The oxidation behavior of Zircaloy‐4 in steam between 600 and 1600°C

DOI Web Site Web Site 被引用文献2件

抄録

<jats:title>Abstract</jats:title><jats:p>With respect to the behavior of Nuclear Pressurized Water Reactor fuel cladding during accidents the oxidation kinetics of Zircaloy‐4 tubing in steam and the related changes in the mechanical properties have been investigated. Short tube sections were exposed to steam between 600 and 1600°C under isothermal and temperature transient conditions. Tube capsules were creep‐rupture tested in steam under isothermal/isobaric (600–1300°C, 7–150 bar) and various temperature/pressure transient conditions.</jats:p><jats:p>The oxidation kinetics of Zircaloy‐4 is described in case of shortterm reaction by simple rate laws. The long‐term behavior is proved to be highly influenced by the breakaway transition and finally by total consumption of the tube wall. On the basis of the isothermal behavior the short‐term temperature transient case can be understood and also modelled. Deviations from predictable behavior are the consequences of Zr‐ and ZrO<jats:sub>2</jats:sub>‐phase transformations. The protective character of preexistent scales is lost above 1100°C for the same reason.</jats:p><jats:p>The consequences of oxidation on the mechanical properties are increased strength and decreased ductility, whereas the response to creep deformation is an overall increase in oxidation due to oxide cracking.</jats:p>

収録刊行物

被引用文献 (2)*注記

もっと見る

キーワード

詳細情報 詳細情報について

問題の指摘

ページトップへ