Nuclear reactor core two-phase thermal hydraulic analysis method
By combining subchannel 3D geometric modeling and a two-fluid six-equation model with a high-resolution discretization method, the problem of insufficient lateral flow description in two-phase thermal-hydraulic analysis of nuclear reactor cores by traditional subchannel programs is solved, achieving high-precision computational efficiency and stability.
Patent Information
- Authority / Receiving Office
- CN · China
- Patent Type
- Applications(China)
- Current Assignee / Owner
- Filing Date
- 2026-02-04
- Publication Date
- 2026-04-07
AI Technical Summary
In the two-phase thermal-hydraulic analysis of nuclear reactor cores, existing technologies, such as traditional subchannel programs, suffer from insufficient description of lateral flow and local non-uniformity, large computational scale, and difficulty in direct application to engineering-level safety analysis. Furthermore, it is difficult to introduce high-resolution discretization schemes.
By employing sub-channel 3D geometric modeling combined with a two-fluid six-equation model and a high-resolution finite volume discretization method, the 3D geometry and main flow characteristics are explicitly reflected, improving computational accuracy and stability.
While maintaining the advantage of computing scale, it significantly improves the resolution and computational accuracy of physical field distribution, while reducing computation time and hardware resource requirements.
Smart Images
Figure CN121809340A_ABST
Abstract
Description
Technical Field
[0001] This invention belongs to the field of reactor thermal-hydraulic safety analysis technology, specifically relating to a two-phase thermal-hydraulic analysis method for nuclear reactor cores. Background Technology
[0002] Two-phase thermal-hydraulic behavior within the reactor core is one of the core issues in nuclear reactor safety analysis, directly affecting fuel cladding integrity and safety margin assessment. During normal operation and power variations, boiling may occur within the core channels. This process is influenced by factors such as geometry, flow redistribution, and heat transfer coupling, placing high demands on the physical description capabilities and computational efficiency of numerical analysis methods. Existing two-phase thermal-hydraulic analysis of the reactor core mainly relies on system-level programs, subchannel methods, and refined computational fluid dynamics (CFD) methods. System analysis programs offer high computational efficiency but are typically based on one-dimensional or quasi-one-dimensional assumptions, limiting their ability to describe lateral flow and local non-uniformity in complex rod bundle channels. Refined CFD methods offer higher spatial resolution, but their large computational scale makes them difficult to directly apply to engineering-level safety analysis and multi-condition assessments. Therefore, in practical engineering applications, subchannel analysis methods have long been the primary tool balancing computational cost and analytical capabilities. However, traditional subchannel programs mostly use homogeneous flow or drift flow models and are based on 1.5-dimensional modeling assumptions, which still have shortcomings in terms of lateral flow characterization and physical consistency. At the same time, their software architecture is relatively closed and their discretization schemes are conservative, which also limits the introduction of advanced two-phase flow models and high-resolution discretization schemes. Summary of the Invention
[0003] In order to overcome the problems existing in the prior art, the purpose of this invention is to provide a two-phase thermal-hydraulic analysis method for nuclear reactor cores, which can explicitly reflect the three-dimensional geometry and main flow characteristics while maintaining the computational scale advantages of the subchannel method, and improve the computational accuracy while maintaining computational stability by using a high-resolution format.
[0004] To achieve the above objectives, the present invention adopts the following technical solution: A two-phase thermal-hydraulic analysis method for a nuclear reactor core includes the following steps: Step 1: Based on the subchannel structure of the reactor core, determine the modeling parameters of the subchannel and the fuel rod, establish a three-dimensional geometric model of the subchannel, and use mesh generation software to generate the computational mesh of the three-dimensional geometric model; Step 2: Based on the sub-channel modeling parameters determined in Step 1, establish the basic governing equations for the vapor and liquid phases based on the two-fluid six-equation model, and establish a constitutive model for the closed basic governing equations. Step 3: Based on the fuel rod modeling parameters determined in Step 1, establish a fuel rod thermal conductivity model; Step 4: Discretize the two-fluid six-equation model in Step 2 and the fuel rod heat conduction model established in Step 3 using the finite volume discretization method to obtain the fluid and solid discrete equation sets. Step 5: Based on the fluid and solid discrete equations obtained in Step 4, establish a fluid and solid coupled solution process.
