Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding
Source abstract
Abstract
Abstract As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at Oak Ridge National Laboratory. A second generation of these alloys bearing additional alloying elements over the previously tested model alloys were tested in boiling water reactor (BWR) conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe–10Cr–6Al–2Mo, Fe–13Cr–5Al–2Mo, Fe–13Cr–6Al–2Mo, Fe–13Cr–7Al–2Mo, and Fe–13Cr–5Al–2Mo–1Nb were tested in normal water chemistry (NWC) and hydrogen water chemistry (HWC) for 9 months in continuously refreshing autoclaves. Commercial FeCrAl alloy APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Fe–13Cr–7Al–2Mo performed worst with an average mass loss of 1.3 mg/cm2 over 9 months. This mass loss represents an estimated thickness loss of approximately 60 μm over 6 years. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm2 or mass gains up to 0.05 mg/cm2. Based on the results of this testing, the 2nd generation FeCrAl alloys tested exhibit low wall thickness loss and are suitable in terms of corrosion resistance for use as LWR cladding in BWR-HWC and BWR-NWC under normal operating conditions.
Source: openaire · Quality: candidate unverified · Source record ↗
Research context
OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Fusion materials and technologies
SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.
References (38)
Show source-provided references
1. Material selection for accident tolerant fuel cladding
2. Accident tolerant fuel cladding development: promise, status, and challenges10.1016/j.jnucmat.2017.12.043 ↗
3. High temperature properties of a new powder metallurgical FeCrAl alloy10.4028/www.scientific.net/MSF.461-464.455 ↗
4. Oxidation and creep limited lifetime of kanthal APMT®, a dispersion strengthened FeCrAlMo alloy designed for strength and oxidation resistance at high temperatures10.1007/s11085-012-9324-4 ↗
5. R.B. Rebak, Iron-chrome-aluminum Alloy Cladding for Increasing Safety in Nuclear Power Plants, (n.d.). doi:10.1051/epjn/2017029.10.1051/epjn/2017029 ↗
6. Versatile oxide films protect FeCrAl alloys under normal operation and accident conditions in light water power reactors10.1007/s11837-017-2705-z ↗
7. High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments10.1016/j.jnucmat.2013.05.047 ↗
8. Effect of steam on high temperature oxidation behaviour of alumina-forming alloys10.1179/0960340914Z.00000000058 ↗
9. Performance of FeCrAl for accident-tolerant fuel cladding in high-temperature steam10.1515/corrrev-2016-0067 ↗
10. Radiation tolerance of neutron-irradiated model Fe–Cr–Al alloys10.1016/j.jnucmat.2015.06.023 ↗
11. Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys10.1016/j.jnucmat.2017.03.038 ↗
12. Irradiation-enhanced α′ precipitation in model FeCrAl alloys10.1016/j.scriptamat.2016.02.002 ↗
13. A combined APT and SANS investigation of α′ phase precipitation in neutron-irradiated model FeCrAl alloys10.1016/j.actamat.2017.02.077 ↗
14. Yamamoto · 2015
15. Uniform corrosion of FeCrAl alloys in LWR coolant environments10.1016/j.jnucmat.2016.06.047 ↗
16. Robertson · Corrosion Sci. · 1991 · 32 · 44310.1016/0010-938X(91)90125-9 ↗
17. The mechanism of oxide film formation on austenitic stainless steels in high temperature water10.1016/S0010-938X(97)00140-6 ↗
18. Pourbaix diagrams for the iron–water system extended to high-subcritical and low-supercritical conditions10.1016/j.corsci.2011.10.034 ↗
19. Pourbaix diagrams for the ternary system of iron-chromium-nickel10.5006/1.3283945 ↗
20. Influence of dissolved oxygen concentration on the oxide film formed on 304 stainless steel in high temperature water10.1016/j.corsci.2012.06.007 ↗
21. Rumble · 2017
22. Yamamoto · 2016
23. Smerd · 2011 · 430
24. Stehle · 1984
25. The effects of radiation and oxygen on the aqueous oxidation of Zirconium and its alloys at 290°C10.1016/0022-3115(73)90122-0 ↗
26. Corrosion performance of zircaloy-2 and zircaloy-4 PWR fuel cladding
27. Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments10.1016/j.jeurceramsoc.2016.11.033 ↗
28. Characterization of the hydrothermal corrosion behavior of ceramics for accident tolerant fuel cladding
29. Radiolysis driven changes to oxide stability during irradiation-corrosion of 316L stainless steel in high temperature water10.1016/j.jnucmat.2017.05.042 ↗
30. Accelerated corrosion and oxide dissolution in 316L stainless steel irradiated in situ in high temperature water10.1016/j.jnucmat.2017.05.043 ↗
31. Raiman · 2017
32. Oxide growth and dissolution on 316L stainless steel during irradiation in high temperature water10.1016/j.corsci.2019.06.006 ↗
33. Field · 2018
34. Immersion testing of FeCrAl tubes under simulated light water nuclear reactor normal operation conditions
35. Microstructural study of oxide layers formed on Zircaloy-4 in autoclave and in reactor Part i: impact of irradiation on the microstructure of the zirconia layer10.1016/0022-3115(95)00070-4 ↗
36. A model for waterside oxidation of Zircaloy fuel cladding in pressurized water reactors10.1016/0022-3115(91)90487-R ↗
37. Development of a mechanistic model to assess the external corrosion of the Zircaloy claddings in PWRs-
38. Corrosion of Zirconium alloy components in light water reactors