Passivity degradation of nuclear steam generator tubing alloy induced by Pb contamination at high temperature

被引:27
|
作者
Lu, B. T. [1 ]
Luo, J. L. [1 ]
Lu, Y. C. [2 ]
机构
[1] Univ Alberta, Dept Chem & Mat Engn, Edmonton, AB T6G 2G6, Canada
[2] Atom Energy Canada Ltd, Chalk River Labs, Chalk River, ON K0J 1J0, Canada
基金
加拿大自然科学与工程研究理事会;
关键词
STRESS-CORROSION CRACKING; POINT-DEFECT MODEL; IN-SITU STM; ACID-SOLUTION; BASE ALLOYS; FILMS; HYDROGEN; IRON; SURFACE; NICKEL;
D O I
10.1016/j.jnucmat.2012.06.021
中图分类号
T [工业技术];
学科分类号
08 ;
摘要
Effects of Pb contamination on the passivity of a Ni-based alloy (UNS N06690) in a simulated crevice chemistry of steam generator with near-neutral pH at 300 degrees C are elucidated using electrochemical measurements and surface analysis techniques. The experimental observations reveal that Pb impurity can enter anodic film, which results in substantial changes in the films structure via hindering the dehydration during the passivation and retarding the formation of spinel oxides. The presence of Pb-contamination can also increase hydrogen content in anodic film. Finally, the mechanism of passivity degradation induced by Pb contamination is described on the basis of the experimental data and established theory. (C) 2012 Elsevier B.V. All rights reserved.
引用
收藏
页码:305 / 314
页数:10
相关论文
共 50 条
  • [1] Effects of pH on lead-induced passivity degradation of nuclear steam generator tubing alloy in high temperature crevice chemistries
    Lu, B. T.
    Luo, J. L.
    Lu, Y. C.
    ELECTROCHIMICA ACTA, 2013, 87 : 824 - 838
  • [2] Effects of dissolved calcium and magnesium ions on lead-induced stress corrosion cracking susceptibility of nuclear steam generator tubing alloy in high temperature crevice solutions
    Lu, B. T.
    Tian, L. P.
    Zhu, R. K.
    Luo, J. L.
    Lu, Y. C.
    ELECTROCHIMICA ACTA, 2011, 56 (04) : 1848 - 1855
  • [3] Inconel 690 is alloy of choice for steam generator tubing
    Strauss, SD
    POWER, 1996, 140 (02) : 29 - &
  • [4] SCC testing of nuclear steam generator tubing materials
    Babcock and Wilcox International Division, United States
    JOM, 5 (39-41):
  • [5] The SCC testing of nuclear steam generator tubing materials
    Doherty, PE
    Sarver, JM
    Miglin, BP
    JOM-JOURNAL OF THE MINERALS METALS & MATERIALS SOCIETY, 1996, 48 (05): : 39 - 41
  • [6] Stress corrosion cracking aspects of nuclear steam generator tubing materials in the water containing lead at high temperature
    Hwang, SS
    Kim, KM
    Kim, UC
    PROCEEDINGS OF THE EIGHTH INTERNATIONAL SYMPOSIUM ON ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMS - WATER REACTORS, VOLS 1 AND 2, 1997, : 200 - +
  • [7] Mechanical/electrochemical performance of alloy 690 steam generator tubing
    Doherty, PE
    Psaila-Dombrowski, MJ
    Miglin, BP
    Sarver, JM
    Doyle, DM
    PROCEEDINGS OF THE EIGHTH INTERNATIONAL SYMPOSIUM ON ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMS - WATER REACTORS, VOLS 1 AND 2, 1997, : 157 - +
  • [8] The estimation of lifetime distribution of Alloy 800 steam generator tubing
    Pandey, M. D.
    Datla, S.
    Tapping, R. L.
    Lu, Y. C.
    NUCLEAR ENGINEERING AND DESIGN, 2009, 239 (10) : 1862 - 1869
  • [9] Evaluation of Caustic Stress Corrosion Resistance of Steam Generator Tubing Alloy 690 for Nuclear Power Plant
    Tang Zhanmei
    Meng Fanjiang
    Zhang Pingzhu
    Xu Xuelian
    Hu Shilin
    RARE METAL MATERIALS AND ENGINEERING, 2019, 48 (11) : 3541 - 3547
  • [10] Evaluation of Caustic Stress Corrosion Resistance of Steam Generator Tubing Alloy 690 for Nuclear Power Plant
    Tang, Zhanmei
    Meng, Fanjiang
    Zhang, Pingzhu
    Xu, Xuelian
    Hu, Shilin
    Xiyou Jinshu Cailiao Yu Gongcheng/Rare Metal Materials and Engineering, 2019, 48 (11): : 3541 - 3547