Early corrosion behaviour of irradiated FeCrAl alloy in a simulated pressurized water reactor environment

被引：28

作者：

Wu, Weisong ^{[1
,2
]}

Ran, Guang ^{[1
,2
]}

Li, Yipeng ^{[1
,2
]}

Cong, Shuo ^{[1
,2
]}

Ye, Chao ^{[1
,2
]}

Zhang, Ruiqian ^{[3
]}

Sun, Yongduo ^{[3
]}

机构：

[1] Xiamen Univ, Coll Energy, Xiamen 361102, Fujian, Peoples R China

[2] Fujian Res Ctr Nucl Engn, Xiamen 361102, Fujian, Peoples R China

[3] Nucl Power Inst China, Sci & Technol Reactor Fuel & Mat Lab, Chengdu 610041, Peoples R China

来源：

CORROSION SCIENCE | 2020年 / 174卷

基金：

中国国家自然科学基金;

关键词：

FeCrAl alloy; Ion implantation; High temperature corrosion; Oxidation; TEMPERATURE AQUEOUS CORROSION; DEGREES-C EMBRITTLEMENT; STAINLESS-STEEL; OXIDE-FILM; 288-DEGREES-C WATER; OXYGEN CONCENTRATION; DISSOLVED-OXYGEN; MECHANISM; HYDROGEN; AL;

D O I：

10.1016/j.corsci.2020.108824

中图分类号：

T [工业技术];

学科分类号：

08 ;

摘要：

The early corrosion behaviour of an irradiated FeCrAl alloy in a simulated pressurized water reactor environment was investigated. The irradiated alloy showed a Cr-rich oxide layer, which was not observed in the asreceived alloy. The corrosion induced the formation of a duplex-layer oxide structure consisting of an outer hematite alpha-Fe2O3 layer and an inner spinel Cr-rich oxide layer. In the initial corrosion stage, the smooth surface formed by ion sputtering and the formation of the Cr-rich oxide layer suppressed the corrosion. However, in the later corrosion stage, the irradiated alloys exhibited severe corrosion because of the presence of irradiation-induced defects.

引用

页数：10

共 50 条

[1] Effects of precipitation on the stress corrosion crack initiation of alloy 718 in simulated pressurized water reactor primary environment
Ning, Zehao
Zhang, Haoran
Zhang, Shihao
Hao, Xianchao
Han, En-Hou
Kuang, Wenjun
[J]. ACTA MATERIALIA, 2024, 276
[2] Effects of temperature and pH on the electrochemical behaviour of alloy 600 in simulated pressurized water reactor primary water
Yang, Jie
Li, Yanhui
Macdonald, Digby D.
[J]. JOURNAL OF NUCLEAR MATERIALS, 2020, 528
[3] The oxidation of alloy 690 in simulated pressurized water reactor primary water
Kuang, Wenjun
Song, Miao
Wang, Peng
Was, Gary S.
[J]. CORROSION SCIENCE, 2017, 126 : 227 - 237
[4] High-resolution characterization of the fretting corrosion of Alloy 690 in the simulated secondary water of pressurized water reactor
Wang, You
Zhou, Zhangjian
Wang, Hui
Ju, Jiang
Shen, Zhao
Zeng, Xiaoqin
[J]. JOURNAL OF MATERIALS SCIENCE & TECHNOLOGY, 2023, 151 : 30 - 40
[5] Early oxidation behavior of Ni-based Alloy 690 in simulated pressurized water reactor environment with varying surface conditions
Min, Shiling
Liu, Hui
Wang, Jianqiu
Hou, Juan
[J]. CORROSION COMMUNICATIONS, 2021, 4 : 68 - 81
[6] Early oxidation behavior of Ni-based Alloy 690 in simulated pressurized water reactor environment with varying surface conditions
Shiling Min
Hui Liu
Jianqiu Wang
Juan Hou
[J]. Corrosion Communication, 2021, 4 (04) : 68 - 81
[7] Early oxidation behavior of Ni-based Alloy 690 in simulated pressurized water reactor environment with varying surface conditions
Shiling Min
Hui Liu
Jianqiu Wang
Juan Hou
[J]. Corrosion Communications, 2021, 4 (04) : 68 - 81
[8] STRESS-CORROSION CRACKING OF REACTOR PRESSURE-VESSEL MATERIALS IN A SIMULATED PRESSURIZED WATER-REACTOR ENVIRONMENT
BULLOCH, JH
[J]. JOURNAL OF THE SOUTH AFRICAN INSTITUTE OF MINING AND METALLURGY, 1987, 87 (08): : 233 - 242
[9] Advanced TEM characterization of stress corrosion cracking of Alloy 600 in pressurized water reactor primary water environment
Sennour, M.
Laghoutaris, P.
Guerre, C.
Molins, R.
[J]. JOURNAL OF NUCLEAR MATERIALS, 2009, 393 (02) : 254 - 266
[10] A study of the oxidation of FeCrAl alloy in pressurized water and high-temperature steam environment
Park, Dong Jun
Kim, Hyun Gil
Park, Jeong Yong
Jung, Yang Il
Park, Jeong Hwan
Koo, Yang Hyun
[J]. CORROSION SCIENCE, 2015, 94 : 459 - 465

← 1 2 3 4 5 →