Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment

Wenjun Kuang; Miao Song; Chad M. Parish; Gary S. Was

doi:10.1007/978-3-319-67244-1_34

Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment

Wenjun Kuang
, Miao Song
, Chad M. Parish
, Gary S. Was

科研成果: 书/报告/会议事项章节 › 会议稿件 › 同行评审

4 引用（Scopus）

摘要

This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360, °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr₂O₃ which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr₂O₃ dominates at the oxide/substrate interfaces and the very tip region.

源语言	英语
主期刊名	Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors
编辑	Denise Paraventi, Michael Wright, John H. Jackson
出版商	Springer International Publishing
页	535-545
页数	11
ISBN（印刷版）	9783319672434
DOI	https://doi.org/10.1007/978-3-319-67244-1_34
出版状态	已出版 - 2018
已对外发布	是
活动	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017 - Portland, 美国期限: 13 8月 2017 → 17 8月 2017

出版系列

姓名	Minerals, Metals and Materials Series
卷	Part F9
ISSN（印刷版）	2367-1181
ISSN（电子版）	2367-1696

会议

会议	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017
国家/地区	美国
市	Portland
时期	13/08/17 → 17/08/17

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10.1007/978-3-319-67244-1_34

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引用此

Kuang, W., Song, M., Parish, C. M., & Was, G. S. (2018). Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. 在 D. Paraventi, M. Wright, & J. H. Jackson (编辑), Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (页码 535-545). (Minerals, Metals and Materials Series; 卷 Part F9). Springer International Publishing. https://doi.org/10.1007/978-3-319-67244-1_34

Kuang, Wenjun ; Song, Miao ; Parish, Chad M. 等. / Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. 编辑 / Denise Paraventi ; Michael Wright ; John H. Jackson. Springer International Publishing, 2018. 页码 535-545 (Minerals, Metals and Materials Series).

@inproceedings{db33caf5fc1d44d7a2737cbde154b05b,

title = "Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment",

abstract = "This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360, °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.",

keywords = "Alloy 690, Boundary migration, Intergranular oxidation, PWR, SCC",

author = "Wenjun Kuang and Miao Song and Parish, \{Chad M.\} and Was, \{Gary S.\}",

note = "Publisher Copyright: {\textcopyright} The Minerals, Metals \& Materials Society 2018.; 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017 ; Conference date: 13-08-2017 Through 17-08-2017",

year = "2018",

doi = "10.1007/978-3-319-67244-1\_34",

language = "英语",

isbn = "9783319672434",

series = "Minerals, Metals and Materials Series",

publisher = "Springer International Publishing",

pages = "535--545",

editor = "Denise Paraventi and Michael Wright and Jackson, \{John H.\}",

booktitle = "Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors",

}

Kuang, W, Song, M, Parish, CM & Was, GS 2018, Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. 在 D Paraventi, M Wright & JH Jackson (编辑), Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. Minerals, Metals and Materials Series, 卷 Part F9, Springer International Publishing, 页码 535-545, 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017, Portland, 美国, 13/08/17. https://doi.org/10.1007/978-3-319-67244-1_34

Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. / Kuang, Wenjun; Song, Miao; Parish, Chad M. 等.
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. 编辑 / Denise Paraventi; Michael Wright; John H. Jackson. Springer International Publishing, 2018. 页码 535-545 (Minerals, Metals and Materials Series; 卷 Part F9).

科研成果: 书/报告/会议事项章节 › 会议稿件 › 同行评审

TY - GEN

T1 - Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment

AU - Kuang, Wenjun

AU - Song, Miao

AU - Parish, Chad M.

AU - Was, Gary S.

N1 - Publisher Copyright: © The Minerals, Metals & Materials Society 2018.

PY - 2018

Y1 - 2018

N2 - This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360, °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.

AB - This study was aimed at investigating the intergranular attack near a stress corrosion crack (SCC) of alloy 690 in simulated pressurized water reactor (PWR) primary water environment. Solution annealed alloy 690 was evaluated for its SCC initiation susceptibility in 360, °C hydrogenated pure water using slow strain rate tensile technique. After the test, a grain boundary showing SCC initiation was sampled with Focused Ion Beam (FIB) milling. The microstructure and elemental distribution near the crack tip were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM). The results show that intergranular oxidation occurs ahead of the crack tip and is preceded by diffusion induced grain boundary migration. The oxides at the crack tip are mainly composed of NiO and Cr2O3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr2O3 dominates at the oxide/substrate interfaces and the very tip region.

KW - Alloy 690

KW - Boundary migration

KW - Intergranular oxidation

KW - PWR

KW - SCC

UR - https://www.scopus.com/pages/publications/85042498207

U2 - 10.1007/978-3-319-67244-1_34

DO - 10.1007/978-3-319-67244-1_34

M3 - 会议稿件

AN - SCOPUS:85042498207

SN - 9783319672434

T3 - Minerals, Metals and Materials Series

SP - 535

EP - 545

BT - Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors

A2 - Paraventi, Denise

A2 - Wright, Michael

A2 - Jackson, John H.

PB - Springer International Publishing

T2 - 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017

Y2 - 13 August 2017 through 17 August 2017

ER -

Kuang W, Song M, Parish CM, Was GS. Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. 在 Paraventi D, Wright M, Jackson JH, 编辑, Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. Springer International Publishing. 2018. 页码 535-545. (Minerals, Metals and Materials Series). doi: 10.1007/978-3-319-67244-1_34

Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment

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