Development and Preliminary Validation of Three-dimensional Subchannel Thermal-hydraulic Analysis Code CorTAF for Numerical Reactor Core

被引：0
|
作者：
Liu K. [1 ]
Wang M. [1 ]
Tian W. [1 ]
Qiu S. [1 ]
Su G. [1 ]
机构：
[1] School of Nuclear Science and Technology, State Key Laboratory on Power Engineering and Multiphase Flow, Shaanxi Provincial Key Laboratory of Advanced Nuclear Energy Technology, Xi'an Jiaotong University, Xi'an
来源：
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | 2022年 / 56卷 / 02期
关键词：
Coupled heat transfer; OpenFOAM; PWR core; Subchannel analysis;
D O I：
10.7538/yzk.2021.youxian.0860
中图分类号：
学科分类号：
摘要：
Reactor core is the pivotal component of nuclear power system, and its integrity is an important prerequisite for safe operation of reactor. Traditional nuclear reactor core thermal-hydraulic analysis methods cannot meet the high precision simulation requirements of advanced nuclear power systems in the future. In this paper, based on open source CFD platform OpenFOAM, a coolant flow heat transfer model considering diffusion due to turbulent mixing between adjacent coolant channels and rod bundle structure characteristics of pressurized water reactor (PWR) was established. A fuel rod heat conduction model was built to describe the internal temperature distribution formed by multiple nodes placed along the radial direction of fuel rods, and a coupled heat transfer model were proposed according to the mapping relationship between the external boundary condition of fuel rod heat conduction equation and convective heat transfer energy source term in coolant governing equation. The three-dimensional subchannel thermal-hydraulic characteristics analysis code CorTAF for PWR core based on finite volume method was developed. After that, various fuel assembly flow and heat transfer experiments were selected to carry out model validation. For GE3×3 experiment, the calculated results of coolant velocity distribution in each channel of test section agree well with the experimental data, indicating that the CorTAF code can predict the flow characteristics in the fuel assembly effectively against COBRA code with similar computational accuracy. For Weiss experiment, the calculated results of coolant flow rate distribution in each assembly of test section are in good agreement with the experimental data, which illustrates that the CorTAF code can accurately predict the flow characteristics in parallel open assemblies under inlet partially blocked condition. For PNL2×6 experiment, the calculated results of coolant temperature and velocity distribution along center line of experimental section are basically consistent with the experimental data and those calculated by CUPID and MATRA codes, demonstrating that the CorTAF code can obtain the flow and heat transfer characteristics in the rod bundle assembly under condition of transverse non-uniform heating, and the error may be caused by the difference of data acquisition methods. In conclusion, the CorTAF code is proved to be suitable for predicting the flow and heat transfer characteristics in the rod bundle fuel assembly. The work in this paper has reference significance for the development of core safety analysis tools for PWR. Subsequently, further model optimization and thermal-hydraulic characteristics analysis of the full PWR core under both steady-state operation and transient accident conditions will be carried out, and research on neutronics and thermal-hydraulics coupling will be performed based on OpenFOAM. © 2022, Editorial Board of Atomic Energy Science and Technology. All right reserved.
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页码：261 / 270
页数：9
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