[0005] The subchannel modeling parameters determined in step 1 include: identification of various types of subchannels, namely inner channels, side channels and corner channels; flow area and hydraulic diameter of each type of subchannel; fuel rod size; grid spacing between fuel rods; and grid spacing between fuel rods and component edges.
[0006] The two-fluid six-equation model in step 2 includes the two-phase continuity equation, the two-phase momentum equation, and the two-phase energy equation, specifically: (6) (7) (8) In the formula: k — Represents either the vapor phase or the liquid phase; t — Represents time, seconds; — Represents the phase fraction of the vapor or liquid phase; —Represents the density of the vapor or liquid phase, kg / m³ 3 ; — Represents the velocity of the vapor or liquid phase, in m / s; —Represents the mass transfer rate per unit volume of the vapor or liquid phase, kg / (m³) 3 ·s); — Represents pressure, MPa; F w，k — Represents the frictional resistance between the wall surface and the vapor or liquid phase, kg / (m 2 ·s 2 ); F b，k —Represents the gravity of the vapor or liquid phase, kg / (m³) 2 ·s 2 ); F i，k — Represents the frictional resistance between the vapor and liquid phases, kg / (m) 2 ·s 2 ); — Represents the phase interface velocity, in m / s; M t，k — Represents the force caused by turbulent mixing of the vapor or liquid phases, kg / (m 2 ·s 2 ); —Represents the specific enthalpy of the vapor or liquid phase, in J / kg; Q w，k — Represents the heat transfer between the wall and the vapor or liquid phase, in W / m³ 3 ; Q i，k — Represents the heat transfer between the vapor and liquid phases, W / m 3 ; Q t，k — Represents heat transfer caused by turbulent mixing of the vapor or liquid phases, W / m 3 ; — Represents the specific enthalpy of the phase interface, J / kg.
[0007] The constitutive models for closing the basic control equations in step 2 include wall resistance model, interface resistance model, wall heat transfer model, interface heat transfer model, interface mass transfer model, and turbulent mixing model. The specific thermal conductivity model of the fuel rod in step 3 is as follows: (9) In the formula: — Represents the solid temperature, in K; — Represents the density of the solid, kg / m³ 3 ; — Represents the specific heat capacity of a solid, J / (kg·K); — Represents the thermal conductivity of a solid, W / (m·K); — Represents a solid internal heat source, W / m 3 .
[0008] In step 4, when discretizing the two-fluid six-equation model based on the finite volume discretization method, the convection term is discretized using the high-resolution van Leer scheme, which has the following specific form: (10) In the formula: — Represents the face-centered value of a physical quantity in the current element; — Represents the volume center value of a physical quantity in the current unit; — Represents the volume center value of a physical quantity in the downstream unit located in the current unit; — Represents the flux limiter, equal to ; — Represents the gradient ratio, equal to ; — Represents the volume center value of a certain physical quantity in the upstream unit located in the current unit.
[0009] The fluid-solid coupling solution process in step 5 is as follows: Step 1: Read the modeling parameters between sub-channels; Step 2: Set the initial and boundary conditions for the computational domain; Step 3: Solve the constitutive model and update the wall drag model, interface drag model, wall heat transfer model, interface heat transfer model, interface mass transfer model, and turbulent mixing model; Step 4: Solve the two-fluid six-equation model to obtain the phase fraction, velocity, specific enthalpy, and pressure of the vapor and liquid phases; Step 5: Solve the thermal conductivity model of the fuel rod to obtain the fuel rod temperature; Step 6: Determine whether the solution results of Step 3-5 meet the preset residual or maximum iteration step. If not, repeat Step 3-5 until the conditions are met; if they are met, proceed to the next step. Step 7: Output the calculation results of the two-phase thermal-hydraulic parameters of the reactor core; Step 8: Determine whether the set time step has been reached. If the set time step has not been reached, repeat steps 3-8 and use the calculation result in step 7 as the known quantity at the new time. If the set time step has been reached, the calculation ends.
