带划伤缺陷690TT合金传热管在模拟压水堆二回路高温高压水环境中的腐蚀行为

许星星, 王龙, 张晓峰, 李振华, 施建辉, 尤磊, 丁清越, 陆永浩

doi:10.16577/j.issn.1001-1560.2025.0046

材料保护 >

2025 , Vol. 58 >Issue 3: 98 - 105

DOI: https://doi.org/10.16577/j.issn.1001-1560.2025.0046

研究论文

带划伤缺陷690TT合金传热管在模拟压水堆二回路高温高压水环境中的腐蚀行为

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¹阳江核电有限公司； ²核电安全技术与装备全国重点实验室； ³深圳中广核工程设计有限公司； ⁴北京科技大学国家材料服役安全科学中心

陆永浩(1967－)，博士，研究员，主要从事电站材料服役安全评价研究工作，E-mail: lu＿yonghao＠mater.ustb.edu.cn；张晓峰(1996－)，硕士，工程师，主要从事高温高压水环境下材料的应力腐蚀、腐蚀疲劳等工作，E-mail:zhangxiaofeng＠ustb.edu.cn

收稿日期: 2024-09-18

修回日期: 2024-11-06

录用日期: 2024-11-10

网络出版日期: 2025-04-16

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Study on the Corrosion Behavior of 690TT Alloy Heat Transfer Tubes with Scratch Defects in a Simulated High-Temperature and High-Pressure Water Environment of the Pressurized Water Reactor Secondary Circuit

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(1.Yangjiang Nuclear Power Co.， Ltd.， Yangjiang 529500， China；2.State Key Laboratory of Nuclear Power Safety Technology and Equipment， Shenzhen 518000， China；3.China Nuclear Power Design Co.， Ltd.(Shenzhen)， Shenzhen 518000， China；4.National Center for Materials Service Safety， University of Science and Technology Beijing， Beijing 100083， China)

LU Yonghao (1967－)， Ph.D.， Researcher， Research Focus: Evaluation of Service Safety of Power Station Materials， E-mail: lu＿yonghao＠mater.ustb.edu.cn；ZHANG Xiaofeng (1996－)，Master，Engineer，Research Focus:Stress Corrosion and Corrosion Fatigue of Materials in High Temperature and High-Pressure Water Environment， E-mail: zhangxiaofeng＠ustb.edu.cn

Received date: 2024-09-18

Revised date: 2024-11-06

Accepted date: 2024-11-10

Online published: 2025-04-16

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摘要

为了评估划伤缺陷对690TT 合金传热管在模拟二回路高温高压水环境中应力腐蚀性能的影响，以核电蒸汽发生器传热管为研究对象，采用划伤装置在管材表面制备出含划伤缺陷的C 型环试样，研究带划伤缺陷690TT 合金传热管在模拟二回路高温高压水环境中的应力腐蚀行为。采用SEM、TEM、EDS、拉曼光谱和白光干涉仪测试了试样在腐蚀前后的表面缺陷形貌、氧化形貌、氧化物元素分布和表面轮廓。结果表明:在模拟二回路高温高压水环境中，受应力加载的C 型环划痕槽表面上形成尖晶石氧化物和富铬氧化物，外表面尖晶石氧化物颗粒尺寸和数量随试验周期延长而增加。划伤形成过程会导致划伤槽区域组织出现机械微裂纹，经长期浸泡后微裂纹没有明显扩展，表明含划伤缺陷690TT 合金在二回路模拟环境下具有较高的抗应力腐蚀开裂能力。

关键词： 690TT 合金传热管; 划伤; 模拟二回路高温高压水; 应力腐蚀

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许星星, 王龙, 张晓峰, 李振华, 施建辉, 尤磊, 丁清越, 陆永浩 . 带划伤缺陷690TT合金传热管在模拟压水堆二回路高温高压水环境中的腐蚀行为[J]. 材料保护, 2025 , 58(3) : 98 -105 . DOI: 10.16577/j.issn.1001-1560.2025.0046

Abstract

In order to evaluate the effect of scratch defects on the stress corrosion performance of 690TT alloy heat transfer tubes in a simulated high-temperature and high-pressure water environment of the secondary circuit， a nuclear power steam generator heat transfer tube was chosen as the research subject.A scratch device was used to create C-ring specimens with scratch defects on the surface of the tube.The stress corrosion behavior of the scratched 690TT alloy heat transfer tubes was investigated in the simulated secondary circuit high-temperature and highpressure water environment.The surface defect morphology， oxidation morphology， distribution of oxide elements and surface profile of the specimens before and after corrosion were analyzed using scanning electron microscopy (SEM)， transmission electron microscopy (TEM)， energy dispersive spectrometer (EDS)，Raman spectroscopy and white light interferometer.Results showed that in the simulated secondary circuit high-temperature and high-pressure water environment， spinel oxide and chromium-rich oxide were formed on the surface of the C-ring scratch groove under stress loading.Additionally， the particle size and quantity of the spinel oxide on the outer surface increased with the test duration.The scratching process led to the formation of mechanical microcracks in the scratch groove area， but after long-term immersion， these microcracks did not show significant propagation， indicating that the scratched 690TT alloy has a high resistance to stress corrosion cracking in the secondary circuit simulated environment.

Key words： 690TT alloy heat transfer tube; scratch; simulated secondary circuit high-temperature and high-pressure water; stress corrosion

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