Research on Fundamental Characteristics of Nuclear Grade 316H Stainless Steel at Ultra High Temperature

A fundamental feature of the fourth-generation-reactor is that most of the designed operating temperature are between 500℃ to 800℃, while the traditional material system and data of the PWR are below 350℃, which cannot meet the requirements. In this paper, 316H is selected as the research object through demonstration and analysis, which is suitable for most reactors and closest to engineering application. The experimental study on mechanical properties, specific heat capacity, average linear expansion coefficient, intergranular corrosion characteristics and low cycle fatigue at 800℃ was carried out. the result shows that the measured data are significantly higher than the standard values. It is recommented that the temperature limit for the long-term operation shall not exceed 700℃, and the temperature limit for the short-term operation shall not exceed 800℃. This study provides a basis for the selection and evaluation of the structural materials of the fourth-generation-reactor.