A multi-scale joint simulation calculation method for nuclear thermal propulsion micro-reactor

Abstract

The application discloses a kind of nuclear thermal propulsion micro reactor multi-scale combined simulation calculation method, to solve the problem that the core part is analyzed using point reactor model under current system scale, simulation approximation is larger, cannot consider the spatial effect of core, cannot reflect the accurate simulation process of reactor core part in reactor system analysis.Specifically, a nuclear thermal propulsion micro reactor system simulation model is established based on the system program of Modelica language, a core fluid-structure coupling calculation model is established based on the core scale calculation program of open source multi-physics coupling framework MOOSE, real-time data exchange is realized through TCP communication, multi-scale combined simulation calculation of reactor system and core fluid-structure coupling analysis is realized, the real-time interactive simulation technology problem of reactor system scale parameters and core scale parameters is solved, so that the power distribution, temperature distribution and other spatial characteristics of reactor core are described in real time while the reactor system program is calculated.

Description

Technical Field

[0001] This invention relates to a method for multi-physics coupling and multi-scale simulation calculation of reactors, specifically a method for multi-scale joint simulation calculation of nuclear thermal propulsion microreactors. Background Technology

[0002] Nuclear propulsion is currently a cutting-edge and hotly debated technology in the field of space propulsion. Major nuclear powers and those in the field of space propulsion are accelerating breakthroughs in overcoming bottlenecks in nuclear propulsion technology development, aiming to lead the future development of space propulsion technology. Nuclear thermal propulsion, with its high energy density, high specific impulse, and long endurance, has become one of the preferred solutions for space propulsion. The microreactor, the heat source of nuclear thermal propulsion, is considered its heart; obtaining accurate reactor calculation parameters is a key challenge in nuclear thermal propulsion technology.

[0003] Parameter calculations for nuclear thermal propulsion microreactors generally include reactor system, physics, thermal engineering, safety, and nuclear fuel performance analysis, involving reactor system scale and core scale. Traditional coupled calculation methods mainly study the coupled calculation problems of nuclear (reactor physics), thermal (thermal fluids), and mechanical (fuel or structural material mechanical properties), focusing on the reactor core scale. At the reactor system scale, the main considerations are the impact of parameter changes in pumps, steam generators, and reactor system loops within the system loop. The general calculation method abstracts system components as a single point, using a lumped parameter to represent a component in system performance studies. Common system analysis tools include, but are not limited to, RELAP, Modelica programming languages, or custom-written system programs. In this case, the reactor core at the system scale is typically analyzed using a point reactor model (a model for solving reactor dynamics equations that ignores neutron spatial distribution; the reactor is simplified to a point without spatial scale, hence the name point reactor model) to obtain curves of lumped parameters such as power and temperature changing over time, which are then used in system analysis. This type of point reactor model, which does not consider the spatial distribution effects of the reactor and only uses lumped point parameters to replace reactor core parameters, has a relatively high degree of approximation and cannot accurately reflect the simulation process of the reactor core in reactor system analysis. In summary, traditional methods face the following technical problems:

[0004] 1) The analysis objects of the reactor system are the components at the system scale, and each component is equivalent to a point model during the calculation. The reactor core, as a component of the reactor system, requires detailed three-dimensional spatial simulation calculation due to its structural complexity and the necessity of safety analysis. The current analysis of the core at the system scale uses a point reactor model, which cannot consider the core space effect and cannot reflect the accurate simulation process of the reactor core in the reactor system analysis.

[0005] 2) Currently, there is a lack of effective real-time interaction and joint simulation techniques for core-scale and reactor system-scale parameters of nuclear thermal propulsion microreactors. Summary of the Invention

[0006] The purpose of this invention is to provide a multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors, in order to solve the technical problems of using point reactor models to analyze the reactor core at the current system scale, which results in a large degree of approximation, cannot consider the core space effect, cannot reflect the accurate simulation process of the reactor core in the reactor system analysis, and currently lacks effective real-time interaction and co-simulation technology for parameters at the core scale and reactor system scale for nuclear thermal propulsion microreactors.

