Full-loop thermal hydraulic calculation method suitable for large nuclear island

Abstract

The invention discloses a full-loop thermal hydraulic calculation method suitable for a large nuclear island, and belongs to the technical field of computational fluid mechanics. The method comprises the following steps: firstly, determining geometric structures, operating parameters and material attributes of key components of the nuclear island; secondly, fine geometric models of the nuclear main pump, the evaporator and the reactor are established through modeling software, and the actual structural characteristics of all components are considered; thirdly, simplifying the structures of the evaporator and the reactor by adopting a porous medium theory, and retaining key operation characteristics; and then, all components in the nuclear island circulation loop are connected according to the actual operation state to form a closed loop so as to restore the actual working condition. In the calculation process, a source item pipeline method is used for replacing traditional boundary condition setting, and flow, pressure and temperature in a closed circulation loop are accurately regulated and controlled in combination with porous medium source item parameter setting. The core of the method is that the source item regulation and control and porous medium model method is comprehensively applied, and efficient calculation and analysis of the thermal hydraulic characteristics of the large nuclear island are achieved.

Description

Technical Field

[0001] The invention belongs to the field of computational fluid dynamics (CFD), and in particular relates to a full-loop thermal hydraulic calculation method suitable for a large nuclear island. Background Art

[0002] In today's situation of growing demand for clean energy, the demand for the development of nuclear power technology is extremely high. Ensuring the safety of energy infrastructure is crucial to the high-quality development of the energy system. The primary system of the nuclear island is responsible for transporting a large amount of heat released by the reactor and interacting with the power generation equipment through the evaporator, making it a key component of the nuclear power plant. The safe and efficient operation of the primary system is crucial. In the nuclear island system, the nuclear main pump is complexly connected with the pipeline, evaporator and reactor to form a closed circulation loop. Each component has unique characteristics while interacting with other components, which produces interference effects during operation. Most of the current research methods only focus on the reactor coolant pump itself, without considering the impact of the actual operating environment. This omission will cause some inaccuracies in the results. It is of great significance to analyze the temperature change law and pressure change law of the nuclear island primary loop system under different operating conditions through the overall CFD simulation calculation of the nuclear island circulation loop.

[0003] However, for the entire primary side system of the nuclear island, the closed circulation loop model lacks a corresponding solution. In addition, the circulation loop has a large number of components, a large scale, and extremely high computational costs. The present invention proposes a closed circulation loop calculation scheme based on source term pipeline control of fluid flow and combined with a simplified porous medium model to reduce computational costs, providing an effective way to analyze the circulation loop system of a large nuclear island. Summary of the invention

[0004] The purpose of the present invention is to realize the fluid-heat coupling thermal calculation of the full-loop system of a large nuclear island, and to study the actual operating characteristics of the nuclear island circulation loop. By reasonably simplifying the equipment of the nuclear island system and reducing the calculation difficulty, the current problem of overall structural simulation of the nuclear island circulation loop is solved.

[0005] The present invention provides a full-loop thermal hydraulic calculation method applicable to a large nuclear island, which is performed according to the following steps:

[0006] Step 1: Determine the key parameters of the full-loop calculation model of the target large-scale nuclear island, including the geometric parameters, speed, and design flow of the nuclear main pump; the geometric parameters, design flow, thermal power, operating pressure, primary side design temperature, secondary side design temperature, and material properties of the evaporator; the geometric parameters, fuel assembly parameters, thermal power, operating pressure, design temperature, and design flow of the nuclear reactor; the geometric parameters of the pipeline and the properties of the primary side cooling medium;

[0007] Step 2: Based on the key parameters obtained in step 1, a fine geometric model of the nuclear main pump considering the front and rear cavity assembly clearance is established through modeling software. The nuclear main pump is mainly divided into non-blade area (inlet section, outlet section, front and rear cavity clearance, water pressure chamber) and blade area (impeller, guide vane). Then, high-quality hexahedral meshes are used for meshing.

[0008] Step 3: Simplify the evaporator structure with large-scale heat exchange tube bundles, and establish a simplified geometric model of the evaporator without the actual tube bundle structure based on porous media theory. The overall fluid domain geometric model of the evaporator is divided into the inlet section, heat exchange section and outlet section. To ensure that the structural characteristics of the real evaporator are fully retained, the inlet section and outlet section structures are kept unchanged, and the heat exchange section is simplified to a porous media model; the simplified model is meshed using a tetrahedral mesh with a high degree of fitness, and edited using surface mesh processing technology to make the porous media domain consistent with the interface nodes of the inlet section and outlet section. The mesh quality is checked and the division is completed;