[0010] Compared with the prior art, the present invention has the following advantages: 1. While maintaining a computational scale comparable to traditional subchannel methods, the three-dimensional geometry of the channel is adopted, which significantly improves the resolution of physical field distribution compared to traditional quasi-one-dimensional subchannel modeling methods. 2. A two-fluid six-equation model is adopted to avoid the assumption of phase equilibrium / local equilibrium and improve the simulation reliability of boiling and two-phase conversion processes; 3. Improve computational accuracy and stability by using high-resolution convection discretization methods, i.e., finite volume discretization methods; 4. Compared with refined CFD methods, this invention significantly reduces the number of computing units by using channel-level modeling, thereby reducing computation time and hardware resource requirements. Attached Figure Description
[0011] Figure 1 This is a flowchart of the two-phase thermal-hydraulic analysis method for nuclear reactor cores according to the present invention.
[0012] Figure 2 Subchannels are divided for individual components of the core. Detailed Implementation
[0013] This invention provides a two-phase thermal-hydraulic analysis method for nuclear reactor cores. Taking the 64-bar bundle assembly in the core as an example, the specific implementation of the method of this invention is described in detail.
[0014] like Figure 1 As shown, the two-phase thermal-hydraulic analysis method for a nuclear reactor core according to the present invention includes the following steps: Step 1: Based on the subchannel structure of the reactor core, determine the subchannel modeling parameters and fuel rod modeling parameters, establish a three-dimensional geometric model of the subchannel, and use mesh generation software to generate the computational mesh for the three-dimensional geometric model; Step 2: Based on the sub-channel modeling parameters determined in Step 1, establish the basic governing equations for the vapor and liquid phases based on the two-fluid six-equation model, and establish a constitutive model for the closed basic governing equations. Step 3: Based on the fuel rod modeling parameters determined in Step 1, establish a fuel rod thermal conductivity model; Step 4: Discretize the two-fluid six-equation model in Step 2 and the fuel rod heat conduction model established in Step 3 using the finite volume discretization method to obtain the fluid and solid discrete equation sets. Step 5: Based on the fluid and solid discrete equations obtained in Step 4, establish a fluid and solid coupled solution process.
[0015] The subchannel modeling parameters determined in Step 1 include: identification of various subchannel types, namely inner channels, side channels, and corner channels; flow area and hydraulic diameter of each type of subchannel; fuel rod size; grid spacing between fuel rods; grid spacing between fuel rods and component edges; and the 3D geometric model and computational mesh defined based on the subchannel structure, as shown below. Figure 2 As shown.
[0016] The two-fluid six-equation model in step 2 includes the two-phase continuity equation, the two-phase momentum equation, and the two-phase energy equation, specifically: (11) (12) (13) In the formula: k — Represents either the vapor phase or the liquid phase; t — Represents time, seconds; — Represents the phase fraction of the vapor or liquid phase; —Represents the density of the vapor or liquid phase, kg / m³ 3 ; — Represents the velocity of the vapor or liquid phase, in m / s; —Represents the mass transfer rate per unit volume of the vapor or liquid phase, kg / (m³) 3 ·s); — Represents pressure, MPa; F w,k — Represents the frictional resistance between the wall surface and the vapor or liquid phase, kg / (m 2 ·s 2 ); F b,k —Represents the gravity of the vapor or liquid phase, kg / (m³) 2 ·s 2 ); F i,k — Represents the frictional resistance between the vapor and liquid phases, kg / (m) 2 ·s 2 ); — Represents the phase interface velocity, in m / s; M t，k — Represents the force caused by turbulent mixing of the vapor or liquid phases, kg / (m 2 ·s 2 ); —Represents the specific enthalpy of the vapor or liquid phase, in J / kg; Q w，k — Represents the heat transfer between the wall and the vapor or liquid phase, in W / m³ 3 ; Q i，k — Represents the heat transfer between the vapor and liquid phases, W / m 3 ; Q t，k — Represents heat transfer caused by turbulent mixing of the vapor or liquid phases, W / m 3 ; — Represents the specific enthalpy of the phase interface, J / kg.