[0007] To achieve the above objectives, this invention provides a multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors, characterized by the following steps:

[0008] Step 1: Establish a simulation model of the nuclear thermal propulsion microreactor system, set initial parameters, configure the system simulation program's calculation environment, set the communication port and communication parameters, start the system simulation program, begin simulation calculations, and obtain the system simulation program's calculation results; activate the system simulation program's interface functions and wait for the establishment of a communication connection with the reactor core calculation program;

[0009] Step 2: Establish a neutronics calculation model for the core of the nuclear thermal propulsion microreactor, and calculate the core power distribution based on the neutronics calculation model;

[0010] Step 3: Construct a core fluid-structure interaction calculation model for use by the core calculation program. Set initial boundary conditions based on the initial operating conditions of the reactor, introduce the core power distribution calculated in Step 2 as the heat source of the core fluid-structure interaction calculation model, start the fluid-structure interaction calculation, obtain the calculation results of the core calculation program, and automatically activate the interface function of the core calculation program.

[0011] Step 4: The core calculation program sends the calculation results to the system simulation program;

[0012] Step 5: After receiving the calculation results from the core calculation program, the system simulation program transmits the calculation results from the system simulation program obtained in Step 1 to the core calculation program.

[0013] Step 6: After the core calculation program receives the calculation results from the system simulation program, the communication connection between the system simulation program and the core calculation program is interrupted.

[0014] Step 7: Time step, update the current calculation time point;

[0015] Step 8: The system simulation program starts a new simulation calculation. After the calculation is completed, it is determined whether the preset convergence condition of the system simulation program has been met. If it has been met, wait for step 9 to be completed before executing step 10. If it has not been met, the system simulation program interface function is reactivated, and it waits to establish a communication connection with the core calculation program. Steps 4-7 are repeated until the preset convergence condition of the system simulation program is met. Then, wait for step 9 to be completed before executing step 10.

[0016] Meanwhile, the core calculation program starts a new step of fluid-structure interaction calculation. After the calculation is completed, it is determined whether the preset convergence condition of the core calculation program has been met. If it has been met, the calculation result of the last step is sent to the system simulation program, and then step 9 is executed. If it has not been met, the core calculation program interface function is reactivated, a communication connection is established with the system simulation program, and steps 4-7 are repeated until the preset convergence condition of the core calculation program is met, and then step 9 is executed.

[0017] Step 9: The core calculation program obtains the end flag signal, sends the end flag signal to the system simulation program, and the core calculation program ends the operation;

[0018] Step 10: The system simulation program receives the end flag signal sent by the core calculation program, interrupts the communication connection between the system simulation program and the core calculation program, and the system simulation program ends the operation.

[0019] Furthermore, in step 1, the calculation results include coolant outlet flow rate parameters, outlet temperature parameters, and inlet pressure parameters.

[0020] Furthermore, the interface function mentioned in step 1 is specifically:

[0021] The transTCDData interface function is used to exchange data with the core computing program.

[0022] Furthermore, step 2 specifically involves:

[0023] A neutronics calculation model of the core of a nuclear thermal propulsion microreactor was established. The neutron transport equations were solved using a neutronics solver to obtain the core neutron flux density distribution. Multiplying the neutron flux density distribution by the energy production cross-section yielded the core fission power distribution. The neutron transport equations are as follows:

[0024]

[0025] in, Let be the neutron angular flux density at position r with energy E and direction of motion Ω; Σ(r,E) is the macroscopic cross section of a neutron with energy E at position r; Q s (r,Ω,E) represents a scattering neutron source at position r with energy E and direction of motion Ω; Q f(r,Ω,E) represents a fission neutron source with energy E at position r and direction Ω.

[0026] Furthermore, in step 2, the neutronics solution program is the Monte Carlo program OpenMC, the continuous finite element neutron transport program, or the discontinuous finite element neutron transport program.

[0027] Furthermore, in step 3, the core fluid-structure interaction calculation model can be modeled using structured meshes and unstructured meshes.

[0028] Furthermore, in step 3, the interface function of the core calculation program is ConnectToModelica.

[0029] Furthermore, the communication protocol between the core computing program and the system simulation program is TCP.

[0030] Furthermore, in step 1, the simulation model of the nuclear thermal propulsion microreactor system is established based on the system program of the Modelica language.

[0031] Furthermore, in step 3, the core fluid-structure interaction calculation model is constructed based on the core-scale calculation program of the open-source multi-physics coupling framework MOOSE.

