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Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment
L. T. Zhang; J. Q. Wang
2014
Source PublicationJournal of Nuclear Materials
ISSN0022-3115
Volume446Issue:1-3Pages:15-26
AbstractStress corrosion crack growth tests of a cold worked nuclear grade 316L stainless steel were conducted in simulated pressurized water reactor (PWR) primary water environment containing various dissolved oxygen (DO) contents but no dissolved hydrogen. The crack growth rate (CGR) increased with increasing DO content in the simulated PWR primary water. The fracture surface exhibited typical intergranular stress corrosion cracking (IGSCC) characteristics. (C) 2013 Elsevier B.V. All rights reserved.
description.department[zhang, litao ; wang, jianqiu] chinese acad sci, inst met res, state key lab corros & protect, shenyang 110016, peoples r china. ; wang, jq (reprint author), chinese acad sci, inst met res, state key lab corros & protect, shenyang 110016, peoples r china. ; wangjianqiu@imr.ac.cn
KeywordHigh-temperature Water Assisted Cracking Alloy 690tt Pure Water Growth Behavior Dependence Chemistry Coolant Model
URL查看原文
Language英语
Document Type期刊论文
Identifierhttp://ir.imr.ac.cn/handle/321006/72600
Collection中国科学院金属研究所
Recommended Citation
GB/T 7714
L. T. Zhang,J. Q. Wang. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment[J]. Journal of Nuclear Materials,2014,446(1-3):15-26.
APA L. T. Zhang,&J. Q. Wang.(2014).Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment.Journal of Nuclear Materials,446(1-3),15-26.
MLA L. T. Zhang,et al."Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment".Journal of Nuclear Materials 446.1-3(2014):15-26.
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