Article ID Journal Published Year Pages File Type
1565728 Journal of Nuclear Materials 2013 8 Pages PDF
Abstract
The viability of advanced oxidation-resistant Fe-base alloys to protect zirconium from rapid oxidation in high-temperature steam environments has been examined. Specimens were produced such that outer layers of FeCrAl ferritic alloy and Type 310 austenitic stainless steel were incorporated on the surface of zirconium metal slugs. The specimens were exposed to high-temperature 0.34 MPa steam at 1200 and 1300 °C. The primary degradation mechanism for the protective layer was interdiffusion with the zirconium, as opposed to high-temperature oxidation in steam. The FeCrAl layer experienced less degradation and protected the zirconium at 1300 °C for 8 h. Constituents of the Fe-base alloys rapidly diffused into the zirconium and resulted in the formation of various intermetallic layers at the interface and precipitates inside the bulk zirconium. The nature of this interaction for FeCrAl and 310SS has been characterized by use of microscopic techniques as well as computational thermodynamics. Finally, a reactor physics discussion on the applicability of these protective layers in light-water-reactor nuclear fuel structures is offered.
Related Topics
Physical Sciences and Engineering Energy Nuclear Energy and Engineering
Authors
, , , ,