- 中国核心期刊（遴选）数据库收录期刊
- 中国科技论文统计源期刊
- 中国学术期刊综合评价数据库来源期刊
| Citation: |
ZHONG Li, LIU Xueying, MA Chenhao, XU Hong, ZHANG Naiqiang, ZHU Zhongliang. Effect of Strain Rate on Stress Corrosion Cracking Sensitivity of Domestic Nickel-Based Alloy in Supercritical Water Environment[J]. Corrosion & Protection, 2024, 45(3): 7-16. DOI: 10.11973/fsyfh-202403002
|
The stress corrosion cracking (SCC) sensitivity of domestic nickel-based alloy C-HRA-1 in supercritical water environment of 650 ℃ and 25 MPa at different strain rates was studied by slow strain rate test (SSRT). The fracture morphology and surface crack morphology of samples were observed by scanning electron microscope (SEM), and the element composition and distribution near the crack tip were analyzed by energy dispersive spectrometer (EDS). The results show that C-HRA-1 exhibited SCC sensitivity at different strain rates, and the SCC sensitivity increased with the decrease of strain rate. In the process of stress corrosion, the formation and propagation of cracks were affected by the coupling of oxidation and mechanics. The grain boundary oxidation occurred at the crack tip, which reduced the crack tip bonding force and led to crack propagation.
| [1] |
卢征然, 曲生志, 陈亮. C-HRA-1镍基合金焊接接头的高温长期性能研究[J]. 锅炉技术, 2017, 48(1): 42-46.
|
| [2] |
刘正东, 陈正宗, 何西扣, 等. 630～700 ℃超超临界燃煤电站耐热管及其制造技术进展[J]. 金属学报, 2020, 56(4): 539-548.
|
| [3] |
徐松乾, 王婷婷. 700 ℃先进超超临界机组过/再热器材料C-HRA-1合金组织稳定性研究[J]. 热力发电, 2016, 45(6): 63-69.
|
| [4] |
吕太, 白杰, 董璐, 等. 燃煤电厂锅炉辐射受热面温度及热应力分析[J]. 锅炉技术, 2016, 47(3): 6-11.
|
| [5] |
SHEN Z, ZHANG L F, TANG R, et al. SCC susceptibility of type 316Ti stainless steel in supercritical water[J]. Journal of Nuclear Materials, 2015, 458: 206-215.
|
| [6] |
WAS G S, AMPORNRAT P, GUPTA G, et al. Corrosion and stress corrosion cracking in supercritical water[J]. Journal of Nuclear Materials, 2007, 371(1/2/3): 176-201.
|
| [7] |
MUTHUKUMAR N, LEE J H, KIMURA A. SCC behavior of austenitic and martensitic steels in supercritical pressurized water[J]. Journal of Nuclear Materials, 2011, 417(1/2/3): 1221-1224.
|
| [8] |
TERACHI T, YAMADA T, MIYAMOTO T, et al. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water[J]. Journal of Nuclear Materials, 2012, 426(1/2/3): 59-70.
|
| [9] |
NOVOTNY R, HAÄGHNER P, SIEGL J, et al. Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions[J]. Journal of Nuclear Materials, 2011, 409(2): 117-123.
|
| [10] |
陈震宇, 张乃强, 周密, 等. 先进超超临界机组候选镍基617合金在超临界水中的腐蚀疲劳裂纹扩展速率试验研究[J]. 动力工程学报, 2023, 43(4): 414-420,435.
|
| [11] |
李力, 张乐福, 唐睿. 奥氏体不锈钢和镍基合金在550 ℃/25 MPa超临界水中的应力腐蚀开裂敏感性[J]. 腐蚀与防护, 2012, 33(11): 928-931.
|
| [12] |
李冬升, 凌瀚宇, 王威, 等. IN740H镍基高温合金的应力腐蚀性能[J]. 材料热处理学报, 2023, 44(2): 85-92.
|
| [13] |
刘宇桐国产镍基合金C-HRA-2应力腐蚀开裂性能试验研究北京华北电力大学2022刘宇桐. 国产镍基合金C-HRA-2应力腐蚀开裂性能试验研究[D]. 北京: 华北电力大学, 2022.
|
| [14] |
李季, 唐丽英, 周荣灿, 等. C-HRA-1锅炉管在700 ℃实炉条件下的力学性能及组织演变研究[J]. 中国电机工程学报, 2022, 42(18): 6580-6588.
|
| [15] |
刘保平, 张志明, 王俭秋, 等. 核用结构材料在高温高压水中应力腐蚀裂纹萌生研究进展[J]. 中国腐蚀与防护学报, 2022, 42(4): 513-522.
|
| [16] |
FORD F P. Quantitative prediction of environmentally assisted cracking[J]. Corrosion, 1996, 52(5): 375-395.
|
| [17] |
MACDONALD D D, LU P C, URQUIDI-MACDONALD M, et al. Theoretical estimation of crack growth rates in type 304 stainless steel in boiling-water reactor coolant environments[J]. Corrosion, 1996, 52(10): 768-785.
|
| [18] |
MACDONALD D D, URQUIDI-MACDONALD M. A coupled environment model for stress corrosion cracking in sensitized type 304 stainless steel in LWR environments[J]. Corrosion Science, 1991, 32(1): 51-81.
|
| [19] |
SCOTT P MAn overview of internal oxidation as a possible explanation of intergranular stress corrosion cracking of alloy 600 in PWRsProceedings of the 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors1999SCOTT P M. An overview of internal oxidation as a possible explanation of intergranular stress corrosion cracking of alloy 600 in PWRs[C]//Proceedings of the 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. [S. l.]: [s. n.],1999.
|
| [20] |
VAN BUEREN H G. Theory of the formation of lattice defects during plastic strain[J]. Acta Metallurgica, 1955, 3(6): 519-524.
|
| [21] |
KUANG W J, WAS G S. A high-resolution characterization of the initiation of stress corrosion crack in alloy 690 in simulated pressurized water reactor primary water[J]. Corrosion Science, 2020, 163: 108243.
|
| [22] |
朱忠亮, 马辰昊, 李宇旸, 等. 镍基合金Inconel617B在700 ℃超临界水环境中的氧化行为研究[J]. 中国腐蚀与防护学报, 2022, 42(4): 655-661.
|
| [23] |
谢冬柏, 周游宇, 鲁金涛, 等. Cr对镍基合金在超临界水中氧化行为的影响研究[J]. 中国腐蚀与防护学报, 2018, 38(4): 358-364.
|
Supported by: Beijing Renhe Information Technology Co., Ltd. 
DownLoad: