Behavior of stainless steels in pressurized water reactor primary circuits
| Autor: | D. Féron, B. Tanguy, E. Herms |
|---|---|
| Rok vydání: | 2012 |
| Předmět: |
Austenite
Nuclear and High Energy Physics Materials science Metallurgy Pressurized water reactor Oxide law.invention Corrosion Stress (mechanics) chemistry.chemical_compound Nuclear Energy and Engineering chemistry law Fracture (geology) General Materials Science Irradiation Stress corrosion cracking |
| Zdroj: | Journal of Nuclear Materials. 427:364-377 |
| ISSN: | 0022-3115 |
| DOI: | 10.1016/j.jnucmat.2012.03.034 |
| Popis: | Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture. |
| Databáze: | OpenAIRE |
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