[0009] Step 4: Simplify the reactor structure with large-scale fuel assemblies, and establish a simplified geometric model of the reactor based on porous media theory. The reactor is mainly divided into four parts: the inlet section, the shroud area, the core and the outlet section. The main structure and the inlet and outlet of the heat exchange section are divided through the shroud area structure. The regularly arranged fuel rods inside the reactor affect the flow and heat transfer of the coolant. The fluid channels in the reactor core are simpler than the heat exchange tube bundle structure of the evaporator. The reactor core is simplified into a porous media model, and the reactor inlet and outlet sections and the shroud area are retained. The simplified model is meshed using a tetrahedral mesh with a higher adaptability, and edited using the surface mesh processing technology so that the porous medium domain is consistent with the inlet and outlet section interface nodes. The mesh quality is checked and the division is completed;

[0010] Step 5: Connect all nuclear main pumps, evaporators, reactors, and pipeline components in the nuclear island circulation loop according to the actual operating status, and set interfaces between the components according to the fluid domain type to form a closed loop, ensuring that the calculation conditions of all components of the nuclear island system restore the actual operating conditions;

[0011] Step 6: Based on the nuclear main pump geometric model and mesh constructed in step 2, calculate and set up multiple nuclear main pumps in the nuclear island circulation loop. There are no inlets and outlets in the closed circulation loop, and the inlet and outlet boundary conditions of each nuclear main pump cannot be set by conventional means. Therefore, the rotation center is set for the nuclear main pumps at different positions in the nuclear island circulation loop, and the multi-reference system model in the rotating coordinate system is used to process the impeller area, set the speed of the nuclear main pump, and set the interface between the rotating and stationary parts to the "frozen rotor" mode;

[0012] Step 7: Different from the numerical calculation of the open structure with artificially assigned inlet and outlet boundary conditions, the closed circulation loop lacks objective conditions for setting inlet and outlet boundary conditions, and cannot directly define the flow, pressure, temperature and other values ​​of each nuclear main pump and circulation branch of the nuclear island circulation loop. Therefore, source term pipelines are set between the reactor and the evaporator in the closed circulation pipeline to introduce external force source terms, and the flow and pressure operating conditions of the nuclear island are restored in combination with the reference pressure setting. A uniform external force is applied to the source term pipeline section to control the flow of the fluid, and the momentum equation is corrected to:

[0013]

[0014] Among them, ρ f is the fluid density; U is the flow velocity; p is the fluid pressure; t f is the shear force tensor; g is the acceleration due to gravity; k is a proportional constant;

[0015] Step 8: The evaporator and reactor are the key heat exchange components of the nuclear island circulation loop. They regulate the temperature and pressure changes of the entire nuclear island circulation loop. The inlet and outlet sections of the evaporator and reactor are set as fluid domains, and the heat exchange section is set as a porous medium domain. The corresponding interface types are set between different types of fluid domains. Secondly, momentum source terms and heat source terms are set for the porous medium domain, replacing the simplified geometric boundaries to regulate the flow and heat transfer process of the porous medium domain, solving the porous medium momentum source terms and heat source terms, and selecting the required heat transfer model. The decomposition process is as follows;

[0016] Step 8-1: The porosity is defined as: ε = V1 / V, where V1 is the volume of the fluid in the porous medium region, and V is the total volume of the porous medium region;

[0017] Step 8-2: Combine Darcy's law with the Erg's equation in porous media flow and consider the total pressure drop in porous media flow to be:

[0018]

[0019] Where ΔP is the pressure difference; Δn is the length of the porous medium flow domain; μ is the dynamic viscosity; ΔP v is the pressure difference caused by viscosity, ΔP i is the pressure difference caused by inertia; K per is the permeability, K loss is the inertial resistance coefficient; based on the velocity definition of the fluid in the porous medium domain, the specific value of the resistance coefficient is determined by fitting the quadratic polynomial relationship between the pressure drop and velocity of the target fluid domain.

[0020] △P=AU+BU 2

[0021] Among them, A and B are empirical coefficients;

[0022] Permeability: Drag coefficient: Step 8-3: Derived the heat source term of the evaporator according to the geometric structure parameters and physical property parameters obtained in step 1, obtained the heat source area in the porous medium through primary and secondary side heat transfer analysis and determined the heat source term therein:

[0023] Heat source term per unit volume:

[0024]

[0025] Among them, Nu z1 is the Nusselt number on the primary side of the evaporator; Re z1 is the Reynolds number on the primary side of the evaporator; Pr z1 is the Prandtl number on the primary side of the evaporator, which is 0.97; v actual is the actual flow rate of coolant; z1 is the thermal conductivity of the primary coolant; δ zwall is the tube wall thickness; zwall is the thermal conductivity of the tube wall; T zwall is the wall temperature of the heat exchange tube; T zsat is the saturation temperature on the boiling heat exchange side; P zsat is the saturated pressure on the boiling heat exchange side; q z A is the heat release power per unit volume; z is the total heat exchange area; V z is the volume of the porous medium region;

[0026] Different from the heat exchange method of the primary and secondary fluids of the evaporator through the tube wall, the reactor generates heat through nuclear fission to heat the coolant. The momentum source term is fitted and solved using the same method as the evaporator. When solving the heat source term, considering that the heat exchange of each part of the reactor core is relatively uniform and is less affected by the coolant flow, the average body heat source method is used to set the reactor heat source term:

[0027]

[0028] Among them, H in is the total enthalpy of the inlet fluid, H out is the total enthalpy of the outlet fluid, Q fyd is the reactor heat transfer power, V fyd is the total volume of the reactor core.

