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:10.1007/978-3-030-04639-2_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

Nuclear Energy Materials Microanalysis

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

2 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 Cr ₂ O ₃ which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr ₂ O ₃ dominates at the oxide/substrate interfaces and the very tip region.

Original language	English
Title of host publication	Minerals, Metals and Materials Series
Publisher	Springer International Publishing
Pages	535-545
Number of pages	11
ISBN (Print)	9783030046385, 9783030046392, 9783319515403, 9783319651354, 9783319728520, 9783319950211
DOIs	https://doi.org/10.1007/978-3-030-04639-2_34
State	Published - 2019
Event	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019 - Boston, United States Duration: Aug 18 2019 → Aug 22 2019

Publication series

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

Conference

Conference	18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019
Country/Territory	United States
City	Boston
Period	08/18/19 → 08/22/19

Funding

The authors gratefully acknowledge financial support through the DOE I-NERI program contract 2011-01-K. The authors would like to thank Young Suk Kim and Sung Soo Kim from Korea Atomic Energy Research Institute for providing the materials for this study, and Alex Flick from the University of Michigan for his assistance with preparation of the high temperature autoclave systems. This research was performed, in part, using instrumentation provided by the Department of Energy, Office of Nuclear Energy, Fuel Cycle R&D Program and the Nuclear Science User Facilities.

Funders	Funder number
U.S. Department of Energy	2011-01-K

Keywords

Alloy 690
Boundary migration
Intergranular oxidation
PWR
SCC

Access to Document

10.1007/978-3-030-04639-2_34

Cite this

@inproceedings{af8886302f864e7eaec20ad10a822091,

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 Cr 2 O 3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr 2 O 3 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.}",

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language = "English",

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series = "Minerals, Metals and Materials Series",

publisher = "Springer International Publishing",

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Kuang, W, Song, M, Parish, CM & Was, GS 2019, Microstructural study on the stress corrosion cracking of alloy 690 in simulated pressurized water reactor primary environment. in Minerals, Metals and Materials Series. Minerals, Metals and Materials Series, Springer International Publishing, pp. 535-545, 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors 2019, Boston, United States, 08/18/19. https://doi.org/10.1007/978-3-030-04639-2_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.
Minerals, Metals and Materials Series. Springer International Publishing, 2019. p. 535-545 (Minerals, Metals and Materials Series).

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.

PY - 2019

Y1 - 2019

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 Cr 2 O 3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr 2 O 3 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 Cr 2 O 3 which maintain rigid orientations with the neighboring grains. The adjacent migration zone is free of oxidization as a compact layer of Cr 2 O 3 dominates at the oxide/substrate interfaces and the very tip region.

KW - Alloy 690

KW - Boundary migration

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Y2 - 18 August 2019 through 22 August 2019

ER -

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

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