1. A Review of Development of Zirconium Alloys as a Fuel Cladding Material and its Oxidation Behavior at High-Temperature Steam
- Author
-
Wasim M. k. Helal, John N. Njoroge, Mamoun I. A. Sagiroun, and Xin Rong Cao
- Subjects
Thesaurus (information retrieval) ,Materials science ,Chemical substance ,020209 energy ,Metallurgy ,Zirconium alloy ,technology, industry, and agriculture ,02 engineering and technology ,equipment and supplies ,Cladding (fiber optics) ,01 natural sciences ,010305 fluids & plasmas ,0103 physical sciences ,0202 electrical engineering, electronic engineering, information engineering - Abstract
Currently, Zr-alloys are widely used in nuclear power reactors for fuel cladding and structural components. Many types of zr-based alloys were developed to overcome the challenges encountered in the progress of nuclear reactors (high-burnup and high-duty). Oxygen diffused into the cladding, hydrogen absorbed in the cladding (breakaway oxidation and ruptured balloons) and rapid oxidation rate are results of chemical interaction of cladding material with steam at high temperature. Zirconium alloys seem to be the most suitable for use in fuel cladding, if they can overcome the rapid oxidation at temperature higher than 1200 °C. Previous studies on the oxidation behavior for some Zr-alloys nuclear fuel cladding tubes in steam and steam–air atmospheres at high temperatures are reviewed. The oxidation behavior of zirconium-alloys is strongly affected by the chemical composition of alloys and its surface conditions.
- Published
- 2020