The oxidation kinetics of the classical pressurized water reactors (PWR)
cladding alloy Zircaloy-4 have been extensively investigated over a wide temper-
ature 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 (both AREVA), ZIRLOTM (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 light-
microscopy 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 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.