非均匀来流条件下核主泵两相降级机理及非稳态激励特性研究

The two-phase degradation characteristic of the reactor coolant pump is a key parameter for assessing the safety of a nuclear power plant. In-depth study of the two-phase degradation mechanism of the reactor coolant pump and its unsteady excitation characteristics caused by the non-uniform inflow are the fundamental way to ensure the reliability of the reactor coolant pump and the safety of the nuclear power plant. The project takes the reactor coolant pump non-uniform inflow as the research object. The Particle Image Velocimetry (PIV) and the Pulsed Shadowgraphy Technique (PST) will be used to study the gas-liquid two-phase flow of the nuclear main pump inlet under non-uniform inflow condition. Combined with the LES and VOF multiphase flow model numerical calculation method, the high speed photography technology is utilized to reveal the spatiotemporal evolution characteristics of the gas-liquid two-phase flow in the reactor coolant pump. Based on the fluid feature recognition and extraction technique, the spatiotemporal distribution characteristics and evolution laws of vortex structures with different frequencies are discussed by using Proper Orthogonal Decomposition (POD). The inflow mechanism of non-uniform inflow and blade interference is analyzed. The inner flow mechanisms of non-uniform inflow interference with impeller blade are analyzed. At the same time, the pressure and velocity pulsation signals are collected by the pressure pulsation acquisition system and the LDV technology, and the unsteady excitation characteristics induced by the non-uniform inflow are explored. Combined with the two-phase degradation characteristics of the reactor coolant pump, the internal correlation mechanisms of vortex structure, gas-liquid two-phase flow, hydraulic unsteady excitation characteristics, pump inner flow and two-phase degradation characteristics caused by non-uniform inflow are constructed. The project research can provide theoretical basis and technical support for the accurate assessment of nuclear power plant safety under accident conditions.

核主泵两相降级特性是评估核电厂安全性的关键参数，非均匀来流条件下核主泵气液两相降级机理及非稳态激励特性是保障核主泵可靠性和核电厂安全性的关键问题。项目以核主泵非均匀来流为研究对象，采用粒子图像测速技术PIV和脉冲阴影技术PST，研究非均匀来流条件下核主泵进口气液两相流流态。采用高速摄像技术，并结合LES和VOF数值计算方法，揭示核主泵气液两相流流动时空演变特性。基于流态特征识别与提取技术，并采用本征正交分解POD探讨不同频率涡结构时空分布特征及演变规律，分析非均匀来流与叶片干涉的内流作用机制。同时采用压力脉动采集系统和LDV技术对压力、速度脉动信号进行采集，探索非均匀来流诱发的非稳态激励特性，并结合核主泵两相降级特性，构建非均匀来流引起的涡结构、气液两相流流动、非稳态激励特性、泵内流场与两相降级特性的内在关联机制。项目研究可以为事故工况下核电厂安全性准确评估提供理论依据和技术支撑。

核主泵两相降级特性是评估核电厂安全性的关键参数，非均匀来流条件下核主泵气液两相降级机理及非稳态激励特性是保障核主泵可靠性和核电厂安全性的关键问题。项目针对非均匀来流条件下核主泵两相降级机理及激励特性开展系统研究，搭建了泵外特性、内部流动结构可视化及其诱导时频域特性一体化同步测量的非均匀来流核主泵实验台，通过时间分辨层析粒子图像测速技术模拟来流情况下的速度和压力，重构了三维空间非均匀流场，以评估其对核主泵的影响。采用乘法代数重建技术重构了组成整个非均匀流场的五个体积速度场。采用本征正交分解方法对流场进行分析研究。开展了直管均匀来流与非均匀来流对核主泵外特性、流场、空化、空化脉动的影响，基于精细化数值计算方法分析了核主泵入口及内部流动非稳态时空演变特性，开展了叶片泵空化两相流共性基础理论研究。最后基于研究发现和成果，提出了多项非均匀来流主动控制和抑制措施，提出了多项叶片泵性能提升措施。项目研究可以为事故工况下核电厂安全性准确评估、后续不同比尺对性能影响和泵综合性能优化研究提供理论依据和技术支撑。