SYMPOSIA PAPER Published: 01 January 1994

Fatigue Behavior of Irradiated and Unirradiated Zircaloy and Zirconium


As a normal part of reactor operation, Zircaloy components in the core of boiling water reactors (BWRs) are subjected to oscillating loads. It then becomes important to assess the fatigue behavior of core materials. These include Zircaloy-2 used as fuel cladding, zirconium used as liners in barrier fuel, and Zircaloy-4 used for fuel channels.

Fatigue testing was performed on unirradiated and irradiated materials. Fully reversed uniaxial fatigue tests were conducted at constant total strain amplitudes between 0.3 and 1.4% at 616 K in air. Fatigue crack growth testing was conducted on conventional compact tension (CT) specimens at 293 and 561 K. Unirradiated material was tested in air and water, and irradiated material was tested in air.

Crack growth rates, expressed as a function of ΔK (applied stress intensity range), were determined to be insensitive to neutron irradiation, but were increased by a factor of 2 to 4 in water, depending on oxygen content of the water.

Fatigue life is shown to be strongly dependent on the partitioning of plastic and elastic strain during the test. In general, the softer zirconiums have longer fatigue lives than Zircaloy. Irradiation reduces the fatigue life of all materials in the low cycle regime, particularly when applied plastic strain is used as the test variable.

The results obtained support the current design basis for fuel rods and channels and reflect the excellent observed performance of BWR core components.

Author Information

Wisner, SB
GE Nuclear Energy, Fuel Materials Technology, Pleasanton, CA
Reynolds, MB
GE Nuclear Energy, Fuel Materials Technology, Pleasanton, CA
Adamson, RB
GE Nuclear Energy, Fuel Materials Technology, Pleasanton, CA
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Developed by Committee: B10
Pages: 499–520
DOI: 10.1520/STP15206S
ISBN-EB: 978-0-8031-5286-1
ISBN-13: 978-0-8031-2011-2