| Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water | |
| Li Xiaohui1; Wang Jianqiu2; Han En-Hou2; Guo Yanjun1; Zheng Hui3; Yang Shuangliang3 | |
| Corresponding Author | Wang Jianqiu(wangjianqiu@imr.ac.cn) |
| 2020-11-11 | |
| Source Publication | ACTA METALLURGICA SINICA
 |
| ISSN | 0412-1961 |
| Volume | 56Issue:11Pages:1474-1484 |
| Abstract | The abnormal shutdown of the pressurized water reactor (PWR) nuclear power plants can be primarily attributed to the rupturing of the heat transfer tube of the steam generator. Regardless, stress corrosion cracking is the most important ageing mechanism associated with the primary water of the PWR. In this work, the damage behavior of alloy 690 was systematically investigated using high-temperature and high-pressure in situ scratching and electrochemical techniques to understand its corrosion behavior and failure mode and provide a reference for controlling the manufacturing, processing, and installation of the alloy 690 tubing. Further, the polarization behavior of alloy 690 at different temperatures was investigated using the self-built high-temperature and high-pressure water circulation circuit system and the high-temperature and high-pressure in situ scratching device. Subsequently, the single-pass scratch in air and in situ reciprocating scratch of alloy 690 obtained using high-temperature and high-pressure water for 11 and 100 h, respectively, were studied. The samples after scratching were observed and analyzed via SEM and EDS. The results revealed the occurrence of microcracks at the bottom of the scratch during the single-pass scratch of alloy 690. The TiN inclusions with large particles were prone to fragmentation, whereas those with smaller particles were susceptible to cracking at the joint of the matrix. During the reciprocating scratch process in high-temperature and high-pressure water, a portion of the metal substrate debris at the bottom of the scratch groove was peeled off along with oxide particles, microcracks, and chipped debris. Further, the TiN inclusions with large particles were fragmented, whereas those with smaller particles easily cracked at the bonding interface of the substrate. The electrochemical signals of alloy 690 during the reciprocating scratch processes were measured using the high-temperature and high-pressure in situ electrochemical technology. The instantaneous peak current density at the scratch during the scratch process is 149 similar to 326 times of that associated with the substrate. |
| Keyword | alloy 690 high temperature high pressure water corrosion in situ scratch |
| Funding Organization | National Science and Technology Major Project |
| DOI | 10.11900/0412.1961.2020.00091 |
| Indexed By | SCI |
| Language | 英语 |
| Funding Project | National Science and Technology Major Project[2015ZX06002005] |
| WOS Research Area | Metallurgy & Metallurgical Engineering |
| WOS Subject | Metallurgy & Metallurgical Engineering |
| WOS ID | WOS:000584345200004 |
| Publisher | SCIENCE PRESS |
| Citation statistics | |
| Document Type | 期刊论文 |
| Identifier | http://ir.imr.ac.cn/handle/321006/141419 |
| Collection | 中国科学院金属研究所 |
| Corresponding Author | Wang Jianqiu |
| Affiliation | 1.Huadian Elect Power Res Inst Co Ltd, Hangzhou 310030, Peoples R China 2.Chinese Acad Sci, Inst Met Res, Shenyang 110016, Peoples R China 3.State Nucl Power Plant Serv Co, Shanghai 200233, Peoples R China |
| Recommended Citation GB/T 7714 | Li Xiaohui,Wang Jianqiu,Han En-Hou,et al. Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water[J]. ACTA METALLURGICA SINICA,2020,56(11):1474-1484. |
| APA | Li Xiaohui,Wang Jianqiu,Han En-Hou,Guo Yanjun,Zheng Hui,&Yang Shuangliang.(2020).Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water.ACTA METALLURGICA SINICA,56(11),1474-1484. |
| MLA | Li Xiaohui,et al."Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water".ACTA METALLURGICA SINICA 56.11(2020):1474-1484. |
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