A multi-dimensional coupling calculation method for sodium-cooled fast reactor

Abstract

The application discloses a kind of sodium-cooled fast reactor multidimensional coupling calculation methods, steps are as follows:1, establish sodium-cooled fast reactor core CFD porous medium geometric model;2, establish sodium-cooled fast reactor main loop, intermediate loop model;3, set initial condition completes the preliminary calculation of sodium-cooled fast reactor core computational fluid dynamics software;4, the calculation result of computational fluid dynamics software is calculated as the boundary condition of sodium-cooled fast reactor system program;5. The calculation result obtained in the fourth step is calculated as the boundary condition of neutron dynamics module;6, using the calculation result obtained in the fifth step as the boundary condition of computational fluid dynamics software to calculate the thermal hydraulic parameter of core;7, repeat step four to step six, until the whole system completes full period calculation.

Description

Technical Field

[0001] This invention belongs to the field of nuclear reactor system technology, specifically relating to a multi-dimensional coupled calculation method for a sodium-cooled fast reactor. Background Technology

[0002] Sodium-cooled fast reactors (SNCRs) have over 400 reactor-years of operational experience. During their development, numerous SNCR system analysis programs have been developed, and the thermo-hydraulic characteristics of SNCR systems under transient accident conditions have been a hot research topic internationally. SNCR system analysis programs typically perform a holistic model of the entire SNCR system, dividing the reactor into control volumes for each structural component and solving the physical fields using one-dimensional conservation equations. However, these programs cannot capture the detailed three-dimensional thermo-hydraulic parameters of the reactor core, leading to inaccurate simulations of key phenomena. In contrast, computational fluid dynamics (CFD) software, based on the Navier-Stokes three-dimensional flow equations, can perform precise numerical simulations of flow and heat transfer phenomena in complex geometric regions, thus overcoming the limitations of SNCR system analysis programs. However, for the complex structure of a sodium-cooled fast reactor (NSCRT) system, the number of grids required for full-system simulation using CFD methods would be in the hundreds of millions, posing a significant challenge to current computing power. Therefore, CFD methods have not been widely applied in the simulation of full-system NSCRT systems. With the development of NSCRTs, the interaction between physics and thermal engineering in nuclear reactors has received increasing attention. Changes in thermal parameters can affect neutron physics feedback in NSCRTs and may also lead to changes in neutron flux distribution, affecting heat source distribution. Therefore, nuclear-thermal coupling analysis is crucial. Summary of the Invention

[0003] To overcome the problems existing in the prior art, the present invention aims to provide a multi-dimensional coupled calculation method for sodium-cooled fast reactors (NSFRs), designed to meet the needs of cross-dimensional, multi-physics field phenomenon analysis of the entire NFR system under reactive introduction conditions, and to improve the calculation accuracy of transient conditions in NFRs. This method combines the advantages of three-dimensional computational fluid dynamics software and one-dimensional NFR system analysis programs, which can save computational resources and improve efficiency, and can also obtain important thermo-hydraulic parameters in the NFR as accurately as possible.

[0004] To achieve the above objectives, the present invention is implemented through the following technical solution:

[0005] A multi-dimensional coupled computational method for sodium-cooled fast reactors includes the following steps:

[0006] Step 1: Establish a CFD porous media geometric model of the reactor core, including fuel assemblies, steel assemblies, control rod assemblies, reflector assemblies, and measurement assemblies. The fuel and control rod assemblies are divided into four parts from bottom to top: lower grid plate segment, fuel segment, fission gas chamber segment, and upper grid plate segment. The steel, reflector, and measurement assemblies are divided into three parts from bottom to top: lower grid plate segment, fuel segment, and upper grid plate segment. After obtaining the geometric model, mesh generation is performed.

[0007] Step 2: Establish the main loop and intermediate loop models of the sodium-cooled fast reactor using the sodium-cooled fast reactor system analysis program. First, establish the node diagram of the sodium-cooled fast reactor system. Fill in the input cards of the sodium-cooled fast reactor system analysis program according to the node diagram. The main loop model includes the cold sodium pool, hot sodium pool, pressure pipelines, valves, main pump, primary side of the intermediate heat exchanger, inlet chamber and outlet chamber. The intermediate loop model includes the inlet pipeline, outlet pipeline and secondary side of the intermediate heat exchanger.

