研究论文

Fe13Cr5Al4Mo合金在高温高压水环境中的腐蚀行为

  • 林晓冬 ,
  • 马海滨 ,
  • 任啟森 ,
  • 孙蓉蓉 ,
  • 张文怀 ,
  • 胡丽娟 ,
  • 梁雪 ,
  • 李毅丰 ,
  • 姚美意
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  • 1.上海大学 材料研究所 上海 200072
    2.中广核研究院有限公司 核燃料与材料研究所 深圳 518026
    3.上海大学 微结构重点实验室 上海 200444
林晓冬,男,1991年生,博士

收稿日期: 2021-12-22

  修回日期: 2022-03-27

  网络出版日期: 2022-05-05

基金资助

国家自然科学基金项目(51871141)

Corrosion Behaviors of Fe13Cr5Al4Mo Alloy in High-Temperature High-Pressure Water Environments

  • Xiaodong LIN ,
  • Haibin MA ,
  • Qisen REN ,
  • Rongrong SUN ,
  • Wenhuai ZHANG ,
  • Lijuan HU ,
  • Xue LIANG ,
  • Yifeng LI ,
  • Meiyi YAO
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  • 1.Institute of Materials, Shanghai University, Shanghai 200072, China
    2.Nuclear Fuel and Materials Department, China Nuclear Power Technology Research Institute, Shenzhen 518026, China
    3.Laboratory for Microstructures, Shanghai University, Shanghai 200444, China
YAO Meiyi, professor, Tel: (021)56338586, E-mail: yaomeiyi@shu.edu.cn
MA Haibin, Tel: (0755)88617459, E-mail: mahaibin@cgnpc.com.cn;

Received date: 2021-12-22

  Revised date: 2022-03-27

  Online published: 2022-05-05

Supported by

National Natural Science Foundation of China(51871141)

摘要

利用XRD、SEM和TEM等测试技术研究了Fe13Cr5Al4Mo合金在360℃、18.6 MPa去离子水和360℃、18.6 MPa、3.5 mg/L Li + 1000 mg/L B水溶液中的腐蚀行为。结果表明,Fe13Cr5Al4Mo合金的腐蚀增重远低于参比锆合金,且腐蚀增重速率较慢,表明Fe13Cr5Al4Mo合金的耐腐蚀性能优于参比锆合金。在2种水环境下,Fe13Cr5Al4Mo合金表面均形成了一层Fe(Cr, Al)2O4纳米尖晶石结构的氧化膜,在去离子水中还产生了Fe3O4外层氧化颗粒。Fe(Cr, Al)2O4尖晶石氧化膜整体上较为致密,可以阻碍氧离子和金属阳离子在氧化膜内的扩散,提高合金的耐腐蚀性能。此外,高温高压水中添加Li + B导致Fe13Cr5Al4Mo合金的腐蚀增重和氧化膜厚度发生变化,且抑制了外层氧化颗粒的产生,这与Li + B水溶液较高的pH值和Li+、B3+的相互作用有关。

本文引用格式

林晓冬 , 马海滨 , 任啟森 , 孙蓉蓉 , 张文怀 , 胡丽娟 , 梁雪 , 李毅丰 , 姚美意 . Fe13Cr5Al4Mo合金在高温高压水环境中的腐蚀行为[J]. 金属学报, 2022 , 58(12) : 1611 -1622 . DOI: 10.11900/0412.1961.2021.00574

Abstract

FeCrAl alloys are promising candidate materials for accident-tolerant-fuel (ATF) claddings owing to their good high-temperature mechanical property, irradiation-swelling resistance, and high-temperature steam-oxidation performance. However, excellent corrosion resistance is also required in high-temperature high-pressure water environments when the alloys are used as ATF claddings. Therefore, in this work, the corrosion behavior of a Fe13Cr5Al4Mo alloy in 360oC, 18.6 MPa deionized water and 360oC, 18.6 MPa, 3.5 mg/L Li + 1000 mg/L B aqueous solution was studied. Results revealed that the weight gain and growth rate of the Fe13Cr5Al4Mo alloy were lower than that of the reference zirconium alloy, indicating a better corrosion property of the Fe13Cr5Al4Mo alloy. Moreover, an oxide film comprising Fe(Cr, Al)2O4 nanospinels formed on the Fe13Cr5Al4Mo alloy in both water environments, and Fe3O4 outer-oxide particles were observed in deionized water. The good corrosion performance of Fe13Cr5Al4Mo alloy was attributed to the compact spinel-oxide film, which could inhibit the diffusion of oxygen ions and metal cations. Adding Li + B into water changed the corrosion weight gain and oxide-film thickness of the Fe13Cr5Al4Mo alloy and impeded the formation of outer-oxide particles, which may be related to the high pH of alkaline Li + B aqueous solution and the interactions between Li+ and B3+.