[0032] The beneficial effects of this invention are:

[0033] 1. The method provided by this invention realizes multi-scale joint simulation calculation of reactor system and core fluid-structure interaction analysis, and solves the technical problem of real-time interactive simulation of reactor system scale parameters and core scale parameters, so as to plot the spatial characteristics of reactor core such as power distribution and temperature distribution in real time while the reactor system program is calculating.

[0034] 2. This invention can significantly improve the computational accuracy of nuclear thermal propulsion microreactor system analysis.

[0035] 3. This invention can comprehensively consider the impact of system simulation calculation and core fluid-structure interaction calculation on the core space characteristics.

[0036] 4. The core fluid-structure interaction calculation model in this invention can be modeled using structured meshes and unstructured meshes, making core geometry modeling highly flexible;

[0037] 5. This invention can assess the safety of nuclear thermal propulsion microreactor systems while also obtaining the three-dimensional temperature field distribution at the reactor core scale, thus enabling a more comprehensive assessment of the operational safety of the microreactor core.

[0038] 6. The joint simulation data transmission method established by this invention is based on the system simulation program interface function and the core calculation program interface function, and the communication protocol is TCP. It has good adaptability and is not limited to micro reactors. Attached Figure Description

[0039] Figure 1 This is a flowchart of an embodiment of the multi-scale co-simulation calculation method for a nuclear thermal propulsion microreactor according to the present invention. Detailed Implementation

[0040] The technical solutions of the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings. Obviously, the described embodiments are only some embodiments of the present invention, and not all embodiments. Based on the embodiments of the present invention, all other embodiments obtained by those skilled in the art without creative effort are within the scope of protection of the present invention.

[0041] The basic framework of this invention is as follows:

[0042] A simulation model of a nuclear thermal propulsion microreactor system was established using a system program based on the Modelica language. A core fluid-structure interaction calculation model was established using a core-scale calculation program based on the open-source multiphysics coupling framework MOOSE. Data was exchanged in real time via TCP communication.

[0043] like Figure 1 As shown, the multi-scale co-simulation calculation method for a nuclear thermal propulsion microreactor provided in this embodiment includes the following steps:

[0044] Step 1: Establish a simulation model of the nuclear thermal propulsion microreactor system, set initial parameters, configure the simulation environment, set the communication port number (e.g., 8000), set the communication time interval (as the simulation calculation step size), set the array size for transmission input and output, set the simulation time, then start the system simulation program to begin simulation calculations, and obtain the system simulation program calculation results. These simulation program calculation results refer to the initial parameters of the system simulation program.

[0045] The system simulation program implements communication requirements through external function calls to an interface. A communication class, TCPServerConnect (custom class name), is defined to construct TCPSocket objects (object name constructed by the custom class). The entire communication transmission process is implemented by activating the transTCDData interface function (inventor's function name). The communication class and interface function are as follows:

[0046] 1) The TCPServerConnect class is called during the initialization function to establish a communication connection.

[0047] 2) The transTCDData interface function is used to exchange data with the core computing program.

[0048] Step 2: Establish a neutronics calculation model for the core of a nuclear thermal propulsion microreactor, and use neutronics solvers, including but not limited to the Monte Carlo program OpenMC and deterministic neutron transport programs (continuous finite element or discontinuous finite element neutron transport programs), to solve the neutron transport equations and obtain the core power distribution.

[0049] The neutron transport equation is as follows:

[0050]

[0051] In the formula, Let be the neutron angular flux density at position r with energy E and direction of motion Ω; Σ(r,E) is the macroscopic cross section of a neutron with energy E at position r; Q s (r,Ω,E) represents a scattering neutron source at position r with energy E and direction of motion Ω; Q f (r,Ω,E) represents a fission neutron source at position r with energy E and a direction of motion of Ω. Solving the neutron transport equations yields the neutron flux density distribution in the reactor core. Multiplying this neutron flux density distribution by the energy production cross-section gives the core power distribution q. v ;

[0052] Step 3: Construct a core fluid-structure interaction calculation model for use by the core calculation program. Set initial boundary conditions based on the initial operating conditions of the reactor, including initial parameters such as coolant inlet flow rate, inlet temperature, and outlet pressure. Start the core calculation program, which will automatically activate the core calculation program interface function ConnectToModelica (custom interface function name). In this function, call the transmission process using the set port number and IP address, as shown in the example below:

[0053] void ConnectToModelica(TCPSocketConnect*tcpData)

[0054] {

[0055] tcpData->Init();

[0056] tcpData->Connect(const_cast<char*> (ip.data()),port);

[0057] tcpData->SendData(send_data,4);

[0058] tcpData->GetData();

[0059] tcpData->Close();

[0060] }

[0061] Step 4: The above function ConnectToModelica is responsible for sending the calculation results of the core calculation program to the system simulation program. The data transmitted before the first step of the calculation is the initial boundary condition data.