[0029] Step 9: The above parameter solution expressions are secondary developed through CEL expression language to obtain the thermodynamic process expression of the momentum source term and heat source term coupled with the fluid temperature and velocity.

[0030] Step 10: The steady Reynolds average method is used for numerical simulation. The NS equations and energy equations are discretized based on the finite volume method. The Thermal Energy model is selected as the heat transfer model, and the SST k-ω model is used as the turbulence model. With the help of CEL, the equations required for flow and heat transfer in the porous media domain are defined, and the convergence residual of the calculation results is set to 10 -7 .

[0031] The nuclear main pump modeling method described in step 2 takes into account the gaps and leakages existing in the actual model of the nuclear main pump, and includes these features when modeling and drawing the grid, so as to achieve refined modeling of the nuclear main pump, and the overall model of the nuclear main pump adopts hexahedral structured grids to ensure the accuracy of the calculation results from the model and grid levels. In step 3, the porous medium model is used to simplify the evaporator and reactor models. When simplifying, only the regularly distributed heat exchange tube bundles and cores are simplified, and their structural features are retained as much as possible while simplifying the calculation model. In the nuclear island circulation loop calculation model described in step 5, multiple nuclear main pumps, evaporators, and key reactor components contained in the nuclear island are considered, as well as the actual pipeline arrangement, to form a closed circulation loop. Different from the conventional open calculation model, the calculation model fully considers the mutual interference between components and is more in line with the actual operating state of the nuclear island primary loop system.

[0032] The closed loop flow control method described in step 7 is different from the open structure numerical calculation with artificially given inlet and outlet boundary conditions. The closed loop lacks the objective conditions for setting inlet and outlet boundary conditions. An external force source term is introduced into the pipeline, and the flow and pressure operating conditions of the nuclear island are restored in combination with the reference pressure setting. A uniform external force is applied to the source pipeline section to control the flow of the fluid, and the external force is expressed as F external =ρa external , where a external is the acceleration caused by the external force. The momentum equation in the momentum source pipeline becomes:

[0033] Momentum source term F external , which can be understood as a volume force, indicating the magnitude of the external force on a unit volume of fluid. Its value is obtained by calculating the magnitude of the external force on the fluid and the volume of the target source area, which can realize the demand for providing power for the closed loop. It is a linear function related to the velocity field U:

[0034]

[0035] In step 8, the porous medium coupled source term method is used to simplify the numerical calculation of the evaporator. The porous medium model is used to reduce the calculation cost. The following momentum source term and heat source term are added to the momentum equation and heat equation in the porous medium domain respectively to control the flow and heat transfer to ensure the calculation accuracy:

[0036]

[0037] Where ε is the porosity; K = K ij is a second-order symmetric tensor, is the area porosity tensor; μ e is the effective viscosity; h hot is the enthalpy of the fluid; e is the effective thermal conductivity, S M is the momentum source term, S E is the heat source term (energy source term).

[0038] The present invention has the following beneficial effects:

[0039] 1) The present invention is based on a large-scale nuclear island full-loop calculation method, which connects all key components of the nuclear island system in series to form a closed-loop calculation model. By introducing momentum source term pipelines to control flow characteristics such as flow and pressure in the circulation loop, the integrated thermal-hydraulic calculation of the nuclear island full loop is realized, breaking the limitations of traditional research on the characteristics of a single component.

[0040] 2) The full-loop thermal hydraulic calculation method for a large nuclear island of the present invention reasonably simplifies the heat exchange area of ​​the evaporator and reactor heat exchange components according to their structure and performance characteristics, and reduces the difficulty and cost of the full-loop calculation of a large nuclear island while retaining as many structural features as possible. In addition, by adding momentum source terms and heat source terms to the porous medium domain to regulate the flow and heat exchange in the components, the influence of geometric structure simplification is compensated to ensure the calculation accuracy.

[0041] In the calculation process, the source term pipeline method is used to replace the traditional boundary condition setting, combined with the porous medium source term parameter setting to accurately control the flow, pressure and temperature in the closed loop. The core of the invention is the comprehensive application of source term control and porous medium model method to achieve efficient calculation and analysis of the thermal hydraulic characteristics of large nuclear islands. BRIEF DESCRIPTION OF THE DRAWINGS

[0042] Figure 1 It is a complete flow chart of the present invention.

[0043] Figure 2 This is a schematic diagram of the structural disassembly of the CAP1400 nuclear main pump.

[0044] Figure 3 This is a schematic diagram of the distribution and flow of the nuclear main pump, evaporator, and reactor in the primary loop system of the CAP1400 nuclear island.

[0045] Figure 4 This is a simplified front and back cross-sectional comparison diagram of the CAP1400 nuclear island evaporator.

