Numerical simulation calculation method and device for thermal characteristics of reactor core of heat pipe reactor and medium

Abstract

The invention discloses a heat pipe reactor core nuclear thermal characteristic numerical simulation method, electronic equipment and a storage medium, and the heat pipe reactor core nuclear thermal characteristic numerical simulation method comprises the steps: building a neutron physical model according to a reactor core geometric configuration and material parameters given by a heat pipe reactor design scheme; obtaining three-dimensional power distribution of the heat pipe reactor core under the critical working condition; carrying out normalization processing on the output power distribution counting result, and converting the result into a fuel rod axial heat source distribution curve through a polynomial fitting method; and sequentially obtaining heat pipe temperature distribution, reactor core three-dimensional temperature distribution and reactor core three-dimensional thermal stress strain distribution. According to the method, nuclear thermal characteristic calculation software is connected through an external coupling method, thermal characteristic calculation is achieved through an internal coupling mode, and nuclear thermal characteristic coupling calculation of the reactor core of the heat pipe reactor can be achieved efficiently and rapidly. According to the method, the advantages of each piece of calculation software are fully exerted, the reliability of numerical analysis is ensured, and the efficiency of reactor core design work is improved.

Description

Technical Field

[0001] The present invention relates to the field of nuclear reactor engineering analysis and design, and in particular to a numerical simulation calculation method for the core thermal characteristics of a heat pipe cooled nuclear reactor (referred to as a heat pipe reactor). Background Art

[0002] A heat pipe reactor (HPRR) utilizes heat pipes instead of traditional primary circuits to passively transfer heat generated by the reactor core directly to a secondary circuit system or thermoelectric converter. This system offers advantages such as high heat transfer efficiency, inherent safety, avoidance of single-point failures, and ease of modularization and expansion. It has significant advantages in scenarios requiring low power of 5MWe or less, operating in isolated grids with limited personnel, and requiring high long-term reliability. It also holds broad application prospects in deep space exploration, satellite surface power stations, land-based mobile power sources, and deep-sea exploration.

[0003] The most important feature that distinguishes heat pipe reactors from traditional cores is their solid-state properties. Therefore, during the design and development of heat pipe reactors, in addition to analyzing the core's neutron physics and thermal properties, it is also important to focus on studying the core's structural mechanical properties. Nuclear thermal characteristics analysis, as a core step in the design and development of nuclear reactor types, provides data support for the evaluation of the core's inherent safety characteristics and a reference for core design optimization. For the design and development of reactor types with different power requirements and application scenarios, numerical simulation calculations can be used to determine the preliminary core design scheme at low cost and high efficiency.

[0004] Heat pipe reactors are currently in the exploratory development phase, with various design concepts proposed for different requirements. The geometry and design requirements of heat pipe reactors differ significantly from those of traditional pressurized water reactors. Currently, most existing models and methods are developed for traditional pressurized water reactors and are not suitable for the computational analysis of heat pipe reactors. Designing and developing heat pipe reactor cores for different power requirements and application scenarios is computationally cumbersome and time-consuming. Therefore, a method for rapid and accurate numerical simulation of nuclear thermal characteristics is needed during the development phase of heat pipe reactor core design. Summary of the Invention

[0005] Technical Problem: A numerical simulation method for the nuclear thermal characteristics of a heat pipe reactor core, electronic equipment, and storage media are proposed to enable efficient and reliable core design and development.

[0006] Technical solution:

[0007] The present invention provides a method for numerical simulation calculation of the thermal characteristics of a heat pipe reactor core, comprising the following steps:

[0008] Step 1: Based on the core geometry and material parameters given in the heat pipe reactor design, a neutron physics model is established. Input parameters such as the neutron source term and statistical card are set. The neutron transport equation is solved using the Monte Carlo method to obtain the three-dimensional power distribution of the heat pipe reactor core under critical conditions.

[0009] Step 2: Determine the credibility of the neutron transport calculation results. If the effective neutron multiplication coefficient convergence, statistical error and other parameters meet the preset standards, the model is considered credible.

[0010] Step 3: Establish heat pipe heat transfer models and reactor core heat transfer models in the finite element calculation platform, set material thermophysical parameters and boundary conditions, and perform meshing and independence verification on the models.

[0011] Step 4: Normalize the power distribution count results output by the neutron transport calculation in Step 1 and convert them into the fuel rod axial heat source distribution curve by fitting a polynomial method;

[0012] Step 5: Based on the power distribution obtained in step 4, the heat flux of the evaporation section of the heat pipe is obtained. The heat transfer model of the heat pipe is imported as the heat source boundary condition. The wall of the adiabatic section is set to adiabatic, and the condensing section is set to be coupled with the heat exchange system boundary condition to obtain the heat pipe temperature distribution.

