Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions
Source abstract
Abstract
The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Subcritical and Supercritical Water Processes · Advanced Materials Characterization Techniques
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References (12)
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1. D. Bittermann, D. Squarer, T. Schulenberg, Y. Oka, Economic Prospects of the HPLWR, GENES4/ANP2003, Kyoto, Japan, September 15–19, 2003, Paper 1003.
2. T. Schulenberg, J. Starflinger, J. Heinecke, Three-pass core design proposal for a high performance light water reactor, in: Proceedings of 2nd COE-INES-2 International Conference on Innovative Nuclear Energy Systems, Yokohama, Japan, November 26–30, 2006.
3. A. Wank, T. Schulenberg, A. Class, Coolant mixing in the plenum of a HPLWR three pass core, in: Proceedings of ICAPP ’08 Anaheim, California, USA, June 8–12, 2008, Paper 8100.
4. Was · J. Nucl. Mater. · 2007 · 371 · 17610.1016/j.jnucmat.2007.05.017 ↗
5. C. Koehly, T. Schulenberg, J. Starflinger, HPLWR reactor design concept, in: 4th International Symposium on Supercritical Water-Cooled Reactors, Heidelberg, Germany, March 8–11, 2009, Paper 37.
6. Siegl · 2009
7. S. Penttilä, A. Toivonen, L. Heikinheimo, R. Novotny, Corrosion Studies of Candidate Materials for European HPLWR, ICAPP ‘08, Anaheim, California, USA, June 8–12, 2008.
8. Y. Tsuchiya, F. Kano, N. Saito, A. Shioiri, S. Kasahara, K. Moriya, H. Takahashi, GENES4/ANP2003, Kyoto, Japan, September 2003, Paper 1096.10.1299/jsmezairiki.2003.0_741 ↗
9. R. Fujisawa, K. Nishimura, T. Kishida, M. Sakaihara, Y. Kurata, Y. Watanabe, Corrosion 2005, NACE International, Houston, Texas, USA, 2005, Paper 05395.
10. G.S. Was, S. Teysseyre, Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, The Minerals, Materials and Metals Society, Warrendale, PA, USA, 2005, Paper 1343.
11. G.S. Was, S. Teysseyre, J. McKinley, Corrosion 2004, NACE International, TX, Paper 04492, Houston, Texas, USA, 2004, Paper 04492.
12. M. Tsubota, Y. Katayama, Y. Kanazawa, Relation between the mechanical properties and scc behavior of the alloys used in high temperature water, in: 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems, Whistler, British Columbia, Canada, August 19–23, 2007.