[0008] Step 3: Set up the computational fluid dynamics software to solve the computational model, set the initial boundary conditions of the core at the initial time T0, and calculate the outlet coolant temperature, flow rate, and outlet pressure of the core;

[0009] Step 4: Place T n The core outlet coolant temperature and flow rate, calculated by computational fluid dynamics software, are used as boundary conditions at the core outlet chamber in the sodium-cooled fast reactor system analysis program. Here, n = 0, 1, 2, 3…; T is calculated by the sodium-cooled fast reactor system analysis program. n+1 The thermal-hydraulic parameters of the main loop and intermediate loop at all times, including the temperature, pressure, flow rate and enthalpy of each node;

[0010] Step 5: By using T n+1 The thermo-hydraulic parameters calculated by the sodium-cooled fast reactor system program at time T are used as boundary conditions. The neutron dynamics module solves the neutron diffusion equation and updates the temperature distribution and reactivity feedback parameters of the neutron physics field to obtain T. n+1 The total power, decay power, and total reactivity of the reactor core at any given time;

[0011] Step 6: Place T n+1 The reactor core power calculated by the neutron dynamics module at each moment is used as the boundary condition for computational fluid dynamics software to obtain the three-dimensional thermal-hydraulic parameters of the core, and simultaneously obtain T. n+1 The core outlet coolant flow rate, temperature, and core outlet pressure at any given time are further calculated by the sodium-cooled fast reactor system analysis program to determine T. n+1 Temperature, pressure, flow rate, and enthalpy at each node in the time loop;

[0012] Step 7: Repeat steps four through six until the preset full-time calculation is completed, realizing multi-dimensional coupled calculation of the sodium-cooled fast reactor.

[0013] The present invention has the following advantages and beneficial effects:

[0014] 1. The method of the present invention can perform one-dimensional modeling of the main loop and intermediate loop of the sodium-cooled fast reactor system, while simultaneously performing three-dimensional modeling of the core of key parts and core-thermal coupling analysis. This can improve computational efficiency and obtain detailed three-dimensional thermal-hydraulic parameters of key parts.

[0015] 2. The method of the present invention provides a coupling scheme for the analysis program of sodium-cooled fast reactor system, which is applicable to most reactor thermal-hydraulic analysis programs, such as FLUENT, RELAP5, etc.

[0016] 3. The method of the present invention performs data interaction and resets boundary conditions at each time step to ensure high calculation accuracy. Attached Figure Description

[0017] Figure 1 A schematic diagram illustrating the multi-dimensional coupled calculation of similar operating conditions for the reactivity of a sodium-cooled fast reactor. Detailed Implementation

[0018] The invention will now be described in further detail with reference to the accompanying drawings, taking the sodium-cooled fast reactor reactivity introduction condition as an example:

[0019] This invention provides a multi-dimensional coupled calculation method for sodium-cooled fast reactors, the specific steps of which are as follows:

[0020] Step 1: Use 3D geometric modeling software to create a CFD porous media geometric model of the reactor core, including the fuel assembly, steel assembly, control rod assembly, reflector assembly, and measurement assembly, ignoring other assemblies. The fuel assembly and control rod assembly are divided into four parts from bottom to top: lower grid plate segment, fuel segment, fission gas chamber segment, and upper grid plate segment. The steel assembly, reflector assembly, and measurement assembly are divided into three parts from bottom to top: lower grid plate segment, fuel segment, and upper grid plate segment. After obtaining the geometric model, use a meshing tool to create a hexahedral mesh.

[0021] Step 2: Model the entire sodium-cooled fast reactor system using the sodium-cooled fast reactor system analysis program, establishing the main loop and intermediate loop models: First, establish the node diagrams for the main loop and intermediate loop. Fill in the input cards of the sodium-cooled fast reactor system analysis program according to the node diagrams, including the length, wetted perimeter, flow area, zenith angle, and azimuth angle of each part. The main loop model equipment includes the cold sodium pool, hot sodium pool, pressure pipes, valves, main pump, primary side of the intermediate heat exchanger, inlet chamber, and outlet chamber. The intermediate loop model equipment includes the inlet pipe, outlet pipe, and secondary side of the intermediate heat exchanger. After establishing the sodium-cooled fast reactor main loop and intermediate loop models, the sodium-cooled fast reactor system analysis program performs node division and subsequent simulation calculations; fill in the neutron dynamics module cards, including Doppler coefficient, fuel expansion feedback coefficient, coolant density feedback coefficient, and cladding expansion feedback coefficient. Set the accident sequence for overpower transient conditions in the sodium-cooled fast reactor system analysis program; calculate the total pump head using the pump head calculation formula.

[0022] In the formula: The coolant mass flow rate is expressed in kg·s. -1 ; Represents pump head / m; This indicates the pump speed in rpm; b1, b2, b3, and b4 are constants.

[0023] Step 3: Set the computational model to be solved by the computational fluid dynamics software according to the operating conditions, set the initial boundary conditions at the initial time T0 of the core, compile the user-defined function to realize the functions of data transfer and calling the sodium-cooled fast reactor system analysis program, and calculate the outlet coolant temperature, flow rate and outlet pressure of the core after the settings are completed.

[0024] Step 4: As Figure 1 As shown, T n The core outlet coolant temperature and flow rate, calculated by computational fluid dynamics software, are used as boundary conditions at the core outlet chamber in the sodium-cooled fast reactor system analysis program. Here, n = 0, 1, 2, 3…; T is calculated by the sodium-cooled fast reactor system analysis program. n+1 The thermal-hydraulic parameters of the main loop and intermediate loop at all times, including the temperature, pressure, flow rate and enthalpy of each node.

[0025] Step 5: By using T n+1 The thermal-hydraulic parameters of the core inlet chamber, calculated by the instantaneous sodium-cooled fast reactor system program, including temperature, flow rate, pressure, and enthalpy, are used as boundary conditions. The neutron dynamics module calculates and solves the neutron diffusion equation to obtain the neutron flux distribution at different locations in the core and the overall effective multiplication coefficient. It also updates the temperature distribution, Doppler feedback coefficient, coolant density feedback coefficient, fuel expansion feedback coefficient, and cladding expansion feedback coefficient of the neutron physics field, ultimately obtaining T.n+1 The total power, decay power, and total reactivity of the reactor core at any given time.

[0026] Step 6: Place T n+1 The reactor core power calculated by the neutron dynamics module at each moment is used as the boundary condition for computational fluid dynamics software to obtain detailed three-dimensional thermal-hydraulic parameters of the core, and simultaneously obtain T. n+1 The core outlet coolant flow rate, temperature, and core outlet pressure at any given time are further calculated by the sodium-cooled fast reactor system analysis program to determine T. n+1 Temperature, pressure, flow rate, and enthalpy at each node in the time loop.

[0027] Step 7: Repeat steps four through six until the sodium-cooled fast reactor system analysis program and computational fluid dynamics software complete the preset full-time period calculation of the sodium-cooled fast reactor full-field power outage transient condition, realizing multi-dimensional coupled calculation of the sodium-cooled fast reactor.

[0028] This invention combines the features of a system program and computational fluid dynamics software, coupling the two programs to interact with each other during the calculation process. This allows for the capture of detailed three-dimensional thermal-hydraulic physical field information of the reactor core, as well as the acquisition of transient changes under various complex operating conditions in the sodium-cooled fast reactor loop. Simultaneously, the sodium-cooled fast reactor system program can add a neutron dynamics calculation module for nuclear-thermal coupling analysis. By inputting reactivity coefficients such as coolant density, fuel expansion, and cladding expansion into input cards, the impact of thermal parameter changes on the neutron physics parameters of the sodium-cooled fast reactor core can be further captured, enabling more accurate calculation of reactor power. Therefore, the multi-dimensional coupling calculation method for sodium-cooled fast reactor reactivity-introduced operating conditions proposed in this paper has reference value and guiding significance for the transient operating condition analysis of sodium-cooled fast reactors.

