Insights into the superior stress corrosion cracking resistance of FeCrAl alloy in high temperature hydrogenated water: The critical role of grain boundary oxidation
Source abstract
Abstract
The stress corrosion cracking (SCC) initiation behavior of FeCrAl alloy was studied through constant extension rate tensile (CERT) test in high temperature hydrogenated water. Compared with a nuclear-grade austenitic 304 stainless steel (SS), the FeCrAl alloy is much more resistant to SCC initiation. The difference in SCC resistance results from the different oxidation behaviors. 304 SS shows typical preferential intergranular oxidation (PIO) which can degrade the strength of grain boundary. In contrast, FeCrAl is immune to PIO and forms homogeneous nano-grained oxide layer. The original grain boundary vanishes after oxidation, thus is not a weak site for SCC initiation.
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Research context
OpenAlex topics: Nuclear Materials and Properties · Hydrogen embrittlement and corrosion behaviors in metals · Fusion materials and technologies
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (40)
Show source-provided references
1. Oxidation of Zircaloy by steam10.1016/0022-3115(91)90428-A ↗
2. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding10.1016/j.jnucmat.2013.06.041 ↗
3. Current knowledge on core degradation phenomena, a review10.1016/S0022-3115(98)00899-X ↗
4. High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments10.1016/j.jnucmat.2013.05.047 ↗
5. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions10.1016/j.jnucmat.2013.09.052 ↗
6. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions10.1016/j.anucene.2015.06.032 ↗
7. AREVA NP's enhanced accident-tolerant fuel developments: focus on Cr-coated M5 cladding10.1016/j.net.2017.12.004 ↗
8. P.J. Doyle, S.S. Raiman, R. Rebak, K.A. Terrani, Characterization of the Hydrothermal Corrosion Behavior of Ceramics for Accident Tolerant Fuel Cladding, in: Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, 2019, pp. 1485–1496.10.1007/978-3-030-04639-2_99 ↗
9. Volatility of common protective oxides in high-temperature water vapor: current understanding and unanswered questions10.4028/www.scientific.net/MSF.461-464.765 ↗
10. Preliminary assesment of the performance of SiC based accident tolerant fuel in commercial LWR systems
11. Intermediate temperature corrosion behaviour of Fe-12Cr-6Al-2Mo-0.2Si-0.03Y alloy (C26M) at 300–600 °C10.1016/j.corsci.2019.05.029 ↗
12. Uniform corrosion behavior of FeCrAl alloys in borated and lithiated high temperature water10.1016/j.jmst.2020.07.026 ↗
13. Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants10.1051/epjn/2017029 ↗
14. A study of the oxidation of FeCrAl alloy in pressurized water and high-temperature steam environment10.1016/j.corsci.2015.02.027 ↗
15. Uniform corrosion of FeCrAl alloys in LWR coolant environments10.1016/j.jnucmat.2016.06.047 ↗
16. Microstructural characteristics of the oxide scale formed on 304 stainless steel in oxygenated high temperature water10.1016/j.corsci.2010.07.015 ↗
17. The oxidation behaviour of 304 stainless steel in oxygenated high temperature water10.1016/j.corsci.2010.09.001 ↗
18. Behavior of stainless steels in pressurized water reactor primary circuits10.1016/j.jnucmat.2012.03.034 ↗
19. The mechanism of oxide film formation on austenitic stainless steels in high temperature10.1016/S0010-938X(97)00140-6 ↗
20. Stress corrosion cracking of current structural materials in commercial nuclear power plants10.5006/0801 ↗
21. General corrosion and stress corrosion cracking of Alloy 600 in light water reactor primary coolants10.1016/j.jnucmat.2019.04.023 ↗
22. An ATEM study of oxidation behavior of SCC crack tips in 304L stainless steel in high temperature oxygenated water10.1016/j.jnucmat.2005.07.006 ↗
23. Multi-scale characterization of stress corrosion cracking of cold-worked stainless steels and the influence of Cr content10.1016/j.actamat.2009.07.040 ↗
24. Crack initiation of neutron-irradiated 304L stainless steel in PWR primary water10.1016/j.corsci.2021.109902 ↗
25. A high-resolution characterization of irradiation-assisted stress corrosion cracking of proton-irradiated 316L stainless steel in simulated pressurized water reactor primary water10.1016/j.corsci.2022.110187 ↗
26. How oxidized grain boundaries fail10.1016/j.actamat.2013.05.012 ↗
27. Micro-mechanical measurement of fracture behaviour of individual grain boundaries in Ni alloy 600 exposed to a pressurized water reactor environment10.1016/j.corsci.2015.10.019 ↗
28. K. Fujii, T. Miura, H. Nishioka, K. Fukuya, Degradation of grain boundary strength by oxidation in alloy 600, 15th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems-water reactors 2011, pp 1447–1461.10.1007/978-3-319-48760-1_89 ↗
29. Resistance of ferritic FeCrAl alloys to stress corrosion cracking for light water reactor fuel cladding applications10.5006/3632 ↗
30. SCC resistance of irradiated and unirradiated High Cr ferritic steels
31. The mechanism of oxide film formation on Alloy 690 in oxygenated high temperature water10.1016/j.corsci.2011.07.038 ↗
32. S. Zhang, S..Wang, X..Feng, J. Hou, W. Kuang, On the enhanced stress corrosion cracking resistance of a selective laser melted 304L stainless steel in high-temperature hydrogenated water, to be sumitted.
33. TEM investigations of the oxide layers formed on a 316L alloy in simulated PWR environment10.1007/s10853-012-6975-0 ↗
34. The effect of grain boundary structure on the intergranular degradation behavior of solution annealed alloy 690 in high temperature, hydrogenated water10.1016/j.actamat.2019.10.041 ↗
35. Correlation between grain boundary migration and stress corrosion cracking of alloy 600 in hydrogenated steam10.1016/j.actamat.2020.01.020 ↗
36. The intergranular oxidation behavior of low-angle grain boundary of alloy 600 in simulated pressurized water reactor primary water10.1016/j.actamat.2021.117533 ↗
37. Nanoscale characterisation of grain boundary oxidation in cold-worked stainless steels10.1016/j.corsci.2012.06.030 ↗
38. A detailed TEM and SEM study of Ni-base alloys oxide scales formed in primary conditions of pressurized water reactor10.1016/j.jnucmat.2010.05.010 ↗
39. Effect of machining on stress corrosion crack initiation in warm-forged type 304L stainless steel in high temperature water10.1016/j.actamat.2018.11.046 ↗
40. Effect of machining on oxide development in type 316L stainless steel in high-temperature hydrogenated water10.1016/j.corsci.2021.109444 ↗