[0062] Step 5: After receiving the initial boundary condition data sent by the core calculation program, namely the coolant outlet flow rate parameters, outlet temperature parameters, and inlet pressure parameters, the system simulation program then transmits the new boundary condition data, such as the coolant inlet flow rate parameters, inlet temperature parameters, and outlet pressure parameters, to the core calculation program.

[0063] Step 6: After the core calculation program receives the boundary condition data transmitted from the system simulation program, the system simulation program and the core calculation program simultaneously interrupt their communication connection.

[0064] Step 7: Time stepping. Here, the stepping time refers to the one-step time set by the system simulation program, updating the current time. Subsequently, the system simulation program begins one step of simulation calculation, and the core calculation program begins one step of fluid-structure interaction calculation. The fluid-structure interaction calculation includes solving the three-dimensional heat conduction equation to obtain the temperature distribution inside the core, with the heat source being the core power distribution q obtained in Step 2. v Fluid dynamics calculations solve the mass conservation equation, momentum conservation equation, and energy conservation equation to obtain parameters such as coolant outlet flow rate, outlet temperature, and inlet pressure.

[0065] The above three-dimensional heat conduction equation is:

[0066] The above mass conservation equation is:

[0067] The above momentum conservation equation is:

[0068] The above energy conservation equation is:

[0069] In the formula, k is the thermal conductivity coefficient, T is the temperature, and q v The core power distribution is given by z, the flow channel direction coordinate, A, and ρ. f Let u be the fluid density and u be the fluid velocity. E represents the average pressure at the boundary of the curved surface. f Let H be the total fluid energy, f be the friction factor, and H be the total fluid energy. w Let α be the wall heat transfer coefficient. w T is the heat transfer surface density.wall For wall temperature, T bulk Let P be the surface average volumetric temperature of the liquid, and D be the fluid pressure. h It is the hydraulic diameter.

[0070] Step 8: After the simulation calculation is completed, determine whether the preset convergence condition of the system simulation program has been met. If it has been met, wait for step 9 to be completed before executing step 10. If it has not been met, reactivate the system simulation program interface function and wait for the calculation result of the core calculation program. Repeat steps 4-7 until the preset convergence condition of the system simulation program is met, wait for step 9 to be completed before executing step 10.

[0071] Meanwhile, the core calculation program starts a new step of fluid-structure interaction calculation. After the calculation is completed, it is determined whether the preset convergence condition of the core calculation program has been met. If it has been met, the calculation result of the last step is sent to the system simulation program, and then step 9 is executed. If it has not been met, the core calculation program interface function is reactivated, a communication connection is established with the system simulation program, and steps 4-7 are repeated until the preset convergence condition of the core calculation program is met, and then step 9 is executed.

[0072] The preset convergence condition refers to: the relative error limit of the calculation results of the solid thermal conductivity temperature in this step and the previous step; and the relative error limit of the calculation results of the fluid temperature in this step and the previous step.

[0073] Step 9: The core calculation program obtains the end flag signal and sends the end flag signal is_end to the system simulation program, and the core calculation program ends the operation;

[0074] Step 10: The system simulation program receives the end flag signal sent by the core calculation program, interrupts the communication connection between the system simulation program and the core calculation program, and the system simulation program ends the operation.

[0075] The termination flag signal is an is_end flag set during communication transmission. This flag is transmitted by the core computing program to control the calling of the simulation program's interface functions. If the core computing program terminates its operation, the system simulation program also needs to terminate the calling of its interface functions. The core computing program determines whether the termination flag signal has been received; if so, it sends this termination flag signal to the system simulation program to end the operation.

