SYMPOSIA PAPER Published: 01 January 1990

Fatigue Crack Growth Behavior of Different Stainless Steels in Pressurized Water Reactor Environments


An experimental program has been conducted in order to determine the fatigue crack growth rate (FCGR) curves of different stainless steels for nuclear pressure vessels and pipings in pressurized water reactor environments and to define reference fatigue crack growth rate curves for these materials. The parameters studied are: the steel structure, the load ratio (R), and the water chemistry. A slight, but clear effect of environment was observed for the five steels studied. The paper presents the FCGR curves and the fractographic examinations for the different testing conditions and stresses the differences between FCGR behavior of stainless steels and low-alloy steels.

Author Information

Amzallag, C
UNIREC, Centre de Recherches, Firminy, France
Maillard, J-L
E.C.A.N. INDRET, La Montagne, France
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Developed by Committee: G01
Pages: 463–494
DOI: 10.1520/STP24081S
ISBN-EB: 978-0-8031-5114-7
ISBN-13: 978-0-8031-1276-6