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Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

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Abstract

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® (both AREVA), ZIRLO™ (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.

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Abbreviations

ACM:

Advanced cladding materials

DX-D4:

AREVA duplex cladding

E110:

Russian cladding alloy (Zr1Nb)

FZK:

Forschungszentrum Karlsruhe

ISTC:

International Science and Technology Center

KIT:

Karlsruhe Institute of Technology

km :

Rate constant

LOCA:

Loss of coolant accident

LWR:

Light water reactor

M5® :

AREVA cladding alloy (Zr1Nb)

PWR:

Pressurized Water Reactor

QUENCH:

Research program on reflood of an overheated reactor core at KIT

RIA:

Reactivity initiated accident

S:

Surface area

SARNET:

Severe accident network (EC Program)

STA:

Simultaneous thermal analysis

TG:

Thermo-gravimetry

VVER:

Russian PWR reactor

ZIRLO™:

Westinghouse cladding alloy (Zr1Nb1Sn)

Zry-4:

Zircaloy-4 cladding alloy (Zr1Sn)

∆m:

Mass gain

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Acknowledgments

This work was sponsored by the HGF Program NUKLEAR at the Karlsruhe Institute of Technology and partially done within the framework of the SARNET Mobility Program (contract FI6O-CT-2004-509065). The authors are very grateful to P. Severloh and U. Stegmaier (KIT) for sample preparation and microscopic examinations, as well as to C. Vorpahl for performing a number of TG tests. We also thank Tim Haste (IRSN) for his thorough review of the paper. The Zr1Nb (E110) was provided by Russian institutions in the context of the EU program ISTC 1648.2. M5® and Duplex-D4 rod cladding were delivered by AREVA, ZIRLO™ by Westinghouse.

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Authors and Affiliations

  1. Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021, Karlsruhe, Germany

    Martin Steinbrück & Mirco Große

  2. AEKI, Budapest, Hungary

    Nóra Vér

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  1. Martin Steinbrück
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  2. Nóra Vér
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  3. Mirco Große
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Correspondence to Martin Steinbrück.

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Steinbrück, M., Vér, N. & Große, M. Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C. Oxid Met 76, 215–232 (2011). https://doi.org/10.1007/s11085-011-9249-3

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  • DOI: https://doi.org/10.1007/s11085-011-9249-3

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