Oxidation kinetics and related phenomena of zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions

Abstract

With respect to the behavior of Nuclear Pressurized Water Reactor fuel cladding during accidents the oxidation kinetics of Zircaloy-4 tubing in steam and hydrogen-steam mixtures and the related changes in the mechanical properties have been investigated. Short tube sections were exposed to steam in surplus between 600 and 1600°C under isothermal and temperature transient conditions and to steam of limited supply and hydrogen-steam mixtures between 800 and 1300°C under isothermal conditions. Tube capsules were creep-rupture tested in steam under isothermal/isobaric (600–1300°C, 7–150 bar) and various temperature/pressure transient conditions. The oxidation kinetics of Zircaloy-4 is described in case of short-term 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 2-phase transformations. The protective character of preexistent scales is lost above 1100°C for the same reason. The limits of steam supply rate dependent oxidation and its interrelation with the competing hydrogen take-up are described. 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.