核电站铸造奥氏体不锈钢热老化-疲劳失效机理与寿命预测

Thermal aging induced hardening and embrittlement occurs in the cast austenitic stainless steel(CASS) used for the primary coolant circuit piping (PCCP)in the pressurized- water-reactor (PWR) nuclear power plant during the long term service, posing a threat to the integrity of reactor coolant pressure boundary. This work plans to build a systematic knowledge base for the thermal aging-fatigue properties of Z3CN20.09M CASS which is being widely used for PCCP in our country.Low cycle fatigue tests under varied strain amplitudes will be first conducted on thermally aged specimens after varied durations at different temperatures. Scanning electron microscopy, transmission electron microscopy, and neutron diffraction technique will then be used to explore the failure mechanisms of thermal aging-fatigue. Finally, based on the microscopic mechanism investigation, an appropriate fatigue parameter will be determined to establish a fatigue life prediction model that takes thermal aging into account. It is worth to note that the real-time in situ neutron diffraction is an advanced technique that enables the quantitative analysis of the content and load distribution of constituent phases of the material. The output of this research project will help accumulate important data for the nuclear power safety system, and provide realistic guidelines for the extended service life of PWR in our country.

压水堆核电站一回路主管道材料采用的铸造奥氏体不锈钢长期在服役温度下会发生硬化和脆化等热老化现象，对反应堆冷却剂压力边界的完整性构成威胁。本项目拟对目前我国核电站一回路主管道广泛采用的Z3CN20.09M铸造奥氏体不锈钢的热老化-疲劳性能进行较为系统而深入的研究。针对不同温度和不同时长的热老化处理材料进行不同幅值的低周疲劳试验；综合利用扫描电子显微术、透射电子显微术和中子衍射技术来探究材料的热老化-疲劳失效机理；基于微观机理确定适宜的疲劳参数来构建考虑热老化影响的疲劳寿命预测模型。其中，实时原位中子衍射技术是本项目拟采用的先进技术手段，可以用来量化材料组成相的含量和承载分配状况。本项目的研究成果将为我国核电站安全体系的数据库积累数据，为第二代核电站的延寿运行提供预测和指导依据。

本项目首先对我国第二代核电站一回路管道材料铸造奥氏体不锈钢Z3CN20.09M进行了不同时长的热老化处理，进而通过力学试验研究了热老化对该材料的机械性能和低周疲劳性能的影响，通过原位中子衍射试验及显微观察手段探究了热老化影响该材料低周疲劳性能的微观机理，通过将疲劳参量与微观硬度相关联的手段提出了适用于该材料的热老化-低周疲劳寿命预测模型。本项目的研究成果有潜力被纳入实际工程结构设计和评估标准中，也将为我国核电站安全体系的数据库积累数据，为第二代核电站的延寿运行提供预测和指导依据。