参考文献

1 Charit I. Accident tolerant nuclear fuels and cladding materials [J]. JOM, 2018, 70: 173
2 Zinkle S J, Terrani K A, Gehin J C, et al. Accident tolerant fuels for LWRs: A perspective [J]. J. Nucl. Mater., 2014, 448: 374
3 Field K G, Yamamoto Y, Pint B A, et al. Accident tolerant FeCrAl fuel cladding: Current status towards commercialization [A]. Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. New York: Springer, 2019: 1381
4 Huang X, Li X Y, Fang X D, et al. Research progress in FeCrAl alloys for accident-tolerant fuel cladding [J]. J. Mater. Eng., 2020, 48(3): 19
4 黄希, 李小燕, 方晓东 等. 容错事故燃料包壳用FeCrAl合金的研究进展 [J]. 材料工程, 2020, 48(3): 19
5 Tao X K, Huang Z G, Guo Q M, et al. Research progress of FeCrAl alloy for cladding material of new type of light water reactor [J]. Hot Work. Technol., 2018, 47(6): 23
5 陶小康, 黄重国, 郭青苗 等. 新型轻水反应堆包壳材料FeCrAl合金的研究进展 [J]. 热加工工艺, 2018, 47(6): 23
6 Zhang Y Y, Wang H, An X G, et al. Dynamic strain aging behavior of accident tolerance fuel cladding FeCrAl-based alloy for advanced nuclear energy [J]. J. Mater. Sci., 2021, 56: 8815
7 Opila E J, Myers D L. Alumina volatility in water vapor at elevated temperatures [J]. J. Am. Ceram. Soc., 2004, 87: 1701
8 Pan D, Zhang R Q, Wang H, et al. Formation and stability of oxide layer in FeCrAl fuel cladding material under high-temperature steam [J]. J. Alloys Compd., 2016, 684: 549
9 Park D J, Kim H G, Park J Y, et al. A study of the oxidation of FeCrAl alloy in pressurized water and high-temperature steam environment [J]. Corros. Sci., 2015, 94: 459
10 Parker S S, White J, Hosemann P, et al. Oxidation kinetics of ferritic alloys in high-temperature steam environments [J]. JOM, 2018, 70: 186
11 Pint B A. Performance of FeCrAl for accident-tolerant fuel cladding in high-temperature steam [J]. Corros. Rev., 2017, 35: 167
12 Stott F H, Wood G C, Stringer J. The influence of alloying elements on the development and maintenance of protective scales [J]. Oxid. Met., 1995, 44: 113
13 Yamamoto Y, Pint B A, Terrani K A, et al. Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors [J]. J. Nucl. Mater., 2015, 467: 703
14 Rebak R B. Versatile oxide films protect FeCrAl alloys under normal operation and accident conditions in light water power reactors [J]. JOM, 2018, 70: 176
15 Rebak R B, Larsen M, Kim Y J. Characterization of oxides formed on iron-chromium-aluminum alloy in simulated light water reactor environments [J]. Corros. Rev., 2017, 35: 177
16 Ning F Q, Wang X, Yang Y, et al. Uniform corrosion behavior of FeCrAl alloys in borated and lithiated high temperature water [J]. J. Mater. Sci. Technol., 2021, 70: 136
17 Song L J, Liu F H, Li C T, et al. Effect of B-Li water chemistry on corrosion of metal materials of nuclear power plant [J]. Nucl. Sci. Eng., 2014, 34: 97
17 宋利君, 刘飞华, 李成涛 等. B-Li水化学对核电站金属材料腐蚀的影响 [J]. 核科学与工程, 2014, 34: 97
18 Betova I, Bojinov M, Karastoyanov V, et al. Effect of water chemistry on the oxide film on alloy 690 during simulated hot functional testing of a pressurised water reactor [J]. Corros. Sci., 2012, 58: 20
19 Wei K J, Wang X P, Zhu M H, et al. Effects of Li, B and H elements on corrosion property of oxide films on ZIRLO alloy in 300oC/14 MPa lithium borate buffer solutions [J]. Corros. Sci., 2021, 181: 109216
20 Molander A, Norring K, Andersson P O, et al. Environmental effects on PWSCC initiation and propagation in alloy 600 [A]. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. New York: Springer, 2011: 1699
21 Vankeerberghen M, Weyns G, Gavrilov S, et al. Crack propagation rate modelling for 316SS exposed to PWR-relevant conditions [J]. J. Nucl. Mater., 2009, 384: 274
22 Liu W Q, Zhou B X, Li Q, et al. Detrimental role of LiOH on the oxide film formed on zircaloy-4 [J]. Corros. Sci., 2005, 47: 1855