[0046] Figure 5 This is a schematic diagram comparing the CAP1400 nuclear island reactor core before and after simplification.

[0047] Figure 6 Schematic diagram of the calculation model of the primary loop system of the CAP1400 nuclear complete island.

[0048] Figure 7 The figure is a comparison curve between the external characteristics of the nuclear main pump calculated by the method of the present invention and the actual operating values ​​of the CAP1400 nuclear island circulation loop.

[0049] In the figure: 1. nuclear main pump inlet section, 2. nuclear main pump impeller, 3. nuclear main pump gap, 4. nuclear main pump guide vane, 5. nuclear main pump water pressure chamber, 6. first nuclear main pump, 7. first evaporator, 8. second nuclear main pump, 9. nuclear reactor, 10. third nuclear main pump, 11. third evaporator, 12. fourth nuclear main pump, 13. evaporator outlet section, 14. evaporator inlet section, 15. evaporator heat exchange tube bundle, 16. evaporator porous medium domain, 17. evaporator 1, 18. first source item pipeline, 19. second source item pipeline. DETAILED DESCRIPTION

[0050] Combine the following Figure 1 The present invention is further described in detail with reference to the flowchart in FIG.

[0051] The present invention provides a full-loop thermal hydraulic calculation method applicable to a large nuclear island, which is performed according to the following steps:

[0052] Step 1: Determine the key parameters of the full-loop calculation model of the target large-scale nuclear island, including the geometric parameters, speed, and design flow of the nuclear main pump; the geometric parameters, design flow, thermal power, operating pressure, primary side design temperature, secondary side design temperature, and material properties of the evaporator; the geometric parameters, fuel assembly parameters, thermal power, operating pressure, design temperature, and design flow of the nuclear reactor; the geometric parameters of the pipeline and the properties of the primary side cooling medium; the distribution and flow schematic diagram of the nuclear main pump, evaporator, and reactor of the primary loop system of the CAP1400 nuclear island is shown in the figure below: Figure 3 As shown in the figure, the nuclear reactor 9 includes a first nuclear main pump 6, a first evaporator 7 and a second nuclear main pump 8 on one side, and a third nuclear main pump 10, a second evaporator 11 and a fourth nuclear main pump 12 on the other side. The structure of the nuclear main pump is as shown in FIG. Figure 2 As shown, it includes a nuclear main pump inlet section 1, a nuclear main pump impeller 2, a nuclear main pump gap 3, a nuclear main pump guide vane 4 and a nuclear main pump water pressure chamber.

[0053] Step 2: Based on the key parameters obtained in step 1, a fine geometric model of the nuclear main pump considering the front and rear cavity assembly clearance is established through modeling software. The nuclear main pump is mainly divided into non-blade area (inlet section, outlet section, front and rear cavity clearance, water pressure chamber) and blade area (impeller, guide vane). Then, high-quality hexahedral meshes are used for meshing.

[0054] Step 3: Simplify the evaporator structure with large-scale heat exchange tube bundles, and establish a simplified geometric model of the evaporator without the actual tube bundle structure based on porous media theory. The overall fluid domain geometric model of the evaporator is divided into the inlet section, heat exchange section and outlet section. To ensure that the structural characteristics of the real evaporator are fully retained, the inlet section and outlet section structures are kept unchanged, and the heat exchange section is simplified to a porous media model; the simplified model is meshed using a tetrahedral mesh with a high degree of fitness, and edited using surface mesh processing technology to make the porous media domain consistent with the interface nodes of the inlet section and outlet section. The mesh quality is checked and the division is completed;

[0055] Step 4: Simplify the reactor structure with large-scale fuel assemblies, and establish a simplified geometric model of the reactor based on porous media theory. The reactor is mainly divided into four parts: the inlet section, the shroud area, the core and the outlet section. The main structure and the inlet and outlet of the heat exchange section are divided through the shroud area structure. The regularly arranged fuel rods inside the reactor affect the flow and heat transfer of the coolant. The fluid channels in the reactor core are simpler than the heat exchange tube bundle structure of the evaporator. The reactor core is simplified into a porous media model, and the reactor inlet and outlet sections and the shroud area are retained. The simplified model is meshed using a tetrahedral mesh with a higher adaptability, and edited using the surface mesh processing technology so that the porous medium domain is consistent with the inlet and outlet section interface nodes. The mesh quality is checked and the division is completed;

[0056] Step 5: Connect all nuclear main pumps, evaporators, reactors, and pipeline components in the nuclear island circulation loop according to the actual operating status, and set interfaces between the components according to the fluid domain type to form a closed loop, ensuring that the calculation conditions of all components of the nuclear island system restore the actual operating conditions;

[0057] Step 6: Based on the nuclear main pump geometric model and mesh constructed in step 2, calculate and set up multiple nuclear main pumps in the nuclear island circulation loop. There are no inlets and outlets in the closed circulation loop, and the inlet and outlet boundary conditions of each nuclear main pump cannot be set by conventional means. Therefore, the rotation center is set for the nuclear main pumps at different positions in the nuclear island circulation loop, and the multi-reference system model in the rotating coordinate system is used to process the impeller area, set the speed of the nuclear main pump, and set the interface between the rotating and stationary parts to the "frozen rotor" mode;

