Stress Corrosion Cracking Behavior of Forged 316L Stainless Steel Used for Nuclear Power Plants in Alkaline Solution at 330 ℃

doi:10.11902/1005.4537.2014.073

Journal of Chinese Society for Corrosion and protection

2015, Vol. 35

Issue (3): 205-212 DOI: 10.11902/1005.4537.2014.073

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Stress Corrosion Cracking Behavior of Forged 316L Stainless Steel Used for Nuclear Power Plants in Alkaline Solution at 330 ℃

Zhiming ZHANG¹,Qingjiao PENG^1,²,Jianqiu WANG¹(

),En-Hou HAN¹,Wei KE¹

1. Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, Institute of Metal Research,Chinese Academy of Sciences, Shenyang 110016, China
2. Zoomlion Heavy Industry Science &Technology Development Co., Ltd., Changsha 410007, China

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Abstract

The stress corrosion cracking behavior of forged 316L stainless steel used for the main pipe of pressured water reactors was investigated in sodium hydroxide solution at 330 ^oC using U-bent samples. The results showed that, after immersion for 720 h, obvious stress corrosion cracks were found on the sample surfaces. The formed surface oxide film consisted of an inner layer rich in Fe, a middle layer rich in Ni and an outmost layer of oxide particles rich in Fe and Ni, which could not protect the base metal from further corrosion. EBSD and fracture morphology observation showed that the steel suffered mainly from intergranular cracking and the fractured surface exhibitedmainly rock candy-like patternwith partly river-like pattern and quasi-cleavage steps. The cracks were propagated by the brittle fracture of grain boundary oxides rich in Fe and Ni under the external stress. The stress corrosion cracking mechanism was ascribed to anodic dissolution-brittle fracture of grain boundary oxides by the applied stress.

Key words: 316L stainless steel high temperature and high pressure NaOH stress corrosion cracking

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	Zhiming ZHANG
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	Wei KE

Cite this article:

Zhiming ZHANG,Qingjiao PENG,Jianqiu WANG,En-Hou HAN,Wei KE. Stress Corrosion Cracking Behavior of Forged 316L Stainless Steel Used for Nuclear Power Plants in Alkaline Solution at 330 ℃. Journal of Chinese Society for Corrosion and protection, 2015, 35(3): 205-212.

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https://www.jcscp.org/EN/10.11902/1005.4537.2014.073 OR https://www.jcscp.org/EN/Y2015/V35/I3/205

Fig.1 Schematic drawings of the 316L stainless steel main pipe and the sampling position of the U-bent sample

Fig.2 Three-dimensional metallographic image of the sample of 316L stainless steel

Fig.3 Dimensions of the U-bent sample used for stress corrosion cracking test

Fig.4 Macrograph (a) and micrograph (b) of 316L U-bent sample after immersion in 4%NaOH solution at 330 ℃ for 720 h

Fig.5 Cross-sectional morphology (a) and chemical compositions (b) of the surface oxide film formed on 316L U-bent sample after immersion in 4%NaOH solution at 330 ℃ for 720 h

Fig.6 Cross-sectional images of the 316L U-bent sample after immersion in 4%NaOH solution at 330 ℃ for 720 h (a) and the magnified image of Fig.6a (b)

Fig.7 Images of the crack paths (a), grain boundaries (b) and grains (c) of 316L stainless steel U-bent sample after immersionin 4%NaOH solution at 330 ℃ for 720 h (GB: grain boundary, LAB: low angle boundary, CSL: coincidence site lattice grain boundary, RGB: random grain boundary)

Fig.8 OM (a) and SEM (b, c) images of fracture surface of 316L U-bent sample after immersion in 4%NaOH solution at 330 ℃ for 720 h

Fig.9 SEM images of the crack paths (a, b) and included oxides (c) existed in the tested sample; elemental line scannings along the lines 1 (d) and 2 (e) in Fig.9c, respectively

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