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SCC growth behavior of austenitic stainless steels in PWR primary water conditions



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M. Le Calvar, G. Turluer (retraité), C. Guerre, O. Raquet, E. Herms,
12th International Conference on Environmental Degradation of Materials in Nuclear Power System - Water Reactors, Salt Lake City, Utah, USA, 14-18 août 2005,
Rapport DSR 112

Type de document > *Rapport/contribution à GT (papier ou CD-Rom), *Congrès/colloque

Mots clés > corrosion, réacteurs à eau sous pression (REP)

Unité de recherche > IRSN/DSR/SAMS

Auteurs > LE CALVAR Marc

Date de publication > 30/03/2006


Fatigue air pre-cracked Compact Tensile (CT) specimens in cold-worked austenitic stainless steel were tested in the VENUS corrosion experiment, a high temperature recirculating loop, which simulates the primary water of Pressurized Water Reactors (PWR). In order to assess the effect of loading conditions and temperature on the susceptibility of stainless steels to Stress Corrosion Cracking (SCC), CT specimen were tested at constant loads or low frequency cyclic loading at 325 °C and 289° C. The material studied was highly cold-worked with a generalized deformation percentage of 60%. Fracture surfaces were characterised by macroscopic and microscopic obervations. PWSCC crack growth rates were evaluated by post-mortem observations. At both temperatures, cracks propagate with cyclic loading conditions as well as at constant load at K values above 25 MPa.m1/2. Comparison between average crack growth rates obtained at constant load and cyclic loading showed that no systematic accelerating effect was observed for the cyclic conditions tested.