期刊文献+

奥氏体321不锈钢在550℃静态铅铋共晶合金中的腐蚀行为 被引量:4

Corrosive Behavior of Austenitic 321 Stainless Steel in Static Lead-bismuth Eutectic Alloy at 550 ℃
下载PDF
导出
摘要 奥氏体321不锈钢常用作核反应堆冷却剂主管道结构材料,铅铋共晶合金是第四代核能系统(GenⅣ)铅冷快堆冷却剂的主要候选材料。为研究321不锈钢与高温液态铅铋共晶合金的相容性,对321不锈钢在550℃液态铅铋共晶合金中的200、400、600 h腐蚀现象进行了研究。对不同腐蚀时间后腐蚀试样的表面和截面分别进行了XRD和SEM、EDS检测。结果发现:在321不锈钢试样表面产生了一种随腐蚀时间增加先生长后脱落的含O、Ti、Pb元素的化合物(Ti_2O和Pb_2O_3);在321不锈钢基体与铅铋共晶合金交界处会产生一层随腐蚀时间增加不断增厚的扩散层;321不锈钢在铅铋共晶合金中发生溶解腐蚀,在Fe、Cr元素不断向铅铋共晶合金中溶解时,伴随着Pb、Bi元素向基体中的渗透。 Austenitic 321 stainless steel is always used as the main pipeline material for nuclear reactor coolant. The lead-bismuth eutectic(LBE) alloy is a primary candidate coolant for the lead-cooled fast reactor for the fourth generation nuclear power system(Gen Ⅳ). The compatibility of 321 stainless steel and LBE alloy at 550 ℃ for 200, 400 and 600 h was studied. X-ray diffraction(XRD), scanning electron microscope(SEM) and energy dispersive spectrometry(EDS) were carried out on the surface and cross-section of corrosive samples with different corrosive time. A compound containing O, Ti and Pb elements(Ti2O and Pb2O3) is developed firstly on the surface of 321 stainless steel samples and peeled off with the increase of corrosive time successively. At the junction of 321 stainless steel matrix and LBE alloy, diffusion layer is thickened with time. In the process of corrosion, dissolution corrosion of 321 stainless steel occurs in the LBE alloy. Fe and Cr elements dissolve continuously in LBE alloy, meanwhile, Pb and Bi elements infiltrate into the base metal.
作者 鞠娜 雷玉成 陈钢 朱强 李天庆 王丹 JU Na;LEI Yucheng;CHEN Gang;ZHU Qiang;LI Tianqing;WANG Dan(School of Material Science and Engineering, Jiangsu University, Zhenjiang 212013, China)
出处 《原子能科学技术》 EI CAS CSCD 北大核心 2019年第3期427-433,共7页 Atomic Energy Science and Technology
基金 国家自然科学基金资助项目(51875264) 国家自然科学基金青年基金资助项目(51505197)
关键词 铅冷快堆 铅铋共晶合金 321不锈钢 溶解腐蚀 lead-cooled fast reactor lead-bismuth eutectic alloy 321 stainless steel dissolution corrosion
  • 相关文献

参考文献2

二级参考文献10

  • 1Schulenberg T,Cheng X,Stieglitz R.Thermal-Hy-draulics of Lead Bismuth for Accelerator Driven System. The 11thInternational Topical Meeting on NuclearReactor Thermal-Hydraulics (NURETH-11) . 2005
  • 2I.V. Gorynin,G.P. Karzov,V.G. Markov, et al.Structure materials for power plants with heavy liquid metals as coolants. Proceedings:Heavy Liquid Metal Coolants in Nuclear Technology (HLMC-98) . 1998
  • 3Kapoor S S.Roadmap for development of Accelerator Driven Sub-critical Reactor Systems (ADS):An interim report of the co-ordination committee on ADS. BARC/2001/R/ 004 . 2001
  • 4Gromov B F,Orlov Y I,Martynov P N,et al.Issues of technology of heavy liquid metal coolants (leadbismuth and lead). Proceedings of Heavy Liquid Metal Coolants in Nuclear Technology,HLMC’’98 . 1998
  • 5Yachmenyov G S,Rusanov A E,et al.Problems of structural materials corrosion in lead-bismuth coolant. Proceedings of Heavy Liquid Metal Coolants in Nuclear Technology,HLMC’’98 . 1998
  • 6吴宜灿,柏云清,宋勇,黄群英,刘超,王明煌,周涛,金鸣,吴庆生,汪建业,蒋洁琼,胡丽琴,李春京,高胜,李亚洲,龙鹏程,赵柱民,郁杰,FDS团队.中国铅基研究反应堆概念设计研究[J].核科学与工程,2014,34(2):201-208. 被引量:65
  • 7王艳青,黄群英,武欣,吴斌,张敏,高胜.铅铋合金中Bi/Bi2O3型氧传感器准确性及稳定性测试研究[J].原子能科学技术,2015,49(3):572-576. 被引量:6
  • 8张敏,王艳青,吴斌,武欣,高胜,黄群英.静态铅铋中Pt/Air型氧传感器性能初步研究[J].核科学与工程,2015,35(1):1-7. 被引量:4
  • 9陈建伟,吴庆生,李京,韩骞,黄群英.包壳材料316Ti在液态铅铋中的腐蚀氧化层分析[J].原子能科学技术,2015,49(B05):187-193. 被引量:4
  • 10田书建,张建武.316L和T91不锈钢在550℃静态铅铋合金中的腐蚀行为[J].中国科学技术大学学报,2015,45(9):751-756. 被引量:15

共引文献18

同被引文献23

引证文献4

二级引证文献9

相关作者

内容加载中请稍等...

相关机构

内容加载中请稍等...

相关主题

内容加载中请稍等...

浏览历史

内容加载中请稍等...
;
使用帮助 返回顶部