[0058] Step 7: Different from the numerical calculation of the open structure with artificially assigned inlet and outlet boundary conditions, the closed circulation loop lacks objective conditions for setting inlet and outlet boundary conditions, and cannot directly define the flow, pressure, temperature and other values ​​of each nuclear main pump and circulation branch of the nuclear island circulation loop. Therefore, source term pipelines are set between the reactor and the evaporator in the closed circulation pipeline to introduce external force source terms, and the flow and pressure operating conditions of the nuclear island are restored in combination with the reference pressure setting. A uniform external force is applied to the source term pipeline section to control the flow of the fluid, and the momentum equation is corrected to:

[0059]

[0060] Among them, ρ f is the fluid density; U is the flow velocity; p is the fluid pressure; t f is the shear force tensor; g is the acceleration due to gravity; k is a proportional constant;

[0061] Step 8: The evaporator and reactor are the key heat exchange components of the nuclear island circulation loop. They regulate the temperature and pressure changes of the entire nuclear island circulation loop. The inlet and outlet sections of the evaporator and reactor are set as fluid domains, and the heat exchange section is set as a porous medium domain. The corresponding interface types are set between different types of fluid domains. Secondly, momentum source terms and heat source terms are set for the porous medium domain, replacing the simplified geometric boundaries to regulate the flow and heat transfer process of the porous medium domain, solving the porous medium momentum source terms and heat source terms, and selecting the required heat transfer model. The decomposition process is as follows;

[0062] Step 8-1: The porosity is defined as: ε = V1 / V, where V1 is the volume of the fluid in the porous medium region, and V is the total volume of the porous medium region;

[0063] Step 8-2: Combine Darcy's law with the Erg's equation in porous media flow and consider the total pressure drop in porous media flow to be:

[0064]

[0065] Where ΔP is the pressure difference; Δn is the length of the porous medium flow domain; μ is the dynamic viscosity; ΔP v is the pressure difference caused by viscosity, ΔP i is the pressure difference caused by inertia; K per is the permeability, K loss is the inertial resistance coefficient; based on the velocity definition of the fluid in the porous medium domain, the specific value of the resistance coefficient is determined by fitting the quadratic polynomial relationship between the pressure drop and velocity of the target fluid domain.

[0066] △P=AU+BU 2

[0067] Among them, A and B are empirical coefficients;

[0068] Permeability: Drag coefficient: Step 8-3: Derive the evaporator energy source term based on the geometric structure parameters and physical property parameters obtained in step 1, obtain the heat source area in the porous medium through primary and secondary side heat transfer analysis and determine the energy source term:

[0069] Heat source term per unit volume:

[0070]

[0071] Among them, Nu z1 is the Nusselt number on the primary side of the evaporator; Re z1 is the Reynolds number on the primary side of the evaporator; Pr z1 is the Prandtl number on the primary side of the evaporator, which is 0.97; v actual is the actual flow rate of coolant; z1 is the thermal conductivity of the primary coolant; δ zwall is the tube wall thickness; zwall is the thermal conductivity of the tube wall; T zwall is the wall temperature of the heat exchange tube; T zsat is the saturation temperature on the boiling heat exchange side; P zsat is the saturated pressure on the boiling heat exchange side; q z A is the heat release power per unit volume; z is the total heat exchange area; V z is the volume of the porous medium region;

[0072] Different from the heat exchange method of the primary and secondary fluids of the evaporator through the tube wall, the reactor generates heat through nuclear fission to heat the coolant. The momentum source term is fitted and solved using the same method as the evaporator. When solving the heat source term, considering that the heat exchange of each part of the reactor core is relatively uniform and is less affected by the coolant flow, the average body heat source method is used to set the reactor heat source term:

[0073]

[0074] Among them, H in is the total enthalpy of the inlet fluid, H out is the total enthalpy of the outlet fluid, Q fyd is the reactor heat transfer power, V fyd is the total volume of the reactor core.

[0075] Step 9: The above parameter solution expressions are secondary developed through CEL expression language to obtain the thermodynamic process expression of the momentum source term and heat source term coupled with the fluid temperature and velocity.

[0076] Step 10: The steady Reynolds average method is used for numerical simulation. The NS equations and energy equations are discretized based on the finite volume method. The Thermal Energy model is selected as the heat transfer model, and the SST k-ω model is used as the turbulence model. With the help of CEL, the equations required for flow and heat transfer in the porous media domain are defined, and the convergence residual of the calculation results is set to 10 -7 .

[0077] Example 1

[0078] The following is a clear and complete description of the technical solutions in the implementation cases of the present invention. The examples described below are merely design condition cases in the calculation process.

