Compatibility of Fe-Cr-Al and Fe-Cr-Al-Mo oxide dispersion strengthened steels with static liquid sodium at 700 °C

Romedenne, M.; Pillai, R.; Harris, B.; Pint, B. A.

doi:10.1016/j.jnucmat.2022.153919

Title: Compatibility of Fe-Cr-Al and Fe-Cr-Al-Mo oxide dispersion strengthened steels with static liquid sodium at 700 °C

Journal Article · Thu Jul 21 00:00:00 EDT 2022 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2022.153919· OSTI ID:1878717

^[1]; Pillai, R. ^[1]; Harris, B. ^[2];

^[1]

Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
US Nuclear Regulatory Commission (NRC), Washington, DC (USA). Office of Nuclear Reactor Regulation

We report the corrosion behavior and tensile properties of oxide dispersion strengthened (ODS) FeCr, FeCrAl, FeCrAlMo steels and 316H stainless steel were evaluated after 1000 h at 700 °C in static sodium with and without an oxygen getter (Zr foil). Without Zr foil to getter the dissolved oxygen in the liquid sodium, combined oxidation (formation of Cr, Fe-rich or Al-rich oxides) and dissolution of the studied materials was observed. For the ODS FeCr, Cr- and Fe-rich oxide formation and dissolution induced Cr depletion in the underlying alloy and led to the formation of laves phases that were observed to be detrimental to the room temperature tensile properties. For ODS FeCrAlMo and FeCrAl Al-rich oxide formation and dissolution resulted in Al depletion in the underlying alloys but did not impact tensile properties at room temperature. For 316H temperature induced sensitization combined with oxidation decreased room temperature tensile properties. With Zr foil added as oxygen getter, roughened surfaces and Al depletion profiles suggested the dissolution of ODS FeCr and FeCrAl into the liquid sodium while limited dissolution was observed for ODS FeCrAlMo and 316H. For the 316H and ODS FeCrAlMo steels, precipitates were formed at the grain boundaries after exposure. Tensile properties of ODS FeCr, FeCrAl and FeCrAlMo were not affected by the 1000 h corrosion exposure with Zr foil added to the sodium but for 316H temperature-induced grain boundary sensitization negatively affected the tensile properties.

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Research Organization:: Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

Sponsoring Organization:: USDOE; USNRC

Grant/Contract Number:: AC05-00OR22725

OSTI ID:: 1878717

Journal Information:: Journal of Nuclear Materials, Vol. 569, Issue 569; ISSN 0022-3115

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (44)

