正激波与纳米流体沸腾耦合作用下的熔融液滴表面触发机理研究

反应堆严重事故下的堆芯熔融物与冷却剂相互作用后果是安全评估的重要内容，其中熔融液滴的触发与细粒化是关键机理，而相应的模型对评估结果影响很大。纳米流体作为新型冷却介质，是未来优化堆外注水策略的重要选择，在此种情况下堆芯熔融物与冷却剂相互作用机理发生了改变，其相应的评估模型也需要发展。本课题针对堆芯熔融物与冷却剂相互作用关键过程中外部引入正激波与新型冷却介质沸腾耦合作用下的熔融液滴表面触发机理开展研究，并发展相应的模型，确立正激波与冷却剂沸腾耦合作用下的熔融液滴表面的气液界面不稳定性传播与发展的主导机制，建立考虑纳米流体影响的触发发生判据。通过本课题的研究，采用理论与实验相结合的方法，阐明纳米流体中气液界面演化和正激波触发机理，获得纳米流体影响气液界面不稳定性的传播与发展规律，建立机理模型。相关研究对开发多相热物理分析工具具有理论意义，对未来新型核反应堆安全设计、安全评估具有实际应用价值。

The consequence of core-melt and coolant interaction in severe accident is an important part of the safety assessment for nuclear power plant. Moreover, triggering and fragmentation of molten droplets during melt/coolant interaction are the crucial mechanisms and their corresponding models have significant impacts on the outcome of safety assessment. In addition, nano-fluid, which can be treated as a newly-designed coolant, may be an important option to optimize the ex-vessel cooling strategy in the future. However, the mechanism of core-melt and coolant interaction is varied under this circumstance, and its relevant model for safety assessment needs to be developed. In this research, triggering mechanism of molten droplet under coupled effects of normal shock waves from external and boiling of new type of coolant, which is a key process of melt/coolant interaction is investigated, with its corresponding model being proposed. The dominant regime which leads to the propagation of instability on the gas-molten droplet interface under coupled effects of normal shock waves and boiling of coolant is determined, based on which, a criterion which judges whether trigger event may occur is given in consideration of the effect of nano-fluid. Through this project, figure out gas-liquid interface evolution and triggering mechanism under normal shock in nano-fluid, and obtain how the nano-fluid can influence the propagation of instability on gas-liquid interface and a mechanistic model, with combining theoretical derivation and experimental tests. The present study would have theoretical significance to the development of multi-phase thermophysical analysis codes, and have practical value to the safety design and assessment for new type of nuclear power plants in the future, as well.