核主泵汽液两相流动机理与瞬态特性研究

The nuclear reactor cooling pump is the only revolving equipment in the nuclear power station and it is also a key nuclear power equipment which has not realized localization.It is a critical period of time after the water loss accident happens, from the water loss accident happening to nuclear reactor cooling pump stopping, the medium of whole system is mainly gas-liquid two-phase. If it does not control pro- perly，it will lead to serious accidents, such as reactor core melt and so on. Water loss accident condition of transient two-phase flow is critical important to the safety of reliable operation of the nuclear power station. This field has become the focus of study at home and abroad in recent years. This project is mainly about unstable features in gas-liquid two-phase transient excessive process of nuclear reactor cooling pump in water loss accident condition.Use the method of combining theoretical derivation，numerical simulation and high temperature and high pressure transient test carry on study; establish mathematical model which is applied totransient gas-liquid two-phase of nuclear reactor cooling pump;revealgas-liquid two-phase transient change rule in water loss accident process;explore the influence of the production and collapse of gas phase toflow characteristics of nuclear reactor cooling pump, Master the influence between pressure fluctuation which produced by different degree of crevasse water loss accident in transient process and high head safety injection and the unstable flow in nuclear reactor cooling pump，Control and realize normal stop of nuclear reactor cooling pump in water loss condition，provide some theoretical basis to realize nuclear reactor cooling pump localization.

核主泵是核电站一回路系统中唯一旋转设备，也是未实现国产化的关键核动力设备。在一回路失水或蒸汽发生器二次侧管路破裂等事故发生后的一段时间非常关键，从事故发生至核主泵停机间整个系统内的介质主要以汽液两相的形式存在，若控制不当将会导致堆芯融化等严重核电事故发生。瞬态汽液两相流对核电站安全可靠运行至关重要，该领域已成为近年来国内外研究的热点和难点。本课题针对核主泵汽液两相瞬态流动特性进行研究，采用理论推导、数值模拟、高温高压瞬态测试相结合进行研究；建立适用于核主泵瞬变汽液两相数学模型；揭示汽液临界两相瞬态变化规律；探索汽相在流动过程中产生和溃灭对核主泵流动规律的影响；掌握不同程度破口失水事故瞬态过渡过程所产生膨胀脉动压力与高压安注所产生的脉动压力对核主泵内部不稳定流动之间的影响；控制和实现核主泵在失水事故下正常可靠停机，为实现核主泵国产化提供部分基础理论。

核主泵是核电站一回路系统中唯一旋转设备，也是未实现国产化的关键核动力设备。失水事故发生后的一段时间非常关键，从失水事故至核主泵停机间整个系统内的介质主要以汽液两相的形式存在，若控制不当将会导致堆芯融化等严重核电事故发生。失水事故工况瞬态汽液两相流对核电站安全可靠运行至关重要，该领域已成为近年来国内外研究的热点和难点。本课题针对核主泵失水事故工况汽液两相瞬态过渡过程不稳定流动特性进行研究，采用理论推导、数值模拟、高温高压瞬态测试相结合进行研究；建立核主泵失水瞬态过渡过程中，适用于不同程度破口时汽液临界两相瞬态流动特性分析的数学模型，为后续的瞬变两相流及水动力特性研究提供理论基础；揭示核主泵失水瞬态过渡过程中不同程度破口下汽液临界两相下核主泵内部流动机理，为采取有效措施实现核主泵在失水事故工况下安全可靠停机提供理论基础，同时也为顺利实现核主泵国产化过程中水力设计和结构设计提供理论依据；探索核主泵失水瞬态过渡过程中破口处的脉动压力与高压安注所产生的脉动压力对核主泵内部汽液临界两相不稳定性的影响，同时也获取核主泵在失水降压时所产生汽蚀变化规律及汽蚀诱导动态压力频谱特征，为改善瞬态过渡过程中抗汽蚀性能和流动稳定性提供理论基础。发表学术论文21篇（SCI收录5篇，EI收录12篇）；在核主泵水力设计与多学科优化设计方法、结构设计、高温高压试验台设计、控制和缓解失水事故不稳定流动等方面，申请发明专利12项，获授权发明专利8项。在核主泵失水事故汽液两相瞬态过渡过程不稳定流动特性、数值、高温高压试验等方面培养博士研究生1名，培养硕士研究生5名。