[0029] The above content is a further elaboration on the coupling method for a specific scheme. The parts not described in detail are common knowledge to those skilled in the art.

Claims

1. A multi-dimensional coupled computational method for sodium-cooled fast reactors, characterized in that: For the sodium-cooled fast reactor (NSCFR) reactivity introduction scenario, computational fluid dynamics (CFD) software was used to perform three-dimensional thermal-hydraulic calculations on the NSCFR core. The NSCFR system analysis program was used to perform one-dimensional thermal-hydraulic calculations on various parts of the main loop and intermediate loop. The neutron dynamics module of the NSCFR system analysis program was used to complete the calculation of the core nuclear physical reactivity during reactivity introduction. Multi-dimensional, full-system coupled calculations of the NSCFR were achieved through boundary parameter transfer between the computational fluid dynamics software, the NSCFR system analysis program, and the neutron dynamics module. The method includes the following steps: Step 1: Establish a CFD porous media geometric model of the reactor core, including fuel assemblies, steel assemblies, control rod assemblies, reflector assemblies, and measurement assemblies. The fuel and control rod assemblies are divided into four parts from bottom to top: lower grid plate segment, fuel segment, fission gas chamber segment, and upper grid plate segment. The steel, reflector, and measurement assemblies are divided into three parts from bottom to top: lower grid plate segment, fuel segment, and upper grid plate segment. After obtaining the geometric model, mesh generation is performed. Step 2: Establish the main loop and intermediate loop models of the sodium-cooled fast reactor using the sodium-cooled fast reactor system analysis program: First, establish the node diagrams of the main loop and intermediate loop of the sodium-cooled fast reactor system. Fill in the input cards of the sodium-cooled fast reactor system analysis program according to the node diagrams. The main loop model includes the cold sodium pool, hot sodium pool, pressure pipelines, valves, main pump, primary side of the intermediate heat exchanger, inlet chamber and outlet chamber. The intermediate loop model includes the inlet pipeline, outlet pipeline and secondary side of the intermediate heat exchanger. Step 3: Set up the computational fluid dynamics software to solve the computational model, set the initial boundary conditions of the core at the initial time T0, and calculate the outlet coolant temperature, flow rate, and outlet pressure of the core; Step 4: Place T n The core outlet coolant temperature and flow rate, calculated by computational fluid dynamics software, are used as boundary conditions at the core outlet chamber in the sodium-cooled fast reactor system analysis program. Here, n = 0, 1, 2, 3…; T is calculated by the sodium-cooled fast reactor system analysis program. n+1 The thermal-hydraulic parameters of the main loop and intermediate loop at all times, including the temperature, pressure, flow rate and enthalpy of each node; Step 5: By using T n+1 The thermo-hydraulic parameters calculated by the sodium-cooled fast reactor system program at time T are used as boundary conditions. The neutron dynamics module solves the neutron diffusion equation and updates the temperature distribution and reactivity feedback parameters of the neutron physics field to obtain T. n+1 The total power, decay power, and total reactivity of the reactor core at any given time; Step 6: Place T n+1 The reactor core power calculated by the neutron dynamics module at each moment is used as the boundary condition for computational fluid dynamics software to obtain the three-dimensional thermal-hydraulic parameters of the core, and simultaneously obtain T. n+1 The core outlet coolant flow rate, temperature, and core outlet pressure at any given time are further calculated by the sodium-cooled fast reactor system analysis program to determine T. n+1 Temperature, pressure, flow rate, and enthalpy at each node in the time loop; Step 7: Repeat steps four through six until the preset full-time calculation is completed, realizing multi-dimensional coupled calculation of the sodium-cooled fast reactor.

2. The multi-dimensional coupled calculation method for a sodium-cooled fast reactor according to claim 1, characterized in that: The reactive feedback parameters mentioned in step 5 include the Doppler feedback coefficient, coolant density feedback coefficient, fuel expansion feedback coefficient, and cladding expansion feedback coefficient.

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