核级690合金辐照加速应力腐蚀开裂机理研究

堆芯内核反应释放的中子对堆芯内结构材料组织产生损伤，导致材料的应力腐蚀开裂问题愈加严重，对反应堆长期运行的安全性和经济性构成威胁。初步研究显示690合金在辐照条件下仍然有优异的抗应力腐蚀开裂性能，是堆芯内应力腐蚀开裂敏感材料的理想替代品。但是，目前对690合金的辐照应力腐蚀开裂机理，辐照对应力腐蚀开裂的加速程度和加速机制并不了解，其在堆芯的长期服役性能还很难预估，故而其入堆辐照考验还未被提上日程。本项目拟进行应力腐蚀裂纹萌生实验和微观表征，探讨辐照组织的变形机制；研究晶界氧化和开裂与组织结构、化学成分、材料变形的联系，揭示辐照加速应力腐蚀开裂的程度和机制；在此基础上构建690合金的辐照应力腐蚀开裂机理模型，预评其在辐照环境下的长期服役性能。该研究将为690合金入堆辐照考验提供理论依据，加速690合金入堆考验进程和在中子辐照环境中的应用，同时为新型抗辐照应力腐蚀开裂材料的开发提供借鉴。

The neutron particles released from nuclear reaction cause damage to the microstructure of structural materials in the nuclear reactor core, which increases the stress corrosion cracking (SCC) problem and threatens the safety and economic efficiency of the nuclear power plant. The preliminary research of Alloy 690 reveals that it has very good SCC resistance under the irradiation condition, making it a perfect candidate for the replacement of the high IASCC sensitive materials. However, the mechanism of irradiation-assisted SCC and the degree of acceleration on crack are still unclear, which makes it difficult to evaluate its longtime performance in a nuclear reactor. Therefore, the test in the nuclear reactor has not been put forward for discussion. This project plans to conduct SCC experiments and associated microstructure characterization to study the feature and mechanism of deformation of irradiated microstructure. The relation of oxidation and cracking of grain boundary with deformation, microstructure, and chemistry will be studied to figure out the critical factor for the acceleration of SCC and reveal the mechanism of IASCC of Alloy 690. On this basis, the mechanism model of IASCC of Alloy 690 will be established, and the longtime performance of this material in the reactor core will be estimated. The results of this project will provide a theoretical foundation for the test of Alloy 690 in the reactor core and the large-scale application under the irradiation condition, and inspire a new path for the development of IASCC resistant materials, as well.