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The oxidation kinetics of the classical pressurized water reactors (PWR) cladding alloy Zircaloy-4 have been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in Western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as the reference material, Duplex DX-D4, M5 Ò (both AREVA), ZIRLO TM (Westinghouse), and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses were performed by lightmicroscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found at up to 1000°C, where the breakaway effect plays a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200°C. Generally, the M5 Ò alloy revealed the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys was considerably dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen.

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