[0079] The present invention provides an overall simplified calculation method based on CFX simulation of CAP1400 large nuclear island circulation loop, combined with Figure 1 The present invention is further explained by the flowchart of the present invention, which includes the following specific steps:

[0080] Step 1: Determine the key parameters of the full-loop calculation model of the target large-scale nuclear island, including the geometric parameters, speed, and design flow of the nuclear main pump; the geometric parameters, design flow, thermal power, operating pressure, primary side design temperature, secondary side design temperature, and material properties of the evaporator; the geometric parameters, fuel assembly parameters, thermal power, operating pressure, design temperature, and design flow of the nuclear reactor; the geometric parameters of the pipeline and the properties of the primary side cooling medium;

[0081] Step 2: Establish a refined model of the nuclear main pump in the 3D modeling software UG. As the core component of the nuclear island, the nuclear main pump has a complex internal structure. In the process of computational fluid dynamics simulation, it is crucial to capture the fluid details with high precision, such as boundary layer flow and secondary flow in the gap, which will significantly affect its performance. In order to ensure accurate numerical calculations, the original pump design remains intact and is not simplified. Figure 3 This is a schematic diagram of the overall fluid domain model, including the disassembly diagram of the five main parts of the inlet section, gap, impeller, guide vane, and pressurized water chamber. The entire nuclear main pump adopts hexahedral structured grids. After grid independence verification, the number of grids in a single nuclear main pump is 4.12 million.

[0082] Step 3: There are 12606 U-shaped heat exchanger tubes in the actual CAP1400 single evaporator. Modeling the performance of the evaporator using the actual tube structure requires a large number of grid points, resulting in a large number of resource requirements and considerable time consumption. Import the established evaporator component geometry model into the 3D modeling software, and perform porous media simplification on the primary side tube bundle heat exchange part. Compare the models before and after simplification. Figure 4As shown; Step 4: Similarly, for the reactor equipment, there are 193 fuel assemblies, each with 17 by 17 fuel rods. The number of grids required for direct CFD calculation is actually incalculable. Import the established reactor component geometry model into the 3D modeling software, simplify the core part, and compare the models before and after simplification. Figure 5 shown.

[0083] Mesh the evaporator and reactor, and import the mesh into CFX; edit the mesh so that the interface mesh nodes are consistent, and import it into ICEM for mesh quality detection. It can be seen that the mesh quality is above 0.2 and the average quality is above 0.8. Step 5: Connect all nuclear main pumps, evaporators, reactors, and piping components in the nuclear island circulation loop according to the actual operating status, and set interfaces between the components according to the fluid domain type to form a closed loop. Step 6: Use ANSYS CFX software for numerical solution. The CAP1400 nuclear island circulation loop contains four nuclear main pumps, and the distribution is as follows: Figure 6 As shown: the nuclear reactor 9 is connected to the first nuclear main pump 6 and the second nuclear main pump 8 through the first source pipeline 18, and then to the first evaporator 7; it is connected to the third nuclear main pump 10 and the fourth nuclear main pump 12 through the second source pipeline 19, and then to the second evaporator 11. The main design parameters include working pressure 15.5MPa, design flow rate 21642m 3 / h, synchronous speed 1485r / min, operating temperature 283.4℃. The four nuclear main pumps take the center of the impeller at their location as the rotation axis, use the multi-reference frame model (MRF) in the rotating coordinate system to process the impeller area, and the interface between the rotating impeller and the stationary parts is set as the "frozen rotor method";

[0084] Step 7: Unlike the open structure, the closed pipeline lacks the objective conditions for setting the inlet and outlet boundary conditions. Therefore, a momentum source area is added between the evaporator and the nuclear main pump. After calculation, the momentum source item of the source pipeline is set to 217000kg / (m 2 ·s 2 ).

[0085] Step 8: Solve the porous media and momentum source parameters in the evaporator and reactor. In the porous media model, the porosity is expressed as the ratio of the volume of the simplified porous media domain in its original, unsimplified target area to the volume of the corresponding simplified porous media area. It is defined as ε = V1 / V. The important parameter values ​​of the porous media model can be calculated, and the specific porosity values ​​are shown in Table 1.

[0086] The other two important parameters, viscosity loss coefficient and inertia loss coefficient, represent the pressure drop of the research object, that is, the momentum loss of the research object. The viscosity loss coefficient is closely related to the permeability parameter, while the inertia loss coefficient is closely related to the local fluid velocity. According to the above process derivation of the viscosity loss coefficient and permeability calculation, the evaporator and reactor characteristic units are selected for pressure drop-velocity fitting and solution, and the results are shown in Table 1;

[0087] Table 1

[0088]

[0089] Solve the energy source terms in the porous media domains of the evaporator and reactor. The single-phase high-temperature and high-pressure coolant heated by the reactor flows inside the evaporator tube bundle, and the gas-liquid two-phase fluid with phase change flows outside the tube. The heat transfer process on the primary and secondary sides of the steam generator heat transfer tube includes forced convection heat transfer of the primary circuit coolant to the tube wall, heat conduction through the tube wall and the dirt layer, and boiling heat transfer of the heat transfer surface tube wall to the secondary circuit working fluid.

