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

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution › peer-review

4 Scopus citations

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.

Original language	English
Title of host publication	Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors
Editors	Denise Paraventi, Michael Wright, John H. Jackson
Publisher	Springer International Publishing
Pages	535-545
Number of pages	11
ISBN (Print)	9783319672434
DOIs	https://doi.org/10.1007/978-3-319-67244-1_34
State	Published - 2018
Externally published	Yes
Event	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017 - Portland, United States Duration: 13 Aug 2017 → 17 Aug 2017

Publication series

Name	Minerals, Metals and Materials Series
Volume	Part F9
ISSN (Print)	2367-1181
ISSN (Electronic)	2367-1696

Conference

Conference	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017
Country/Territory	United States
City	Portland
Period	13/08/17 → 17/08/17

Keywords

Alloy 690
Boundary migration
Intergranular oxidation
PWR
SCC

Access to Document

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

Cite this

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. In D. Paraventi, M. Wright, & J. H. Jackson (Eds.), Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors (pp. 535-545). (Minerals, Metals and Materials Series; Vol. Part F9). Springer International Publishing. https://doi.org/10.1007/978-3-319-67244-1_34

Kuang, Wenjun ; Song, Miao ; Parish, Chad M. et al. / 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. editor / Denise Paraventi ; Michael Wright ; John H. Jackson. Springer International Publishing, 2018. pp. 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. in D Paraventi, M Wright & JH Jackson (eds), Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. Minerals, Metals and Materials Series, vol. Part F9, Springer International Publishing, pp. 535-545, 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, 2017, Portland, United States, 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. et al.
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. ed. / Denise Paraventi; Michael Wright; John H. Jackson. Springer International Publishing, 2018. p. 535-545 (Minerals, Metals and Materials Series; Vol. Part F9).

Research output: Chapter in Book/Report/Conference proceeding › Conference contribution › peer-review

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. In Paraventi D, Wright M, Jackson JH, editors, Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors. Springer International Publishing. 2018. p. 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

Abstract

Publication series

Conference

Keywords

Access to Document

Other files and links

Fingerprint

Cite this