超临界水冷堆候选材料的应力腐蚀行为
Stress Corrosion Behavior of Candidate Materials for Supercritical Water-cooled Reactor
摘 要
采用慢应变速率试验(SSRT)研究了800H合金、825合金、HR3C不锈钢在550 ℃和650 ℃,25 MPa的超临界水环境中的应力腐蚀行为。结果表明,825合金在两个温度下的抗拉强度均最高、延伸率均最大,HR3C不锈钢和800H合金次之。由断口形貌可见,550 ℃和650 ℃时的800H合金、825合金和550 ℃时的HR3C不锈钢的失效模式均为韧性断裂和脆性断裂同时存在的断裂行为,而650 ℃时的HR3C不锈钢失效模式为完全的韧性断裂。
慢应变速率试验
应力-应变曲线
断口形貌
应力腐蚀开裂倾向
supercritical water-cooled reactor
slow strain rate testing
stress-strain curve
fracture morphology
stress corrosion cracking tendency
Abstract
Stress corrosion cracking(SCC) behaviors of alloy 800H, alloy 825 and stainless steel HR3C in supercritical water at temperature of 550 and 650 ℃ and pressure of 25 MPa were studied by slow strain rate testing(SSRT). The results show that alloy 825 had the highest tensile strength and elongation. Fractography indicates that the failure modes of alloy 800H and alloy 825 at 550 ℃ and 650 ℃, and stainless steel HR3C at 550 ℃ were combination of ductile and brittle fracture. HR3C showed fully ductile fracture at 650 ℃.
中图分类号 TG174
所属栏目 试验研究
基金项目 国家“973”重点基础研究发展计划(2007CB209802)
收稿日期 2013/1/8
修改稿日期
网络出版日期
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备注张乐福(1967-),副教授,博士,从事核电用材料相关研究,
引用该论文: SUN Yao,ZHANG Le-fu,LI Li. Stress Corrosion Behavior of Candidate Materials for Supercritical Water-cooled Reactor[J]. Corrosion & Protection, 2013, 34(12): 1053
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参考文献
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