Research on Fundamental Characteristics of Nuclear Grade 316H Stainless Steel at Ultra High Temperature
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摘要: 第四代反应堆的一个基础特征是设计运行温度大多数在500~800℃，而传统压水堆材料体系和数据均在350℃以下得到，无法满足需求。本文通过广泛论证分析，筛选出了适用于大多数反应堆、最接近工程应用的316H不锈钢材料作为研究对象。开展800℃超高温下力学性能、比热容、平均线膨胀系数、晶间腐蚀特性、低周疲劳等试验研究，结果表明，316H不锈钢实测数据结果大幅高于规范标准值，长期应用温度限值建议不超过700℃，短时瞬态运行温度限值建议不超过800℃。该研究为第四代反应堆结构材料筛选和评价提供了依据。Abstract: 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.
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Key words:
- 316H stainless steel /
- Ultra high temperature /
- Mechanical properties /
- Low cycle fatigue
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图 8 不同温度下敏化后的断口形貌图
Figure 8. Fracture Morphology of Sensitized Specimens at Different Temperatures
表 1 拟合得到的高温疲劳参数
Table 1. Fitting for High Temperature Fatigue Parameters
温度/℃ $\sigma {'_{\rm{f}}} $/MPa b ε′f c 600 1053 −0.1577 0.0847 −0.4097 650 632 −0.1118 0.0378 −0.3485 800 452 −0.131 0.1246 −0.4983 
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表 2 不同温度下316H的
$k'$ 和$n'$ Table 2.
$k'$ and$n'$ of 316H at Different Temperature温度/℃ $k'$ $n'$ 600 1345 0.2708 650 1273 0.2663 800 427 0.1683
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