[0013] Step 6: Using command streams, the fuel rod power distribution polynomial formula obtained in Step 4 is imported into the core heat transfer model as a heat source condition. The heat pipe evaporator wall temperature distribution obtained in Step 5 is imported into the core heat transfer model as a cold source condition. Core thermal calculations are performed to obtain the three-dimensional core temperature distribution.

[0014] Step 7: Establish a structural mechanics model in the finite element calculation platform and add material structural mechanics parameters and constraints, such as thermal expansion coefficient, Young's modulus, Poisson's ratio and other parameters;

[0015] Step 8: Based on the finite element calculation platform, the three-dimensional core temperature distribution obtained in step 6 is loaded into the core structural mechanics model through the internal coupling method, and the core structural mechanics calculation is performed to obtain the three-dimensional thermal stress and strain distribution of the core;

[0016] Step 9: Export and analyze the nuclear thermal coupling calculation results as the basis for subsequent core thermal safety characteristics, structural safety characteristics and design optimization, and complete the nuclear thermal characteristics analysis of the heat pipe reactor core.

[0017] The present invention further provides an electronic device, comprising:

[0018] one or more processors;

[0019] a memory for storing one or more programs;

[0020] When the one or more programs are executed by the one or more processors, the one or more processors implement the steps of the method for numerical simulation calculation of the thermal characteristics of the heat pipe reactor core.

[0021] The present invention also provides a storage medium storing a computer program, which, when executed by a processor, implements the steps of the numerical simulation calculation method for the thermal characteristics of the core of a heat pipe reactor.

[0022] Compared with the prior art, the beneficial effects of the invention are:

[0023] 1. This invention utilizes an external coupling method to connect to nuclear thermal characteristics calculation software and implements thermodynamic characteristics calculation through internal coupling. This method can efficiently and quickly perform coupled calculations of nuclear thermodynamic characteristics for heat pipe reactor cores. This method fully leverages the advantages of each calculation software, ensures the reliability of numerical analysis, and improves the efficiency of core design.

[0024] 2. The present invention can obtain the three-dimensional power distribution, three-dimensional temperature field distribution and three-dimensional structural mechanics analysis results of the heat pipe reactor core, thereby being able to more comprehensively evaluate the operating characteristics and inherent safety characteristics of the heat pipe reactor core.

[0025] 3. The present invention performs three-dimensional refined modeling of the heat pipe reactor core, solving the problem that the point pile model cannot take into account the spatial effect of the core. Therefore, more realistic simulation calculation results can be obtained, providing accurate guidance for the design and optimization of the heat pipe reactor core.

[0026] 4. The thermal characteristics calculation is realized through internal coupling. The temperature field is loaded into the structured grid of the structural mechanics model in the finite element calculation platform, which can realize the coupled calculation of the thermal characteristics of the heat pipe reactor core efficiently and quickly.

[0027] 5. Import 3D heat loads or surface temperature distributions through command streams to precisely set model boundary conditions and improve the accuracy of numerical calculation results.

[0028] 6. In this invention, multiple heat pipes within the reactor core are treated as independent individual heat pipes for calculation. By establishing separate heat pipe and core heat transfer models, the core heat transfer model is simplified and the computational complexity is reduced. This effectively improves computational efficiency while ensuring accuracy. BRIEF DESCRIPTION OF THE DRAWINGS

[0029] Figure 1 It is a flow chart of the numerical simulation calculation method of the nuclear thermal characteristics of the heat pipe reactor core of the present invention.

[0030] Figure 2 FIG. 4 is a structural diagram of a heat pipe stack according to an embodiment.

[0031] Figure 3 yes Figure 2 Schematic diagram of the structure of the neutron unit core. DETAILED DESCRIPTION

[0032] The present invention will be described in further detail below with reference to the accompanying drawings.

[0033] like Figure 1 As shown, a numerical simulation calculation method for the thermal characteristics of the core of a heat pipe reactor is provided, and the steps are as follows:

[0034] Step 1: Establish the neutron physics model of the LEGO-30 heat pipe reactor core in the Monte Carlo software MCNP. Figure 2 The LEGO-30 heat pipe stack consists of six LEGO-5 sub-unit cores 1 and six B4C control rods 2 outside the stack. The six LEGO-5 sub-units are arranged in a regular hexagonal core structure. Figure 3 Each LEGO-5 subunit contains 84 fuel rods 3 and 43 heat pipes 4, with 2-3 heat pipes arranged around each fuel rod. Using UO2 fuel pellets, the LEGO-30 reactor core is designed for a total thermal power of 150 kW. The neutron source is set to have a particle count of 100,000 per generation, with 75 inactive generations and 425 active generations. The Monte Carlo method is used to solve the neutron transport equation, which is as follows:

[0035]

[0036] Φ(r,E,t)=∫ψ(r,E,Ω,t)dΩ

[0037] Where Ω is the solid angle; E is the neutron energy; r is the particle position; ψ(r, E, Ω, t) is the neutron flux density; Σ T is the total cross section; ∑ s (r, E′, Ω′→E, Ω) is the scattering cross section, which represents the probability that a neutron will be scattered from energy E′ and direction Ω′ to energy E and direction Ω; v is the neutron velocity; Φ(r, E, t) is the neutron fluence; and Q(r, E, Ω, t) represents the neutron source.

