Corrosion Behaviors of Candidate Materials for Supercritical-Cooled Water Reactor

ZHU Fa-wen; ZHANG Le-fu; QIAO Pei-peng; LIU Rui-qin; BAO Yi-chen; CHEN Yu-qing

Volume 30 Issue 5

Oct. 2009

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ZHU Fa-wen, ZHANG Le-fu, QIAO Pei-peng, LIU Rui-qin, BAO Yi-chen, CHEN Yu-qing. Corrosion Behaviors of Candidate Materials for Supercritical-Cooled Water Reactor[J]. Nuclear Power Engineering, 2009, 30(5): 62-66.

Citation:

ZHU Fa-wen, ZHANG Le-fu, QIAO Pei-peng, LIU Rui-qin, BAO Yi-chen, CHEN Yu-qing. Corrosion Behaviors of Candidate Materials for Supercritical-Cooled Water Reactor[J]. Nuclear Power Engineering, 2009, 30(5): 62-66.

Citation:

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Corrosion Behaviors of Candidate Materials for Supercritical-Cooled Water Reactor

School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai, 200240, China

Received Date: 2008-07-31
Rev Recd Date: 2009-06-30

Available Online: 2025-07-28

Publish Date: 2009-10-15

Abstract

Abstract

The corrosion behaviors of ferritic-martensitic steel P92, austenitic stainless steel type 316L and nickel-based alloy 690 have been investigated in supercritical water （SCW） at 600℃ and 23 MPa. After exposed to SCW for 625 h, the weight gain of alloy 690, 316L and P92 are 0.001 02 g/（m²·h）, 0.060 6 g/（m²·h） and 0.101 27g/（m²·h）, respectively. SEM observation shows that a three-layer oxide film is formed on P92, a single-layer oxide film is formed on 316L and a very thin oxide film with some pitting is formed on alloy 690.
- Stainless steels,
- Nickel-base alloys,
- Supercritical water,
- Oxidation layer,
- General corrosion