核动力堆芯流态的空间分布特性与机理研究

The marine nuclear power is of great important to the national security and economic development of China. In order to ensure the safety, accuracy and efficiency of the power output and regulation, it is necessary to improve the prediction of the thermal hydraulic status of nuclear power reactor core. There are tens of thousands of narrow coolant channels, and nearly 1000 spacer grids with numerous small and complex mixing vanes, dimples, and springs in a typical reactor core, which causes the significant spatial differences of the thermal hydraulic status, such as the turbulence characteristics, and the flow mixing effect. The mesoscale computational fluid dynamics (CFD) analysis technology and the subchannel analysis technology which can consider the status in the fuel cell scale are two effective analysis tools. However, the former one needs the turbulence model, and the later one needs the computing model for the mixing effect. In current, both models have not been designed for the fine scale thermal hydraulic status of reactor core, and the adaptability is very limited. Thus, this project takes advantages of the fine scale experimental analysis and theoretical analysis to study the fine scale thermal hydraulic status, law and mechanism under the influence of the characteristic structure and operating conditions of the reactor core. The fine analysis models will also be built to consider the spatial difference feature for the turbulence and mixing effect. Then the analysis technology of thermal hydraulic will be optimized. This project will be beneficial for the nuclear safety and the performance of the marine nuclear power.

船舶核动力是国家安全与经济战略基石。为安全、准确、高效地进行功率调节，需正确认知核动力堆芯的热工水力状态。典型堆芯数万细窄冷却剂流道，以及近千个具有搅混翼、刚凸、弹簧等复杂细小繁多结构特征的定位格架，致使堆芯的湍流特性、流动交混特性存在显著的空间差异性。能够考虑堆芯介观尺度特性的计算流体动力学（CFD）分析技术，以及能够考虑堆芯栅元尺度流域状态的子通道分析技术是有效的分析工具。然而，前者所需的湍流模拟方案，后者所需的流动交混计算模型及方案，目前仍没有针对堆芯细致空间流动特性进行精细化设计，适用性有限。项目通过开展堆芯精细化流场实验、理论分析研究堆芯特征结构与运行工况影响下的精细化流动状态、规律与机理，并建立考虑湍流与流动交混空间分布差异性的计算模型与方案，优化堆芯热工水力计算分析技术。项目成果有利于船舶核动力的核安全性与动力性能。

核反应堆是船舶核动力装置最核心的部分，其内部精细流动状态对核动力反应堆的最大温度、压力、流速等重要参量以及反应堆输出的最大功率与功率调节速度造成显著影响，直接决定核动力船舶的最大航速与航行机动调节性能。为提升核反应堆堆芯热工水力状态预测水平以提升船舶核动力性能，需研究并有效利用堆芯复杂流动状态、规律与机理，为此本项目开展了堆芯内湍流与流动交混状态的精细分布状态与特性研究，完成了堆芯精细建模与计算分析技术研究，并通过我国及美国、韩国棒束组件实验数据及关联式完成测试验证，实现了10亿级精细网格的堆芯特征流域计算分析，完成了堆芯不同流域中冷却剂的速度分布、湍流强度、湍流各向异性程度等湍流状态的量化研究，确定了二次流强度、湍流强度、压降等参量的空间量化分布规律，研发了堆芯各子流域多湍流模型联合应用方案，提升了对堆芯湍流空间差异性分布特性的适用性；并针对堆芯121组全高度典型组件研发了流域分区与耦合连接技术、分布并行及自动化大规模CFD计算技术，提升了计算效率，实现了所有全高度典型组件域的精细计算；确定了交混横流导致的影响域范围，量化分析了堆芯反向横流交混特性对质能输运的影响，明确了反向横流作用对结果精度的重要影响，并研发了考虑精细流动的精细栅元控制体划分与计算方案；研究了冷却剂流动通道中心旋涡竖直方向上的运动特性，研究了旋涡流体与交混流体间的黏性作用特性，研发了子通道横流方程的源项方案。本项目的开展有利于提升堆芯流动状态空间差异性及流动质能交混特性的认知水平，所研发精细化计算加速技术有利于促进堆芯精细化计算分析技术的工程应用，为数字化反应堆及核动力数字孪生等技术的研发提供支持，为核动力堆芯的设计与运行提供辅助支撑。