[0090] Through thermal resistance analysis, it can be known that the heat transfer process contains three parts: the primary side convection heat transfer resistance, the tube wall heat conduction resistance and the secondary side convection heat transfer resistance. After checking, it can be known that the inner diameter of the tube d z1 0.01647m; the heat transfer tube material is INCONEL690, and its thermal conductivity is zwall is 13.5W / (m·K); tube wall thickness δ zwall It is 0.00101m.

[0091] After derivation, the heat source term of the porous medium domain of the CAP1400 nuclear island evaporator is:

[0092]

[0093] The heat source of the evaporator heat transfer adopts the unit volume heat source method. Different from the heat transfer method of the primary and secondary fluids of the evaporator through the tube wall, the reactor generates heat through nuclear fission to heat the coolant. Considering the difficulty of heat transfer analysis of various parts of the reactor core, the average volume heat source method is used to set the reactor heat source source term. According to the thermal power of the CAP1400 nuclear island reactor and the geometric model of the reactor porous medium domain, the average volume heat source of the reactor is 9.083×10 7 W / m 3 .

[0094]

[0095] Step 9: The complete nuclear island closed cycle loop calculation model consisting of all components is as follows Figure 6, all parameter solution processes are written into the ANSYS simulation through CEL coding, the porous medium source term intensity is coupled with the temperature and velocity of the primary side fluid, and numerical calculations are performed to solve it.

[0096] Step 10: The steady Reynolds average method is used for numerical simulation. The NS equations and energy equations are discretized based on the finite volume method. The Thermal Energy model is selected as the heat transfer model, and the SST k-ω model is used as the turbulence model. With the help of CEL, the equations required for flow and heat transfer in the porous media domain are defined, and the convergence residual of the calculation results is set to 10 -7 ;

[0097] After the calculation converges, the external characteristics of the core main pump calculated by the method of the present invention are first compared with the actual operating values. The comparison results are as follows: Figure 7 As shown, compared with the experimental results, the numerical errors of head and efficiency are within 2.7% and 2.6% respectively, which fully meets the requirements of GB / T 3216 and calculation accuracy.

[0098] Secondly, compared Figure 6 The simulation results and actual operating values ​​of the inlet and outlet interface temperatures of each key component S1~S5 are compared as shown in Table 2.

[0099] Table 2

[0100]

[0101] The maximum temperature difference between the calculated value and the running value of each key section is 3.8°C, and the simulation error is kept below 1.5%. These results provide evidence to support the accuracy of the method of the present invention used in closed simulation.

[0102] The simulation results reliably reflect the operating data of the nuclear island circulation loop system, verifying the reliability of the results of the present invention.

