SYMPOSIA PAPER Published: 01 January 2002
STP11405S

The Influence of In-Situ Clad Straining on the Corrosion of Zircaloy in a PWR Water Environment

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Zircaloy fuel element cladding changes dimensions during service in a pressurized water reactor (PWR) as a result of stress-free irradiation-induced growth, creep-driven by fuel pellet expansion and hydriding. The application of a tensile load in a high-temperature autoclave environments has previously been reported to increase the corrosion rate of Zircaloy, and heat treatments (beta quenching) that reduce the irradiation-induced stress-free growth of Zircaloy have previously been reported to reduce Zircaloy corrosion in-reactor. However, the effect of in-situ straining on Zircaloy corrosion in a PWR environment has not been systematically studied and reported in the literature. This paper presents experimental results regarding the effect of in-situ straining on Zircaloy corrosion in a PWR environment, both in-reactor and in an autoclave. In-situ electrochemical data and post-test metallographic data are presented. In-situ straining is seen to increase the corrosion rate of the Zircaloy, likely by breaking the passivating layer that is forming on the surface. However, the effect is a function of the applied strain rate.

Author Information

Kammenzind, BF
Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA
Eklund, KL
Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA
Bajaj, R
Bettis Atomic Power Laboratory, Bechtel Corporation, West Mifflin, PA
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Developed by Committee: B10
Pages: 524–560
DOI: 10.1520/STP11405S
ISBN-EB: 978-0-8031-5468-1
ISBN-13: 978-0-8031-2895-8