the reactor system program and the core three-dimensional neutron dynamics program;

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

[0020] The thermal hydraulic parameters of each node are transferred to the core three-dimensional neutron dynamics program through the interface function;

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

[0022] After the three-dimensional core power distribution is obtained, it is passed to the reactor system program, and the reactor system program determines a new time step according to the parameter state, and then calculates new parameters according to 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 coupling calculation, obtain the key safety parameters of the VVER reactor in a specific accident.

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

[0026] In the compilation of the three-dimensional neutron dynamics dynamic link library, the interface function is prefabricated, and the interface function is jointly compiled 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 a plug-in manner;

[0027] In the interface function, the variable address in the reactor system program is obtained through a hash chart to read the required data, thereby achieving efficient and reliable data transmission between the reactor system program and the core three-dimensional neutron dynamics program.

[0028] In a 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 gives the node information of the coolant and fuel in the core part of the reactor system model through mapping relationships;

[0030] The driving parameters are used to transfer the thermal hydraulic parameters of each axial node 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 three-dimensional neutron dynamics program of the core to each node of the reactor system program, and feed back the data to the reactor system program.

[0032] In a 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 control rod position.

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

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

[0035] Step 2-2: Obtaining pre-generated cross-section table information of various fuel assemblies according to the core loading plan;

[0036] Step 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 a specific embodiment of the present invention, in step 4,

[0038] Process signals that define thermal parameters of thermal fluid objects at specific locations and overall core parameters;

[0039] defining control logic signals for the main systems of the reactor based on the defined process signals;

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

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

[0042] For the reactor system program, the set values ​​of the disabled systems, enabled systems and control signals during the accident process are set according to the conservative assumptions of the accident in the final safety analysis report;

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

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

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

[0046] Step 6-3-2: Arrange 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 transfer 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 transmitted, the core three-dimensional neutron dynamics program independently performs steady-state calculation of three-dimensional core diffusion to obtain the core three-dimensional power distribution that is compatible with 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 power. If the maximum value of the relative deviation is less than 1‰ or the number of iteration steps exceeds 15, the steady state convergence is achieved and the coupling calculation under transient conditions is continued; otherwise, the three-dimensional power distribution of the core needs to be sorted out and mapped to the core grid of the system program, and fed back to the reactor system program, and steps 6-3-1 to 6-3-3 are repeated.

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

[0050] 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 core three-dimensional neutron dynamics program;

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

[0052] Step 6-4-3: passing the three-dimensional core power distribution 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 calculation of the reactor system program,

[0055] The time step of the reactor system program is different from that of the core 3D neutron dynamics program;

[0056] A maximum limit is set for the time step of the core three-dimensional neutron dynamics program in the reactor system program. If the time step of the reactor system program is greater than the maximum limit, 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 the core three-dimensional neutron dynamics calculation alone until the time is accumulated to the time step of the reactor system program, and then data exchange between programs is performed; if the time step of the reactor system program is less than or equal to the maximum limit, the core three-dimensional neutron dynamics program performs the core three-dimensional neutron dynamics calculation until the time is accumulated to the time step of the reactor system program, and then data exchange between programs is performed.

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

[0058] Compared with the prior art, the method for coupling the system program and the core three-dimensional neutron dynamics program for VVER type pressurized water reactor accident analysis of the present invention is that the three-dimensional neutron dynamics program is compiled into a dynamic link library and coupled with the system program in a plug-in manner, thereby realizing data transmission between the system program and the three-dimensional neutron dynamics program in an internal coupling manner; in the coupling calculation, the system program drives the neutron dynamics program as a leading program, that is, within the current time step, the system program first completes the calculation of the core state parameters and control parameters and feeds back to the three-dimensional neutron dynamics program, and then the three-dimensional neutron dynamics program updates the neutron parameters according to 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 transmission, thereby improving the simulation accuracy of the three-dimensional neutron dynamics; the coupling method can simulate various reactive accidents of the VVER type pressurized water reactor, thereby establishing the ability of VVER type pressurized water reactor accident analysis. BRIEF DESCRIPTION OF THE DRAWINGS

[0059] Figure 1 A flow chart showing the method of coupling the VVER reactor system program with the neutron dynamics program;

[0060] Figure 2 Interface functions representing coupling interfaces;

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

[0062] Figure 4 A schematic diagram showing a reactor system control model taking a boron dilution accident as an example;

[0063] Figure 5 It represents the iterative calculation flow chart of the coupling between the reactor system program and the three-dimensional neutron dynamics program;

[0064] Figure 6 Diagram showing the interaction strategy between the reactor system program and the 3D neutron dynamics program data. DETAILED DESCRIPTION

[0065] In order to further understand the present invention, the embodiments of the present invention are described below in conjunction with examples, but it should be understood that these descriptions are only for further illustrating the features and advantages of the present invention, rather than limiting the present invention.