Claims

1. A full-circuit thermal-hydraulic calculation method applicable to a large nuclear island, characterized in that: The following steps are involved: Step 1: Determine the key parameters of the full-loop calculation model of the target large-scale nuclear island, including the geometric structure parameters, speed, and design flow of the nuclear main pump; Evaporator geometry parameters, design flow, thermal power, operating pressure, primary design temperature, secondary design temperature, and material properties; Nuclear reactor geometry parameters, fuel assembly parameters, thermal power, operating pressure, design temperature, and design flow rate; Pipeline geometry parameters, primary side cooling medium properties; Step 2: Based on the key parameters obtained in step 1, a fine geometric model of the nuclear main pump considering the assembly clearance of the front and rear cavities is established through modeling software, and the nuclear main pump is divided into a non-blade area and a blade area, and then a hexahedral grid is used for grid division; The non-blade area includes an inlet section, an outlet section, a front and rear cavity gap, and a pressurized water chamber, and the blade area includes an impeller and a guide vane; Step 3: Based on the porous media theory, a simplified geometric model of the evaporator without the actual tube bundle structure is established, wherein the overall fluid domain geometric model of the evaporator is divided into an inlet section, a heat exchange section, and an outlet section. The structures of the inlet section and the outlet section are kept unchanged, and the heat exchange section is simplified to a porous media model; the simplified model is meshed using tetrahedral meshes, and edited using surface mesh processing technology to make the porous media domain consistent with the interface nodes of the inlet section and the outlet section. The mesh quality is checked and the meshing is completed; Step 4: Based on the porous media theory, a simplified geometric model of the reactor is established. The reactor is divided into four parts: the inlet section, the surrounding tube area, the core and the outlet section. The main structure of the heat exchange section and the inlet and outlet are divided through the surrounding tube area structure; the reactor core is simplified into a porous media model, the reactor inlet and outlet sections and the surrounding tube area are retained, and the simplified model is meshed using tetrahedral meshes. The surface mesh processing technology is used to edit the model so that the porous media domain is consistent with the inlet and outlet section interface nodes. The mesh quality is checked and the meshing is completed. Step 5: Connect all nuclear main pumps, evaporators, reactors, and pipeline components in the nuclear island circulation loop according to the actual operating status, and set interfaces between the components according to the fluid domain type to form a closed loop, ensuring that the calculation conditions of all components of the nuclear island system restore the actual operating conditions; Step 6: Based on the nuclear main pump geometric model and mesh constructed in step 2, calculate and set up multiple nuclear main pumps in the nuclear island circulation loop respectively; set the rotation center of the nuclear main pumps at different positions in the nuclear island circulation loop respectively, use the multi-reference system model in the rotating coordinate system to process the impeller area, set the speed of the nuclear main pump, and set the interface between the rotating and stationary parts to the "frozen rotor" mode; Step 7: Set source term pipelines between the reactor and the evaporator in the closed circulation pipeline to introduce external force source terms, and restore the flow and pressure operating conditions of the nuclear island in combination with the reference pressure setting; apply a uniform external force to the source term pipeline section to control the flow of the fluid, and the momentum equation is corrected to: Among them, ρ f is the fluid density; U is the flow velocity; p is the fluid pressure; τ f is the shear force tensor; g is the acceleration due to gravity; k is the proportionality constant; Step 8: The evaporator and reactor are the key heat exchange components of the nuclear island circulation loop. The temperature and pressure changes of the entire nuclear island circulation loop are regulated. The inlet and outlet sections of the evaporator and reactor are set as fluid domains, and the heat exchange section is set as a porous medium domain. The corresponding interface types are set between different types of fluid domains; secondly, the momentum source term and heat source term are set for the porous medium domain, and the simplified geometric boundaries are replaced to regulate the flow and heat transfer process of the porous medium domain, solve the porous medium momentum source term and heat source term, and select the required heat transfer model. The decomposition process is as follows: Step 8-1: The porosity is defined as: ε = V1 / V, where V1 is the volume of the fluid in the porous medium region, and V is the total volume of the porous medium region; Step 8-2: Combine Darcy's law with the Erg's equation in porous media flow and consider the total pressure drop in porous media flow to be: Where ΔP is the pressure difference; Δn is the length of the porous medium flow domain; μ is the dynamic viscosity; ΔP v is the pressure difference caused by viscosity, ΔP i is the pressure difference caused by inertia; K per is the permeability, K loss is the inertial drag coefficient; Based on the definition of fluid velocity in the porous medium domain, the specific value of the resistance coefficient is determined by fitting the quadratic polynomial relationship between the pressure drop and velocity in the target fluid domain; ΔP=AU+BU 2 Among them, A and B are empirical coefficients; permeability: Drag coefficient: Step 8-3: Derived the heat source term of the evaporator according to the geometric structure parameters and physical property parameters obtained in step 1, obtained the heat source area in the porous medium through primary and secondary side heat transfer analysis and determined the heat source term therein: Heat source term per unit volume: Among them, Nu z1 is the Nusselt number on the primary side of the evaporator; Re z1 is the Reynolds number on the primary side of the evaporator; Pr z1 is the Prandtl number on the primary side of the evaporator; v actual is the actual flow rate of coolant; z1 is the thermal conductivity of the primary coolant; δ zwall is the tube wall thickness; zwall is the thermal conductivity of the tube wall; T zwall is the wall temperature of the heat exchange tube; T zsat is the saturation temperature on the boiling heat exchange side; P zsat is the saturated pressure on the boiling heat exchange side; q z A is the heat release power per unit volume; z is the total heat exchange area; V z is the volume of the porous medium region; The momentum source term is fitted and solved using the same method as the evaporator. When solving the heat source term, the average body heat source method is used to set the reactor heat source term: Among them, H in is the total enthalpy of the inlet fluid, H out is the total enthalpy of the outlet fluid, Q fyd is the reactor heat transfer power, V fyd is the total volume of the reactor core; Step 9: The above parameter solution expressions are secondary developed through CEL expression language to obtain the thermodynamic process expression of the momentum source term and heat source term coupled with the fluid temperature and velocity; Step 10: The steady Reynolds average method is used for numerical simulation. The NS equations and energy equations are discretized based on the finite volume method. The Thermal Energy model is selected as the heat transfer model, and the SST k-ω model is used as the turbulence model. With the help of CEL, the equations required for flow and heat transfer in the porous media domain are defined, and the convergence residual of the calculation results is set to 10 -7 .

2. The method according to claim 1, characterized in that In step 7, a uniform external force is applied to the source pipeline segment, and the external force is represented by F external =ρa external , where a external is the acceleration caused by the external force; the momentum equation in the momentum source pipeline becomes: Momentum source term F external , which represents the magnitude of the external force acting on a unit volume of fluid.

3. The method according to claim 1, characterized in that In step 8, the porous medium coupled source term method is used to simplify the numerical calculation of the evaporator. The momentum equation after adding the momentum source term in the porous medium domain is: The energy equation after adding the heat source term is: Where ε is the porosity; K = K ij is a second-order symmetric tensor, is the area porosity tensor; μ e is the effective viscosity; h hot is the enthalpy of the fluid; e is the effective thermal conductivity, S M is the momentum source term, S E is the heat source term.

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