Investigation on Corrosion Model of Cr-coated Zirconium Alloy Cladding

被引：0
|
作者：
Shen, Yong [1 ,2 ]
Zeng, Xiehu [1 ,2 ]
Duan, Zhengang [1 ,2 ,3 ]
Wen, Qinglong [1 ,2 ,3 ]
Yuan, Bo [1 ,2 ]
He, Liang [4 ]
Gao, Shixin [4 ]
机构：
[1] School of Energy and Power Engineering, Chongqing University, Chongqing,400044, China
[2] Liangjiang Laboratory for New Energy (Nuclear Energy and Power), Chongqing,400044, China
[3] Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Ministry of Education, Chongqing University, Chongqing,400044, China
[4] Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu,610213, China
来源：
Hedongli Gongcheng/Nuclear Power Engineering | 2024年 / 45卷
关键词：
Corrosion rate - Corrosion resistant alloys - Corrosion resistant coatings - Corrosive effects - Lead alloys - Metal cladding - Nuclear energy - Nuclear reactor accidents - Oxidation resistance - Zirconium alloys;
D O I：
10.13832/j.jnpe.2024.S1.0175
中图分类号：
学科分类号：
摘要：
As one of the candidate materials for accident tolerant fuel (ATF) cladding, Cr coating can significantly improve the corrosion resistance and oxidation resistance of zirconium alloy cladding, which is expected to prolong the service life. In order to evaluate the corrosion and oxidation behavior of Cr-coated zirconium alloy cladding, a corrosion model of Cr-coated zirconium alloy cladding under normal operating conditions of PWR was established in this paper, and the model was verified based on the experimental data in the literature. Based on this model, the effects of heat flux and mass flow rate on the corrosion of Cr-coated zirconium alloy cladding were analyzed. The results show that the corrosion thickness increases with the increase of heat flux. In addition, the increase of coolant mass flow rate leads to the decrease of cladding wall temperature, which eventually leads to the decrease of cladding corrosion thickness. © 2024 Atomic Energy Press. All rights reserved.
引用
收藏
页码：175 / 180
相关论文
共 50 条
  • [1] Current progress of Cr-coated zirconium alloy cladding for accident tolerant fuel
    Duan Z.
    Chen P.
    Zhou Y.
    Gao S.
    Jiao Y.
    Yin C.
    He L.
    He Jishu/Nuclear Techniques, 2022, 45 (03):
  • [2] Experimental Study of Cr-coated Zirconium Alloy Cladding under Simulated LOCA Conditions
    Wang, Zhanwei
    Yan, Jun
    Peng, Zhenxun
    Ren, Qisen
    Liao, Yehong
    Li, Sigong
    Zhao, Yahuan
    Hedongli Gongcheng/Nuclear Power Engineering, 2023, 44 (02): : 122 - 128
  • [3] Improving fatigue performance of Cr-coated zirconium alloy cladding by ultrasonic shot-peening proces
    Wang, Yong-jing
    Li, Si -gong
    Liu, Yang
    Zhang, Zhi-jia
    Feng, Rui-xuan
    Wang, Qu-ge
    Zhang, Qian-cheng
    JOURNAL OF MATERIALS RESEARCH AND TECHNOLOGY-JMR&T, 2023, 27 : 6047 - 6057
  • [4] Corrosion behavior of Cr-coated zirconium alloy cladding in LiOH/H3BO3-containing water at 360 °C
    Ma, Haibin
    Zhang, Honglin
    Hu, Lijuan
    Li, Sigong
    Zhang, Jianan
    Zhang, Wenhuai
    Sun, Rongrong
    Yao, Meiyi
    Xue, Jiaxiang
    Ren, Qisen
    Liao, Yehong
    CORROSION SCIENCE, 2023, 222
  • [5] Radiation-Induced Sharpening in Cr-Coated Zirconium Alloy
    Ribis, Joel
    Wu, Alexia
    Guillou, Raphaelle
    Brachet, Jean-Christophe
    Baumier, Cedric
    Gentils, Aurelie
    Loyer-Prost, Marie
    MATERIALS, 2022, 15 (06)
  • [6] Experimental Investigation for the Fretting Wear of Cr-coated Zircaloy Cladding
    Yang, Siyuan
    Yuan, Bo
    Wen, Qinglong
    Wen, Shuang
    Zhang, Ruiqian
    Yang, Hongyan
    Hedongli Gongcheng/Nuclear Power Engineering, 2024, 45 : 103 - 109
  • [7] Oxidation behavior of Cr-coated zirconium alloy cladding in high-temperature steam above 1200 °C
    Ma, Hai-Bin
    Yan, Jun
    Zhao, Ya-Huan
    Liu, Tong
    Ren, Qi-Sen
    Liao, Ye-Hong
    Zuo, Jia-Dong
    Liu, Gang
    Yao, Mei-Yi
    NPJ MATERIALS DEGRADATION, 2021, 5 (01)
  • [8] Experimental Study on Pool Boiling Heat Transfer of Cr-coated Zirconium Cladding
    Zeng X.
    Chen Z.
    Wen Q.
    Du Q.
    Zhang R.
    Du P.
    Hedongli Gongcheng/Nuclear Power Engineering, 2023, 44 (02): : 91 - 97
  • [9] Oxidation behavior of Cr-coated zirconium alloy cladding in high-temperature steam above 1200 °C
    Hai-Bin Ma
    Jun Yan
    Ya-Huan Zhao
    Tong Liu
    Qi-Sen Ren
    Ye-Hong Liao
    Jia-Dong Zuo
    Gang Liu
    Mei-Yi Yao
    npj Materials Degradation, 5
  • [10] Influence of Neutron Irradiation on Mechanical Properties of Cr-coated Zirconium Alloy
    Wu Y.
    Xi H.
    Li G.
    Liu X.
    Zhang H.
    Sun K.
    Ning Z.
    Fang Z.
    Liu S.
    Hedongli Gongcheng/Nuclear Power Engineering, 2023, 44 (02): : 116 - 121
12345