[0066] The embodiment of the present invention discloses a method for coupling a VVER reactor system program with a neutron kinetics program, such as Figure 1 As shown, the following steps are included:

[0067] Step 1: After adding interface functions to the core three-dimensional neutron dynamics program, compile it into a dynamic link library, and access and call it in the reactor system program in a plug-in manner;

[0068] In the compilation of the three-dimensional neutron dynamics dynamic link library, the interface function is prefabricated, and the interface function is jointly compiled 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 a plug-in manner;

[0069] like Figure 2 As shown, the interface function includes an initialization function, a driving function and a feedback function;

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

[0071] The driving parameters are used to transfer the thermal-hydraulic parameters of each axial node of the three-dimensional neutron dynamics program of the core, 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 control rod position.

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

[0073] More specifically, in the interface function, the data required for reading the variable address in the reactor system program is obtained through a hash chart, thereby achieving efficient and reliable data transmission between the reactor system program and the core three-dimensional neutron dynamics program.

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

[0075] The neutron physics model includes: neutron cross-section parameters of fuel assemblies for three-dimensional core neutron calculations; the initial state of the core 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 by giving the initial core neutron physics information before accident analysis.

[0076] The step 2 specifically includes:

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

[0078] Step 2-2: Obtaining pre-generated cross-section table information of various fuel assemblies according to the core loading plan;

[0079] Step 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 of a VVER type pressurized water reactor;

[0081] The accident analysis simulation model simulates the important systems and equipment in the primary and secondary loops of the VVER reactor, including the geometric topology and initialization parameters of each system. The role of the accident analysis system simulation model of the VVER pressurized water reactor is to simulate the operating characteristics of each system equipment, so as to reflect the response changes of the coolant thermal fluid parameters in each system equipment during the accident analysis process.

[0082] Specifically, the primary loop system of the reactor includes the reactor active area, reflector, upper chamber, lower chamber, pressurizer, cold legs and hot legs of the four loops, primary side and main pump of the horizontal steam generator, and core emergency cooling system; the secondary loop system of the reactor includes the secondary side of the horizontal steam generator, bypass valve, atmospheric release valve and steam turbine, etc. Figure 3 As shown;

[0083] The simulation model of the accident analysis system of the VVER pressurized water reactor is established in the reactor system program. Only after this simulation model is established can the thermal hydraulic calculation simulation of the accident be carried out. The simulation model of the accident analysis system of the VVER pressurized water reactor and the neutron physics model of the core three-dimensional neutron dynamics program need to exchange and transfer data at the core, so it is necessary to consider the modeling method and mapping relationship of the core active area components in the two programs. The simulation model of the accident analysis system of the VVER pressurized water reactor established in the reactor system program exchanges and transfers data at the core with the neutron physics model of the three-dimensional neutron dynamics program. To this end, more specifically, in the reactor system program, the core active area is divided into several channels based on the fuel assembly. In the three-dimensional neutron dynamics program, the grid is also divided in the radial direction based on the fuel assembly, and it is divided into 10 to 30 layers in the height direction of the active area;

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

[0085] In some embodiments, the three-dimensional neutron dynamics program will divide more or fewer grids in the radial and axial directions of the core;

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

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

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

[0089] Based on the development process of the specific accident, a complete VVER reactor system control logic is established, namely the triggering of system action signals during the accident and the subsequent action responses 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 of a VVER reactor, the process signals and control signals of the initiating events that caused the accident are defined according to the accident development process in the final safety analysis report of the VVER reactor.

[0092] The step 5 specifically includes:

[0093] For the reactor system program, the set values ​​of the disabled systems, enabled systems and control signals during the accident process are set according to the conservative assumptions of the accident in the final safety analysis report;

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

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

[0096] Specifically include:

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

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

[0099] Step 6-3: Perform coupling calculation under steady-state conditions, and obtain the core state in which the parameters of each node of the reactor system match the neutronics parameters through iterative calculation of the reactor system program and the core three-dimensional neutron dynamics program;

[0100] The step 6-3 specifically includes:

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

[0102] Step 6-3-2: Arrange 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 transfer them to the core three-dimensional neutron dynamics program through the interface function; 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;

[0103] Step 6-3-3: Based on the thermal-hydraulic parameter distribution of each node transmitted, the core three-dimensional neutron dynamics program independently performs steady-state calculation of three-dimensional core diffusion to obtain the core three-dimensional power distribution that is compatible with 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 power. If the maximum value of the relative deviation is less than 1‰ or the number of iteration steps exceeds 15, the steady state convergence is achieved and the coupling calculation under transient conditions is continued; otherwise, the three-dimensional power distribution of the core needs to be sorted out and mapped to the core grid of the system program, and fed back to the reactor system program, and steps 6-3-1 to 6-3-3 are repeated.