[0038] When the six control rods are raised to a height of 10.5 cm, during the neutron multiplication process in a finite-sized breeding medium, the ratio of the number of neutrons in a certain generation to the number of neutrons in the adjacent previous generation, i.e., the effective neutron multiplication coefficient keff, is 1.00215. At this time, the heat pipe reactor core is considered to be in critical operation.

[0039] Step 2: When the heat pipe reactor core is in critical operation, the standard deviation of keff is 0.00007, which meets the credibility criterion. The criticality calculation results of the neutron physics model are considered credible.

[0040] Step 3: Build the heat pipe and reactor core heat transfer models within the finite element calculation platform ANSYS Workbench. Set material thermophysical properties and boundary conditions, then perform meshing and independence verification on each model. A mesh-independent solution is achieved for the heat pipe model when the mesh count reaches 147,000, and for the reactor core model when the mesh count reaches 3.626 million.

[0041] Step 4: Since the statistics card for the input parameters in Step 1 is set to count the output power at 28 discrete points along the axis, the power distribution counting results need to be normalized. The conversion formula is as follows:

[0042]

[0043] Where, P 栅元 is the power of a certain grid element of the fuel element in the core neutron physics calculation; P0 is the designed thermal power of the reactor core; n i,j is the count value of the cell, i is the radial number of the cell, ranging from 1 to 84, j is the axial number of the cell, ranging from 1 to 28; N is the total number of cells of all fuel elements.

[0044] The unit power of each fuel cell is obtained, and then the axial power distribution curve of the fuel rod is fitted with a fifth-order polynomial formula to obtain the axial power distribution curve of each fuel rod;

[0045] Step 5: Based on the power distribution obtained in step 4, the heat flux of the heat pipe evaporation section wall is obtained. The heat pipe heat transfer model is imported as the heat source boundary condition. The adiabatic section wall is set to adiabatic. The heat pipe condensation section is connected to the Stirling engine through a secondary circuit containing potassium working medium. The condensation section wall and liquid potassium undergo convective heat exchange. The boundary conditions of the evaporation section, adiabatic section, and condensation section are set as follows:

[0046]

[0047] Where k w is the thermal conductivity of the heat pipe wall, W / (m·K); T is the heat pipe temperature, K; r is the radial coordinate of the heat pipe, m; q e is the heat flux density of the heat pipe wall at axial height, W / m 2 ; h is the convection heat transfer coefficient, which is 1000W / (m 2 ·K); T ∞ is the potassium working medium temperature of the secondary circuit, which is 1030K; T w is the wall temperature of the heat pipe, K; r outer is the outer wall radius of the heat pipe, m.

[0048] Solve the heat pipe heat transfer model to calculate the heat pipe temperature distribution;

[0049] Step 6: Use the func method to import the fuel rod power polynomial as the body heat source into the core heat conduction model through the APDL command flow, use the heat pipe evaporation section wall temperature as the cooling boundary, and calculate the core temperature field.

[0050] The core's critical operating temperature peaked at 1057.1 K. Furthermore, under the heat pipe failure accident condition, the temperatures of all materials remained within their respective allowable operating temperature limits, demonstrating the heat pipe reactor's excellent thermal safety characteristics.

[0051] Step 7: Establish a structural mechanics model in ANSYS Workbench and add material structural mechanics parameters and constraints. The bottom of the heat pipe reactor core is a fixed constraint, and other structures are free to expand.

[0052] Step 8: Based on the ANSYS Workbench platform, the temperature distribution was loaded into the structural mechanics model through internal coupling. The calculated maximum thermal stress of the structural material reached 992 MPa, exceeding the yield strength of 316L stainless steel. It was necessary to optimize the design of the heat pipe reactor core structure.

[0053] Step 9: Export and analyze the thermal coupling calculation results. The calculation results show that the heat pipe reactor has good thermal safety characteristics, but the structure needs to be optimized to complete the thermal characteristics analysis of the heat pipe reactor core.

[0054] Therefore, the method of the present invention can realize the numerical simulation of the entire process of nuclear thermal response of the heat pipe reactor, support the entire process of engineering design and safety analysis, and provide a reference for the design optimization of the heat pipe reactor.