[0017] The constitutive models for closing the fundamental governing equations in step 2 include the wall drag model, the interface drag model, the wall heat transfer model, the interface heat transfer model, the interface mass transfer model, and the turbulent mixing model. For the liquid phase wall resistance model, the Madam relation is used: (14) In the formula, — Represents the two-phase multiplication factor; — Represents the Reynolds number in the vapor or liquid phase; — Represents the hydraulic diameter, in meters (m); For the wall heat transfer model, the DB relationship is given as follows: (15) In the formula, — Represents the thermal conductivity of the vapor or liquid phase, W / (m·K); —The Prandtl number, representing the vapor or liquid phase; — Represents the heat transfer area per unit volume of the wall, in meters. -1 ; — Represents the solid temperature, in K; — Represents the temperature of the vapor or liquid phase, in K; For boiling scenarios, the superposition method is used to handle the wall heat transfer model: (16) In the formula, The heat flux density of the convective component is expressed in W / m³. 2 ; Heat flux density representing the boiling portion of the pool, W / m 2 ; The heat flux density corresponding to the boiling initiation point, W / m³ 2 ; Furthermore, assuming that the liquid phase covers the wall surface and bears all wall friction and heat transfer under normal flow conditions, the wall friction model and wall heat transfer model of the vapor phase are both equal to 0, so the subscripts in equations (14) and (15) are... k It actually only acts on the liquid phase l.
[0018] The specific thermal conductivity model of the fuel rod in step 3 is as follows: (17) In the formula: — Represents the density of the solid, kg / m³ 3 ; — Represents the specific heat capacity of a solid, J / (kg·K); — Represents the thermal conductivity of a solid, W / (m·K); — Represents a solid internal heat source, W / m 3 .
[0019] The fluid-solid interface employs a mixed boundary condition, namely the inner wall of the tube and the outer wall of the fuel rod, to ensure heat flux conservation.
[0020] In step 4, when discretizing the two-fluid six-equation model using the finite volume discretization method, the high-resolution van Leer scheme is used to discretize the convection terms, which are in the following form: (19) In the formula: — Represents the face-centered value of a physical quantity in the current element; — Represents the volume center value of a physical quantity in the current unit; — Represents the volume center value of a physical quantity in the downstream unit located in the current unit; — Represents the flux limiter, equal to ; — Represents the gradient ratio, equal to ; — Represents the volume center value of a certain physical quantity in the upstream unit located in the current unit.
[0021] The solution process for fluid-solid coupling in step 5 is as follows: Step 1: Read the modeling parameters between sub-channels; Step 2: Set the initial and boundary conditions for the computational domain; Step 3: Solve the constitutive model and update the wall drag model, interface drag model, wall heat transfer model, interface heat transfer model, interface mass transfer model, and turbulent mixing model; Step 4: Solve the two-fluid six-equation model to obtain the phase fraction, velocity, specific enthalpy, and pressure of the vapor and liquid phases; Step 5: Solve the thermal conductivity model of the fuel rod to obtain the fuel rod temperature; Step 6: Determine whether the solution results of Step 3-5 meet the preset residual or maximum iteration step. If not, repeat Step 3-5 until the conditions are met; if they are met, proceed to the next step. Step 7: Output the calculation results of the two-phase thermal-hydraulic parameters of the reactor core; Step 8: Determine whether the set time step has been reached. If the set time step has not been reached, repeat steps 3-8 and use the calculation result in step 7 as the known quantity at the new time. If the set time step has been reached, the calculation ends.
Claims
1. A method for two-phase thermal-hydraulic analysis of a nuclear reactor core, characterized in that: Includes the following steps: Step 1: Based on the subchannel structure of the reactor core, determine the subchannel modeling parameters and fuel rod modeling parameters, establish a three-dimensional geometric model of the subchannel, and use mesh generation software to generate the computational mesh for the three-dimensional geometric model; Step 2: Based on the sub-channel modeling parameters determined in Step 1, establish the basic governing equations for the vapor and liquid phases based on the two-fluid six-equation model, and establish a constitutive model for the closed basic governing equations. Step 3: Based on the fuel rod modeling parameters determined in Step 1, establish a fuel rod thermal conductivity model; Step 4: Discretize the two-fluid six-equation model in Step 2 and the fuel rod heat conduction model established in Step 3 using the finite volume discretization method to obtain the fluid and solid discrete equation sets. Step 5: Based on the fluid and solid discrete equations obtained in Step 4, establish a fluid and solid coupled solution process.
2. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: The subchannel modeling parameters determined in step 1 include: identification of various types of subchannels, namely inner channels, side channels and corner channels; flow area and hydraulic diameter of each type of subchannel; fuel rod size; grid spacing between fuel rods; and grid spacing between fuel rods and component edges.
3. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: The two-fluid six-equation model in step 2 includes the two-phase continuity equation, the two-phase momentum equation, and the two-phase energy equation, specifically: (1) (2) (3) In the formula: k — Represents either the vapor phase or the liquid phase; t — Represents time, seconds; — Represents the phase fraction of the vapor or liquid phase; —Represents the density of the vapor or liquid phase, kg / m³ 3 ; — Represents the velocity of the vapor or liquid phase, in m / s; —Represents the mass transfer rate per unit volume of the vapor or liquid phase, kg / (m³) 3 ·s); — Represents pressure, MPa; F w，k — Represents the frictional resistance between the wall surface and the vapor or liquid phase, kg / (m 2 ·s 2 ); F b，k —Represents the gravity of the vapor or liquid phase, kg / (m³) 2 ·s 2 ); F i，k — Represents the frictional resistance between the vapor and liquid phases, kg / (m) 2 ·s 2 ); — Represents the phase interface velocity, in m / s; M t，k — Represents the force caused by turbulent mixing of the vapor or liquid phases, kg / (m 2 ·s 2 ); —Represents the specific enthalpy of the vapor or liquid phase, in J / kg; Q w，k — Represents the heat transfer between the wall and the vapor or liquid phase, in W / m³ 3 ; Q i，k — Represents the heat transfer between the vapor and liquid phases, in W / m³ 3 ; Q t，k — Represents heat transfer caused by turbulent mixing of the vapor or liquid phases, W / m 3 ; — Represents the specific enthalpy of the phase interface, J / kg.
4. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: The constitutive models for closing the basic governing equations in step 2 include wall resistance model, interface resistance model, wall heat transfer model, interface heat transfer model, interface mass transfer model, and turbulent mixing model.
5. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: The specific thermal conductivity model of the fuel rod in step 3 is as follows: (4) In the formula: — Represents the solid temperature, in K; — Represents the density of the solid, kg / m³ 3 ; — Represents the specific heat capacity of a solid, J / (kg·K); — Represents the thermal conductivity of a solid, W / (m·K); — Represents a solid internal heat source, W / m 3 .
6. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: In step 4, when discretizing the two-fluid six-equation model based on the finite volume discretization method, the convection term is discretized using the high-resolution van Leer scheme, which has the following specific form: (5) In the formula: — Represents the face-centered value of a physical quantity in the current element; — Represents the volume center value of a physical quantity in the current unit; — Represents the volume center value of a physical quantity in the downstream unit located in the current unit; — Represents the flux limiter, equal to ; — Represents the gradient ratio, equal to ; — Represents the volume center value of a certain physical quantity in the upstream unit located in the current unit.
7. The method for two-phase thermal-hydraulic analysis of a nuclear reactor core as described in claim 1, characterized in that: The fluid-solid coupling solution process established in step 5 is as follows: Step 1: Read the modeling parameters between sub-channels; Step 2: Set the initial and boundary conditions for the computational domain; Step 3: Solve the constitutive model and update the wall drag model, interface drag model, wall heat transfer model, interface heat transfer model, interface mass transfer model, and turbulent mixing model; Step 4: Solve the two-fluid six-equation model to obtain the phase fraction, velocity, specific enthalpy, and pressure of the vapor and liquid phases; Step 5: Solve the thermal conductivity model of the fuel rod to obtain the fuel rod temperature; Step 6: Determine whether the solution results of Step 3-5 meet the preset residual or maximum iteration step. If not, repeat Step 3-5 until the conditions are met; if they are met, proceed to the next step. Step 7: Output the calculation results of the two-phase thermal-hydraulic parameters of the reactor core; Step 8: Determine whether the set time step has been reached. If the set time step has not been reached, repeat steps 3-8 and use the calculation result in step 7 as the known quantity at the new time. The calculation ends when the set time step is reached.