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

[0106] The thermal hydraulic parameters of each node are transferred to the core three-dimensional neutron dynamics program through the interface function;

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

[0108] After the three-dimensional core power distribution is obtained, it is passed to the reactor system program, and the reactor system program determines a new time step according to the state of the thermal hydraulic parameters, and then calculates new thermal hydraulic parameters according to the three-dimensional core power distribution;

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

[0110] The step 6-4 specifically includes:

[0111] 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 core three-dimensional neutron dynamics program;

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

[0113] Step 6-4-3: passing the three-dimensional core power distribution 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 the transient calculation of the reactor system program, Figure 6 As shown,

[0116] The time step of the reactor system program is different from that of the core 3D neutron dynamics program;

[0117] The reactor system program sets a maximum limit on the time step for the core three-dimensional neutron dynamics program.

[0118] If the time step of the reactor system program is greater than the maximum limit, the maximum limit will be used as the time step of the core three-dimensional neutron dynamics program, and the core three-dimensional neutron dynamics program will perform the core three-dimensional neutron dynamics calculations alone until the time is accumulated to the time step of the reactor system program, and then data exchange between programs will be performed; if the time step of the reactor system program is less than or equal to the maximum limit, the core three-dimensional neutron dynamics program will perform the core three-dimensional neutron dynamics calculations until the time is accumulated to the time step of the reactor system program, and data exchange between programs will be performed.

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

[0120] Step 7: Based on the results of the coupling calculation, 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, the pressure and liquid level of the pressurizer, the pressure, liquid level, outlet steam flow and temperature of the steam generator, the main feed water flow of the secondary circuit, the flow of the atmospheric release valve and the bypass valve, etc. are obtained from the coupling calculation result file.

[0122] More importantly, the maximum center temperature of the fuel rods, the maximum temperature of the cladding surface, the minimum deviation from nucleate boiling ratio of the cladding surface, and the maximum system pressure in the first and second circuits in the reactor core hot channel under extreme power distribution conditions are obtained.

[0123] The key safety parameters of a VVER pressurized water reactor usually include the maximum center temperature of the fuel rod, the maximum temperature of the cladding surface, the minimum deviation from nucleate boiling ratio of the cladding surface, and the maximum system pressure in the primary and secondary circuits.

[0124] The above embodiments are only used to help understand the method and core idea of ​​the present invention. It should be noted that, for those skilled in the art, several improvements and modifications can be made to the present invention without departing from the principles of the present invention, and these improvements and modifications also fall within the scope of protection of the claims of the present invention.

[0125] The above description of the disclosed embodiments enables one skilled in the art to implement or use the present invention. Various modifications to these embodiments will be apparent to one 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 present invention. Therefore, the present invention will not be limited to the embodiments shown herein, but rather to the widest scope consistent with the principles and novel features disclosed herein.

Claims

1. A method for coupling a VVER reactor system program with a neutron kinetics program, characterized in that: The following steps are involved: Step 1: After adding interface functions to the core three-dimensional neutron dynamics program, compile it into a dynamic link library, and access and call it in the reactor system program in a plug-in manner; Step 2: Establish a neutronics physics model for a specific loading scheme of a VVER reactor; Step 3: Establishing a simulation model of the accident analysis system of a VVER type pressurized water reactor in the reactor system program; Step 4: Based on the simulation model of the accident analysis system of the VVER type pressurized water reactor, the control logic of the main systems and equipment of the primary and secondary loops in the VVER type reactor is established; Step 5: Establish control logic and conservative condition settings for specific accidents; Step 6: Execute the coupled calculation of the reactor system program and the three-dimensional neutron dynamics of the core; Specifically include: Step 6-1: Read the reactor model input file and initialize the calculation of the reactor system program; the model includes a neutronics physics model, an accident analysis system simulation model and control logic; Step 6-2: Initialization calculation of the core three-dimensional neutron dynamics program; Step 6-3: Perform coupled calculation under steady-state conditions, and obtain the core state in which the thermal-hydraulic parameters of each node of the reactor system match the neutronic parameters through iterative calculation of the reactor system program and the core three-dimensional neutron dynamics program; Step 6-4: Perform coupling calculation under transient conditions. Within the initial time step, the reactor system program calculates the thermal hydraulic parameters of each loop of the reactor and each node in the system according to the steady-state power distribution fed back by the three-dimensional neutron dynamics program of the core; The thermal hydraulic parameters of each node are transferred to the core three-dimensional neutron dynamics program through the interface function; The core three-dimensional neutron dynamics program calculates the core three-dimensional power distribution based on the thermal hydraulic parameters of each node; After the three-dimensional core power distribution is obtained, it is passed to the reactor system program, and the reactor system program determines a new time step according to the parameter state, and then calculates new parameters according to the three-dimensional core power distribution; Repeat the above transient calculation process until the accumulated time reaches the set limit; Step 7: Based on the results of the coupling calculation, obtain the key safety parameters of the VVER reactor in a specific accident.

2. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that: In the step 1, In the compilation of the three-dimensional neutron dynamics dynamic link library, an interface function is prefabricated, and the interface function is jointly compiled 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 a plug-in manner; In the interface function, the variable address in the reactor system program is obtained through a hash chart to read the required data, thereby achieving efficient and reliable data transmission between the reactor system program and the core three-dimensional neutron dynamics program.

3. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 2, characterized in that: The interface function includes an initialization function, a driving function and a feedback function; The initialization function is used to initialize the core three-dimensional neutron dynamics program and interface parameters, and gives the node information of the coolant and fuel in the core part of the reactor system model through mapping relationships; The driving parameters are used to transfer the thermal hydraulic parameters of each axial node of the core three-dimensional neutron dynamics program and drive the neutron dynamics program to complete the power distribution calculation; The feedback function is used to map the three-dimensional power distribution calculated by the three-dimensional neutron dynamics program of the core to each node of the reactor system program, and feed back the data to the reactor system program.

4. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 3, characterized in that: 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.

5. The method for coupling a VVER reactor system program with a neutron dynamics program according to claim 1, characterized in that: The step 2 specifically includes: Step 2-1: Complete the layout of the core loading scheme for the specific fuel cycle of the VVER type reactor in the core three-dimensional neutronics program; Step 2-2: obtaining pre-generated cross-section table information of various fuel assemblies according to the core loading plan; Step 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.

6. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that: In step 4, Process signals that define thermal parameters of thermal fluid objects at specific locations and overall core parameters; defining control logic signals for the main systems of the reactor based on the defined process signals; Based on the development process of the specific accident, a complete VVER reactor system control logic is established.

7. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that: The step 5 specifically includes: For the reactor system program, the set values ​​of the disabled systems, enabled systems and control signals during the accident process are set according to the conservative assumptions of the accident in the final safety analysis report; For the core three-dimensional neutron dynamics program, the uncertainties that need to be considered are set in the configuration file according to the conservative conditions of the neutronic parameters under the accident in the final safety analysis report.

8. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 1, characterized in that: The step 6-3 specifically includes: Step 6-3-1: Perform steady-state calculation of the reactor system program based on the read initialization state, and obtain the equilibrium state of the core, primary circuit and secondary circuit matching according to the input power distribution and system equipment boundary conditions; Step 6-3-2: Arrange 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 transfer them to the core three-dimensional neutron dynamics program through the interface function; Step 6-3-3: Based on the thermal-hydraulic parameter distribution of each node transmitted, the core three-dimensional neutron dynamics program independently performs steady-state calculation of three-dimensional core diffusion to obtain the core three-dimensional power distribution that is compatible with the thermal-hydraulic parameters of each node; 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 power. If the maximum value of the relative deviation is less than 1‰ or the number of iteration steps exceeds 15, the steady state convergence is achieved and the coupling calculation under transient conditions is continued; otherwise, the three-dimensional power distribution of the core needs to be sorted out and mapped to the core grid of the system program, and fed back to the reactor system program, and steps 6-3-1 to 6-3-3 are repeated.

9. The method for coupling a VVER reactor system program with a neutron kinetics program according to claim 8, characterized in that: The step 6-4 specifically includes: 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 core three-dimensional neutron dynamics program; Step 6-4-2: Based on the thermal-hydraulic parameters of each node transmitted, the transient calculation module of the core three-dimensional neutron dynamics program is called to perform three-dimensional core diffusion calculation separately, and the neutronic parameters used for subsequent accident analysis are adjusted according to the uncertainties that need to be considered for each neutronic parameter; Step 6-4-3: passing the three-dimensional core power distribution calculated by the three-dimensional core neutron dynamics program to the reactor system program; Step 6-4-4: After obtaining the core power distribution, the reactor system program performs transient calculations; During the transient calculation of the reactor system program, The time step of the reactor system program is different from that of the core 3D neutron dynamics program; The reactor system program sets a maximum limit on the time step for the core three-dimensional neutron dynamics program. If the time step of the reactor system program is greater than the maximum limit, the maximum limit will be used as the time step of the core three-dimensional neutron dynamics program, and the core three-dimensional neutron dynamics program will perform the core three-dimensional neutron dynamics calculation alone until the time is accumulated to the time step of the reactor system program, and then data exchange between programs will be performed; if the time step of the reactor system program is less than or equal to the maximum limit, the core three-dimensional neutron dynamics program will perform the core three-dimensional neutron dynamics calculation until the time is accumulated to the time step of the reactor system program, and then data exchange between programs will be performed.

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

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