Claims

1. A numerical simulation calculation method for the thermal characteristics of a heat pipe reactor core, characterized in that: The following steps are involved: Step 1: Establish a neutron physics model based on the core geometry and material parameters given in the heat pipe reactor design. Set input parameters and solve the neutron transport equations used to quantify the established neutron physics model to obtain the three-dimensional power distribution of the heat pipe reactor core under critical conditions. The input parameters include the neutron source term and the statistical card. Step 2: Determine the credibility of the statistical results of the neutron transport criticality calculation based on the preset standards; Step 3: Establish heat pipe heat transfer models and reactor core heat transfer models in the finite element calculation platform, set material thermophysical parameters and boundary conditions, and perform meshing and independence verification on the models. Step 4: normalize the three-dimensional power distribution of the heat pipe reactor core under critical conditions obtained in step 1 and convert it into a fuel rod axial power distribution curve by fitting a polynomial method; Step 5: Based on the three-dimensional power distribution obtained in step 1, the heat flux of the heat pipe evaporation section wall is obtained. The heat transfer model of the heat pipe is imported as the heat source boundary condition. The wall of the adiabatic section is set to adiabatic, and the condensing section is set to be coupled with the heat exchange system boundary condition to obtain the heat pipe temperature distribution. Step 6: Import the fuel rod axial power distribution curve obtained in Step 4 into the core heat transfer model as a heat source condition, and import the heat pipe temperature distribution obtained in Step 5 into the core heat transfer model as a cold source condition, perform core thermal calculations, and obtain the three-dimensional temperature distribution of the core; Step 7: Establish a structural mechanics model in the finite element calculation platform and add material structural mechanics parameters and constraints; Step 8: Based on the finite element calculation platform, the three-dimensional core temperature distribution obtained in step 6 is loaded into the core structural mechanics model through the internal coupling method, and the core structural mechanics calculation is performed to obtain the three-dimensional thermal stress and strain distribution of the core; Step 9: Derive the three-dimensional temperature distribution and thermal stress-strain distribution of the core to provide data basis for subsequent core thermal safety characteristics, structural safety characteristics and design optimization, and complete the nuclear thermal characteristics analysis of the heat pipe reactor core.

2. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: In step 1, the neutron transport equation is as follows: Φ(r,E,t)=∫ψ(r,E,Ω,t)dΩ Where Ω is the solid angle; E is the neutron energy; r is the particle position; ψ(r, E, Ω, t) is the neutron flux density; Σ T is the total cross section; ∑ s (r, E′, Ω′→E, Ω) is the scattering cross section, which represents the probability that a neutron will be scattered from energy E′ and direction Ω′ to energy E and direction Ω; v is the neutron velocity; Φ(r, E, t) is the neutron fluence; and Q(r, E, Ω, t) represents the neutron source.

3. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: The statistical card for the input parameters in step 1 is set to count the output power at 28 discrete points along the axis. The three-dimensional power distribution counting results are normalized. The conversion formula is as follows: Where, P 栅元 is the power of a certain grid element of the fuel element in the core neutron physics calculation; P0 is the designed thermal power of the reactor core; n i,j is the count value of the cell, i is the radial number of the cell, ranging from 1 to 84, j is the axial number of the cell, ranging from 1 to 28; N is the total number of cells in all fuel elements.

4. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: In step 4, the polynomial fitting method is a quintic polynomial fitting method.

5. The method for numerical simulation calculation of nuclear thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: In step 5, the boundary conditions of the evaporation section are set as: The boundary conditions of the adiabatic section are set as: The boundary conditions of the condensation section are set as: Where k w is the thermal conductivity of the heat pipe wall; T is the heat pipe temperature; r is the radial coordinate of the heat pipe; q e is the heat flux density of the heat pipe wall at axial height; h is the convection heat transfer coefficient; T ∞ is the potassium working medium temperature in the secondary circuit.

6. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: In step seven, the mechanical parameters include thermal expansion coefficient, Young's modulus and Poisson's ratio.

7. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: In step 2, the preset standard is that the standard deviation of the effective neutron multiplication coefficient keff should be less than 0.0001.

8. The method for numerical simulation of thermal characteristics of a heat pipe reactor core according to claim 1, characterized in that: The finite element calculation platform is ANSYS Workbench.

9. An electronic device, characterized in that: include: one or more processors; a memory for storing one or more programs; When the one or more programs are executed by the one or more processors, the one or more processors implement the steps of the method for numerical simulation calculation of nuclear thermal characteristics of a heat pipe reactor core according to any one of claims 1 to 8.

10. A storage medium having a computer program stored thereon, characterized in that: When the program is executed by a processor, the steps of the method for numerical simulation calculation of the thermal characteristics of the core of a heat pipe reactor as claimed in any one of claims 1 to 8 are implemented.

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