设为首页 
您现在的位置: 中国科学院核用材料与安全评价重点实验室 > 通知公告 > 正文
学术报告:The SCC initiation behavior of alloy 690 in simulated pressurized water reactor primary environment
作者:admin    通知公告来源:本站原创    点击数:154    更新时间:2018/5/30           

    报告:The SCC initiation behavior of alloy 690 in simulated pressurized water reactor primary environment

  报告人:匡文军

  2015年9月至2018年5月在密歇根大学核工程系任职助理研究员;2017年入选第十三批中组部青年千人计划,现为西安交通大学教授,青年拔尖人才计划A类;目前主要研究方向是核电结构材料的环境失效。

  时间:5月31日(周四)上午9:30-11:00

  地点:南区报告厅一楼

  内容简介:Alloy 690 is an important structure material in modern light water reactor. Initiation of stress corrosion cracks in Alloy 690 in high temperature water is a rare occurrence and depends on the method by which the sample is loaded. Only in dynamic straining experiments is crack initiation consistently observed. The investigation on the oxidation behavior of alloy 690 shows that this material is prone to penetrative internal oxidation along closely packed lattice planes.  However, the grain boundary forms a protective Chromia layer on the surface with the fast diffusion of Cr along the grain boundary. The surface oxide film would be breached under dynamic straining and the chromium depleted grain boundary is open to the inwards diffusing oxygen, resulting in formation and rupture of oxides down the grain boundary. Microstructure study of the SCC cracks indicates that grain boundary migration and intergranular oxidation are two fundamental precursors in SCC propagation. Cold work can significantly change the grain boundary structure and extends the SCC propagation path.

 

 

通知公告录入:admin    责任编辑:admin 
  • 上一个通知公告:
  • 下一个通知公告:
  •  


    地址:辽宁省沈阳市沈河区文萃路62号 邮编:110014 电话:024-23915772 电子邮件:labnmsa@imr.ac.cn
    Copyright © 中国科学院核用材料与安全评价重点实验室