Sigma-phase formation in high chromium ferritic steels at 650°C Niewolak, L.; Garcia-Fresnillo, L.; Meier, G. H. Journal of Alloys and Compounds, Vol. 638 https://doi.org/10.1016/j.jallcom.2015.03.076	journal	July 2015
Structural changes in materials exposed to liquid sodium Thorley, A. W.; Bardsley, J. A. Journal of the Royal Microscopical Society, Vol. 88, Issue 4 https://doi.org/10.1111/j.1365-2818.1968.tb00625.x	journal	November 1968
The corrosion of molybdenum and tungsten in liquid sodium Barker, Marten G.; Morris, Christopher W. Journal of the Less Common Metals, Vol. 44 https://doi.org/10.1016/0022-5088(76)90128-4	journal	January 1976
Carbon and nitrogen transfer in Fe-9 Cr-Mo ferritic steels exposed to a sodium environment Chopra, O. K.; Natesan, K.; Kassner, T. F. Journal of Nuclear Materials, Vol. 96, Issue 3 https://doi.org/10.1016/0022-3115(81)90571-7	journal	February 1981
Irradiation Behavior of Experimental Mark-II Experimental Breeder Reactor II Driver Fuel Hofman, G. L. Nuclear Technology, Vol. 47, Issue 1 https://doi.org/10.13182/NT80-A32408	journal	January 1980
Compatibility of FBR materials with sodium Furukawa, T.; Kato, S.; Yoshida, E. Journal of Nuclear Materials, Vol. 392, Issue 2 https://doi.org/10.1016/j.jnucmat.2009.03.003	journal	July 2009
Carburization of austenitic and ferritic stainless steels in liquid sodium: Comparison between experimental observations and simulations Romedenne, M.; Rouillard, F.; Hamon, D. Corrosion Science, Vol. 159 https://doi.org/10.1016/j.corsci.2019.108147	journal	October 2019
Potential suitability of ferritic and austenitic steels as interconnect materials for solid oxide fuel cells operating at 600°C Niewolak, L.; Wessel, E.; Singheiser, L. Journal of Power Sources, Vol. 195, Issue 22 https://doi.org/10.1016/j.jpowsour.2010.06.007	journal	November 2010
Corrosion of annealed AISI 316 stainless steel in sodium environment Ganesan, Vaidehi; Ganesan, Vedaraman Journal of Nuclear Materials, Vol. 256, Issue 1 https://doi.org/10.1016/S0022-3115(98)00041-5	journal	July 1998
Materials challenges in nuclear energy Zinkle, S. J.; Was, G. S. Acta Materialia, Vol. 61, Issue 3 https://doi.org/10.1016/j.actamat.2012.11.004	journal	February 2013
In-reactor creep rupture properties of 20% CW modified 316 stainless steel Ukai, S.; Mizuta, S.; Kaito, T. Journal of Nuclear Materials, Vol. 278, Issue 2-3 https://doi.org/10.1016/S0022-3115(99)00233-0	journal	April 2000
Single Crystal and Sintered Alumina Corrosion in Liquid Sodium Courouau, Jean-Louis; Steil, Marlu-César; Fouletier, Jacques Oxidation of Metals, Vol. 87, Issue 5-6 https://doi.org/10.1007/s11085-017-9743-3	journal	February 2017
ODS cladding fuel pins irradiation tests using the BOR-60 reactor Kaito, Takeji; Yano, Yasuhide; Ohtsuka, Satoshi Journal of Nuclear Science and Technology, Vol. 50, Issue 4 https://doi.org/10.1080/00223131.2013.773165	journal	April 2013
Microstructural evaluation of a Fe-12Cr nanostructured ferritic alloy designed for impurity sequestration Massey, Caleb P.; Hoelzer, David T.; Seibert, Rachel L. Journal of Nuclear Materials, Vol. 522 https://doi.org/10.1016/j.jnucmat.2019.05.019	journal	August 2019
Modern nanostructured ferritic alloys: A compelling and viable choice for sodium fast reactor fuel cladding applications Hoelzer, D. T.; Massey, C. P.; Zinkle, S. J. Journal of Nuclear Materials, Vol. 529 https://doi.org/10.1016/j.jnucmat.2019.151928	journal	February 2020
Effect of temperature and dissolved oxygen content on the dissolved iron concentration in liquid sodium at equilibrium Meddeb, Sami; Courouau, Jean-Louis; Rouhard, Marion Journal of Nuclear Materials, Vol. 566 https://doi.org/10.1016/j.jnucmat.2022.153785	journal	August 2022
Compatibility of FeCrAlMo with flowing PbLi at 500°-650 °C Jun, Jiheon; Unocic, Kinga A.; Lance, Michael J. Journal of Nuclear Materials, Vol. 528 https://doi.org/10.1016/j.jnucmat.2019.151847	journal	January 2020
THe corrosion of chromium, iron, and stainless steel in liquid sodium Barker, M. G.; Wood, D. J. Journal of the Less Common Metals, Vol. 35, Issue 2 https://doi.org/10.1016/0022-5088(74)90243-4	journal	April 1974
Thermodynamic stability of Na4FeO3 and threshold oxygen levels in sodium for the formation of this compound on AISI 316 steel surfaces Bhat, N. P.; Borgstedt, H. U. Journal of Nuclear Materials, Vol. 158 https://doi.org/10.1016/0022-3115(88)90147-X	journal	August 1988
Thermodynamic and kinetic aspects of carbon transport in sodium-steel systems Natesan, K.; Kassner, T. F. Journal of Nuclear Materials, Vol. 37, Issue 2 https://doi.org/10.1016/0022-3115(70)90087-5	journal	November 1970
Contribution a l'etude de la corrosion de l'acier austenitique bas carbone (afnor Z3 CN 18-10) par le sodium liquide a 700°C Baque, P.; Lapon, A.; Sermet, E. Journal of Nuclear Materials, Vol. 54, Issue 2 https://doi.org/10.1016/0022-3115(74)90135-4	journal	December 1974
The downstream effect in the corrosion of metallic materials in a type 316 stainless steel sodium loop system Suzuki, Tadashi; Mutoh, Isao Journal of Nuclear Materials, Vol. 140, Issue 1 https://doi.org/10.1016/0022-3115(86)90196-0	journal	August 1986
Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors Natesan, K.; Li, Meimei; Chopra, O. K. Journal of Nuclear Materials, Vol. 392, Issue 2 https://doi.org/10.1016/j.jnucmat.2009.03.039	journal	July 2009
Mass transfer behavior of a modified austenitic stainless steel in lithium Tortorelli, P. F.; DeVan, J. H. Journal of Nuclear Materials, Vol. 123, Issue 1-3 https://doi.org/10.1016/0022-3115(84)90251-4	journal	May 1984
Metallic fast reactor fuels Hofman, G. L.; Walters, L. C.; Bauer, T. H. Progress in Nuclear Energy, Vol. 31, Issue 1-2 https://doi.org/10.1016/0149-1970(96)00005-4	journal	January 1997
Further insights into the mechanisms involved in the corrosion of 316L(N) austenitic steel in oxygenated liquid sodium at 550 °C Rivollier, Matthieu; Courouau, Jean-Louis; Tabarant, Michel Corrosion Science, Vol. 165 https://doi.org/10.1016/j.corsci.2019.108399	journal	April 2020
Sodium compatibility of ODS steel at elevated temperature Yoshida, Eiichi; Kato, Shoichi Journal of Nuclear Materials, Vol. 329-333 https://doi.org/10.1016/j.jnucmat.2004.04.312	journal	August 2004
A Comparison of the Oxidation of Fe&?Cr, Ni?Cr and Co?Cr alloys in oxygen and water vapour Wood, G. C.; Wright, I. G.; Hodgkiess, T. Materials and Corrosion/Werkstoffe und Korrosion, Vol. 21, Issue 11 https://doi.org/10.1002/maco.19700211105	journal	November 1970
Carburization of Austenitic and Ferritic Steels in Carbon-Saturated Sodium: Preliminary Results on the Diffusion Coefficient of Carbon at 873 K Romedenne, Marie; Rouillard, Fabien; Duprey, Brigitte Oxidation of Metals, Vol. 87, Issue 5-6 https://doi.org/10.1007/s11085-017-9733-5	journal	February 2017
Microstructural changes in austenitic stainless steels during long-term aging Minami, Y.; Kimura, H.; Ihara, Y. Materials Science and Technology, Vol. 2, Issue 8 https://doi.org/10.1179/mst.1986.2.8.795	journal	August 1986
Oxidation of 316L(N) stainless steel in liquid sodium at 650 °C Rivollier, Matthieu; Courouau, Jean-Louis; Tabarant, Michel Journal of Nuclear Materials, Vol. 500 https://doi.org/10.1016/j.jnucmat.2017.12.037	journal	March 2018
Corrosion of various ferritic steels in an isothermal sodium loop system Rademakers, P. L. F.; Kolster, B. H. Journal of Nuclear Materials, Vol. 97, Issue 3 https://doi.org/10.1016/0022-3115(81)90479-7	journal	April 1981
1032. Reactions of liquid sodium with transition-metal oxides. Part II. Reduction of chromium trioxide to a ternary oxide of chromium(IV) Addison, C. C.; Barker, M. G. Journal of the Chemical Society (Resumed) https://doi.org/10.1039/jr9650005534	journal	January 1965
Compatibility of FeCrAlMo in Flowing Pb-Li at 600°C to 700°C Pint, B. A.; Jun, J.; Cakmak, E. Fusion Science and Technology https://doi.org/10.1080/15361055.2021.1898305	journal	August 2021
The Kinetics of Decarburization of Cr –1% Mo Steel in Sodium Krankota, J. L.; Armijo, J. S. Nuclear Technology, Vol. 24, Issue 2 https://doi.org/10.13182/NT74-A31477	journal	November 1974
Microstructure, corrosion and mechanical properties characterization of AISI type 316L(N) stainless steel and modified 9Cr-1Mo steel after 40,000 h of dynamic sodium exposure at 525 °C Bharasi, N. Sivai; Pujar, M. G.; Das, C. R. Journal of Nuclear Materials, Vol. 516 https://doi.org/10.1016/j.jnucmat.2019.01.012	journal	April 2019
The corrosion of vanadium and vanadium-base binary alloys in oxygen-saturated liquid sodium Noda, T.; Suzuki, T. Corrosion Science, Vol. 16, Issue 7 https://doi.org/10.1016/0010-938X(76)90062-7	journal	January 1976
Carbon Equilibrium Relationships for Austenitic Stainless Steels in a Sodium Environment Shiels, Stuart A.; Bagnall, Chris; Schrock, Steven L. Nuclear Technology, Vol. 23, Issue 3 https://doi.org/10.13182/NT74-A15919	journal	September 1974
Sodium fast reactor evaluation: Core materials Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon Journal of Nuclear Materials, Vol. 392, Issue 2 https://doi.org/10.1016/j.jnucmat.2009.03.021	journal	July 2009
The role of dissolved oxygen in the corrosion of titanium and zirconium by liquid sodium Barker, M. G.; Wood, D. J. Journal of the Chemical Society, Dalton Transactions, Issue 22 https://doi.org/10.1039/dt9720002451	journal	January 1972
Carbon Transfer Behavior of Materials for LMFBR's Steam-Generator in Flowing Sodium Matsumoto, Keishi; Ohta, Yoshio; Kataoka, Tadayuki Journal of Nuclear Science and Technology, Vol. 13, Issue 4 https://doi.org/10.1080/18811248.1976.9734012	journal	April 1976
The role of dissolved oxygen in the corrosion of niobium and tantalum by liquid sodium Addison, C. C.; Barker, M. G.; Wood, D. J. Journal of the Chemical Society, Dalton Transactions, Issue 1 https://doi.org/10.1039/dt9720000013	journal	January 1972
Structural materials challenges for advanced reactor systems Yvon, P.; Carré, F. Journal of Nuclear Materials, Vol. 385, Issue 2 https://doi.org/10.1016/j.jnucmat.2008.11.026	journal	March 2009
Critical review of the thermodynamics of the Na-Al-O system Barsoum, M. Journal of Materials Science, Vol. 25, Issue 10 https://doi.org/10.1007/BF00581098	journal	October 1990

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