采用2 MeV质子束在360 ℃对国产核用304不锈钢试样进行了辐照实验，利用高温高压水环境的慢应变速率拉伸实验(SSRT)和SEM、EBSD、TEM等研究了核用304不锈钢辐照促进应力腐蚀开裂(IASCC)机理。结果表明，慢应变速率拉伸过程中辐照促进材料晶界和表面滑移台阶处形成应变集中，且其程度随辐照剂量增加而增加。滑移台阶穿过或终止于晶界，终止于晶界的台阶造成晶界处产生不连续滑移，易将位错传输到晶界，在晶界区域形成位错塞积和残余应变集中。而台阶不连续滑移的形成则受毗邻晶粒的Schmidt因子对的类型影响。另一方面，辐照促进晶界发生贫Cr富Ni元素偏析，其偏析程度随辐照剂量增加而增加。SSRT实验后辐照试样表面发生明显的沿晶应力腐蚀开裂，且裂纹数量随辐照剂量和外加应变增加而增加。同时，裂纹尖端区域发生明显晶界腐蚀，且氧化物宽度和长度随辐照剂量增加而增加。分析认为，辐照致晶界应变集中和元素偏析的协同作用造成材料变形行为和晶界腐蚀行为变化是IASCC发生的关键因素。

Irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. Previous studies on the IASCC had mainly focused on the effect of irradiation on changes in deformation modes and interaction of dislocation channels with grain boundary. The role of corrosion in IASCC, however, has not received sufficient attentions. In the process of stress corrosion cracking (SCC), corrosion occurs simultaneously with localized deformation in the vicinity of the crack tip. This indicates that corrosion is one of the potential contributors to IASCC. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel (304SS) was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile (SSRT) tests at 320 ℃ in simulated primary water of pressurized water reactor containing 1200 mg/L B as H₃BO₃ and 2.3 mg/L Li as LiOH·H₂O, with a dissolved hydrogen concentration of 2.6 mg/L. Following the SSRT tests, the localized deformation, corrosion and IASCC of the specimens were characterized. The results revealed that increasing the irradiation dose promoted residual strain accumulation at slip steps and grain boundaries of nuclear grade 304SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually promoted localized deformation at the grain boundary. Specially, the slip step transmitted at grain boundary led to slip continuity at the grain boundary. In contrast, a slip discontinuity was observed at the grain boundary where the slip step terminated, which caused a much higher strain accumulation by feeding dislocations to the grain boundary region. Further, formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains. The irradiation resulted in a depletion of Cr and an enrichment of Ni at grain boundary, while the magnitude of Cr depletion and Ni enrichment increased with increasing the irradiation dose. Following the SSRT tests, intergranular cracking was observed on surfaces of the irradiated specimens, while the number of the cracks was increased by a higher irradiation dose and applied strain. This suggested a higher IASCC susceptibility of nuclear grade 304SS in the primary water. Meanwhile, significant intergranular oxidation ahead of the crack tip was observed, while both the width and length of the oxide were larger at a higher irradiation dose. The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.