23 Billot P, Yagnik S, Ramasubramanian N, et al. The role of lithium and boron on the corrosion of zircaloy-4 under demanding PWR-type conditions [A]. Zirconium in the Nuclear Industry: 13th International Symposium [C]. West Conshohocken: American Society for Testing and Materials, 2002: 169
24 Zhao Y F, Tang M, Jiang E, et al. Inhibition effects of low concentration of boron on corrosion of zirconium alloy [J]. Nucl. Power Eng., 2019, 40(2): 32
24 赵永福, 唐敏, 姜峨 等. 低浓度硼对锆合金缓蚀作用研究 [J]. 核动力工程, 2019, 40(2): 32
25 Li J. The focused-ion-beam microscope—More than a precision ion milling machine [J]. JOM, 2006, 58(3): 27
26 Cao X Y, Zhu P, Wang W, et al. Effect of thermal aging on oxide film of stainless steel weld overlay cladding exposed to high temperature water [J]. Mater. Charact., 2018, 138: 195
27 Hanbury R D, Was G S. Oxide growth and dissolution on 316L stainless steel during irradiation in high temperature water [J]. Corros. Sci., 2019, 157: 305
28 Kuang W J, Han E H, Wu X Q, et al. Microstructural characteristics of the oxide scale formed on 304 stainless steel in oxygenated high temperature water [J]. Corros. Sci., 2010, 52: 3654
29 Terachi T, Yamada T, Miyamoto T, et al. Corrosion behavior of stainless steels in simulated PWR primary water—Effect of chromium content in alloys and dissolved hydrogen [J]. J. Nucl. Sci. Technol., 2008, 45: 975
30 Lister D H, Davidson R D, Mcalpine E. The mechanism and kinetics of corrosion product release from stainless steel in lithiated high temperature water [J]. Corros. Sci., 1987, 27: 113
31 Macdonald D D, Urquidi-Macdonald M. Theory of steady-state passive films [J]. J. Electrochem. Soc., 1990, 137: 2395
32 Robertson J. The mechanism of high temperature aqueous corrosion of stainless steels [J]. Corros. Sci., 1991, 32: 443
33 Matthews R P, Knusten R D, Westraadt J E, et al. Intergranular oxidation of 316L stainless steel in the PWR primary water environment [J]. Corros. Sci., 2017, 125: 175
34 Macdonald D D. Passivity—The key to our metals-based civilization [J]. Pure Appl. Chem., 1999, 71: 951
35 Wu W S, Ran G, Li Y P, et al. Early corrosion behaviour of irradiated FeCrAl alloy in a simulated pressurized water reactor environment [J]. Corros. Sci., 2020, 174: 108824
36 Shen Z, Tweddle D, Yu H B, et al. Microstructural understanding of the oxidation of an austenitic stainless steel in high-temperature steam through advanced characterization [J]. Acta Mater., 2020, 194: 321
37 Robino C V. Representation of mixed reactive gases on free energy (Ellingham-Richardson) diagrams [J]. Metall. Mater. Trans., 1996, 27B: 65
38 Sun H, Wu X Q, Han E H, et al. Effects of pH and dissolved oxygen on electrochemical behavior and oxide films of 304SS in borated and lithiated high temperature water [J]. Corros. Sci., 2012, 59: 334
39 Shu M, Wang C L, Chen Y. Studies on electrochemical corrosion behaviors and 316NG stainless steel in boron-lithium solutions [J]. Nucl. Power Eng., 2018, 39(5): 63
39 舒茗, 王丛林, 陈勇. 316NG不锈钢在硼-锂溶液中的电化学腐蚀行为研究 [J]. 核动力工程, 2018, 39(5): 63
40 Kaczorowski D, Combrade P, Vernot J P, et al. Water chemistry effect on the wear of stainless steel in nuclear power plant [J]. Tribol. Int., 2006, 39: 1503
41 Park Y J, Choi K C, Ha Y K. Solubility study of nickel ferrite in boric acid using a flow-through autoclave system under high temperature and high pressure [J]. Nucl. Eng. Technol., 2016, 48: 554
42 Tremaine P R, Leblanc J C. The solubility of magnetite and the hydrolysis and oxidation of Fe2+ in water to 300oC [J]. J. Solution Chem., 1980, 9: 415
43 Cox B, Ungurelu M, Wong Y M, et al. Mechanisms of LiOH degradation and H3BO3 repair of ZrO2 films [A]. Zirconium in the Nuclear Industry: 11th International Symposium [C]. West Conshohocken: American Society for Testing and Materials, 1996: 114
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