[0076] The above description is merely a specific embodiment of the present invention, but the scope of protection of the present invention is not limited thereto. Any changes or substitutions within the technical scope disclosed in the present invention should be covered within the scope of protection of the present invention. Therefore, the scope of protection of the present invention should be determined by the scope of the claims.

Claims

1. A multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors, characterized in that, Includes the following steps: Step 1: Establish a simulation model of the nuclear thermal propulsion microreactor system, set initial parameters, configure the system simulation program's calculation environment, set the communication port and communication parameters, start the system simulation program, begin simulation calculations, and obtain the system simulation program's calculation results; activate the system simulation program's interface functions and wait for the establishment of a communication connection with the reactor core calculation program; Step 2: Establish a neutronics calculation model for the core of the nuclear thermal propulsion microreactor. Solve the neutron transport equations using a neutronics solver to obtain the core neutron flux density distribution. Multiply the neutron flux density distribution by the energy production cross section to obtain the core fission power distribution. The neutron transport equations are as follows: ； in, Let be the neutron angular flux density at position r with energy E and direction of motion Ω; The macroscopic cross section of a neutron with energy E at position r; This is a scattering neutron source with energy E at position r and direction of motion Ω; This is a fission neutron source at position r with energy E and direction of motion Ω; Step 3: Construct a core fluid-structure interaction calculation model for use by the core calculation program. Set initial boundary conditions based on the initial operating conditions of the reactor, introduce the core power distribution calculated in Step 2 as the heat source of the core fluid-structure interaction calculation model, start the fluid-structure interaction calculation, obtain the calculation results of the core calculation program, and automatically activate the interface function of the core calculation program. Step 4: The core calculation program sends the calculation results to the system simulation program; Step 5: After receiving the calculation results from the core calculation program, the system simulation program transmits the calculation results from the system simulation program obtained in Step 1 to the core calculation program. Step 6: After the core calculation program receives the calculation results from the system simulation program, the communication connection between the system simulation program and the core calculation program is interrupted. Step 7: Time step, update the current calculation time point; Step 8: The system simulation program starts a new simulation calculation. After the calculation is completed, it is determined whether the preset convergence condition of the system simulation program has been met. If it has been met, wait for step 9 to be completed before executing step 10. If it has not been met, the system simulation program interface function is reactivated, and it waits to establish a communication connection with the core calculation program. Steps 4-7 are repeated until the preset convergence condition of the system simulation program is met. Then, wait for step 9 to be completed before executing step 10. Meanwhile, the core calculation program starts a new step of fluid-structure interaction calculation. After the calculation is completed, it is determined whether the preset convergence condition of the core calculation program has been met. If it has been met, the calculation result of the last step is sent to the system simulation program, and then step 9 is executed. If it has not been met, the core calculation program interface function is reactivated, a communication connection is established with the system simulation program, and steps 4-7 are repeated until the preset convergence condition of the core calculation program is met, and then step 9 is executed. Step 9: The core calculation program obtains the end flag signal, sends the end flag signal to the system simulation program, and the core calculation program ends the operation; Step 10: The system simulation program receives the end flag signal sent by the core calculation program, interrupts the communication connection between the system simulation program and the core calculation program, and the system simulation program ends the operation.

2. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 1, characterized in that: In step 1, the calculation results include coolant outlet flow rate parameters, outlet temperature parameters, and inlet pressure parameters.

3. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 1 or 2, characterized in that, The interface function mentioned in step 1 is specifically: The transTCDData interface function is used to exchange data with the core computing program.

4. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 3, characterized in that: In step 2, the neutronics solution program is the Monte Carlo program OpenMC, the continuous finite element neutron transport program, or the discontinuous finite element neutron transport program.

5. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 4, characterized in that: In step 3, the core fluid-structure interaction calculation model is modeled using both structured and unstructured meshes.

6. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 5, characterized in that: In step 3, the interface function of the core calculation program is ConnectToModelica.

7. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 6, characterized in that: The communication protocol between the core calculation program and the system simulation program is TCP.

8. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 7, characterized in that: In step 1, the simulation model of the nuclear thermal propulsion microreactor system is established based on the system program of the Modelica language.

9. The multi-scale co-simulation calculation method for nuclear thermal propulsion microreactors according to claim 8, characterized in that: In step 3, the core fluid-structure interaction calculation model is constructed based on the core-scale calculation program of the open-source multi-physics coupling framework MOOSE.

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