This paper provides a summary and analysis on the initiation of stress -corrosion cracking (SCC) of Alloy 600 (and related alloys) and stainless steels in nuclear reactor environments near 300 °C. Two processes seem to be comparable between traditional slip dissolution/oxidation mechanisms and aging-related crystalline ordering. This paper evaluates various supporting topics, including SCC initiation time, activation energy, stress/strain rate/plastic deformation, electrode potential drop, rate-limiting step, long-term threshold values, and data uncertainties. The understanding of Alloy 600 SCC is extended to stainless steels. SCC mechanisms of nickel-based alloys and stainless steels at lower temperatures are different from nickel-based alloys at higher temperatures. Finally, this paper describes how this work would be potentially considered in the nuclear industry.

本文对在300°C附近的核反应堆环境中600合金（及相关合金）和不锈钢的应力腐蚀开裂（SCC）的萌发进行了总结和分析。在传统的滑移溶解/氧化机理和与老化相关的晶体有序化之间，有两个过程似乎是可比的。本文评估了各种支持主题，包括SCC起始时间，活化能，应力/应变速率/塑性变形，电极电势下降，速率限制步骤，长期阈值和数据不确定性。对600 SCC合金的了解已扩展到不锈钢。较低温度下的镍基合金和不锈钢的SCC机理与较高温度下的镍基合金的SCC机理不同。最后，