Corrosion Behavior of Modified F/M Steel with Ti and Dispersed Oxides: Y2O3 and ZrO2 Under High Temperature in Static Liquid Lead
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Abstract
This study investigates the corrosion resistance of Oxide Dispersion Strengthened (ODS) steel in a lead environment, which is crucial for Lead-Cooled Fast Reactors (LFR), a type of Generation IV nuclear reactor. To improve corrosion resistance, two types of oxides-yttrium oxide (Y2O3) and zirconium oxide (ZrO2)-were added individually and in combination to the ODS steel. The samples were synthesized via powder metallurgy and characterized using optical microscopy, X-ray diffraction (XRD), scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDX), and Vickers hardness testing. Corrosion testing was conducted in static liquid lead at 550°C for 75 hours. The results showed that the addition of Y2O3 and ZrO2 significantly refines grain structure, increases hardness, and promotes the formation of stable, protective oxide layers. Particularly, the dual-oxide (Y2O3+ ZrO2) sample exhibited the most uniform and effective oxide barrier, indicating improved resistance to lead corrosion. These findings demonstrate that dual-oxide dispersion is a promising strategy for improving the durability of structural materials in LFR applications.
References
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Giuranno, D., S. Amore, R. Novakovic, C. Tomasi, and E. Ricci (2021). Wetting Property of Liquid Pb on Different Steel Candidates as Structural Materials for the Generation IV Nuclear Reactors. High Temperatures - High Pressures, 50(1), 49–61.
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Wang, J., S. Liu, B. Xu, J. Zhang, M. Sun, and D. Li (2021b). Research Progress on Preparation Technology of Oxide Dispersion Strengthened Steel for Nuclear Energy. International Journal of Extreme Manufacturing, 3(3), 032001.
Xiao, Z., J. Liu, Z. Jiang, and L. Luo (2022). Corrosion Behavior of Refractory Metals in Liquid Pb at 1000 °C for 1000 h. Nuclear Engineering and Technology, 54(6), 1954–1961.
Yaskiv, O. I. and V. M. Fedirko (2014). Oxidation/Corrosion Behaviour of ODS Ferritic/Martensitic Steels in Pb Melt at Elevated Temperature. International Journal of Nuclear Energy, 2014(1), 657689.
Zhu, C., W. Zhang, M. Zhou, L. Lv, Y. Zhong, Q. Li, and J. Yang (2023). Effect of Y2O3 Content on Microstructure and Corrosion Behavior of ZrO2 Coatings in Liquid Lead-Bismuth Eutectic. Materials Chemistry and Physics, 309, 128392.
Zinkle, S. J., J. L. Boutard, D. T. Hoelzer, A. Kimura, R. Lindau, G. R. Odette, and H. Tanigawa (2017). Development of Next Generation Tempered and ODS Reduced Activation Ferritic/Martensitic Steels for Fusion Energy Applications. Nuclear Fusion, 57(9), 092005.
Şahin, S. and Y. Wu (2018). 3.14 Fission Energy Production. In Comprehensive Nuclear Materials.
Bari, B. E. and D. Irwanto (2021). Wall Effect Analysis in Thermal-Hydraulics Aspect of HTR-10 and PR-3000 Reactors Using Different Porosity Models. Science and Technology Indonesia, 6(4), 235–241.
Bassini, S., S. Cataldo, C. Cristalli, A. Fiore, C. Sartorio, M. Tarantino, and M. Frignani (2020). Material Performance in Lead and Lead-Bismuth Alloy. In Comprehensive Nuclear Materials: Second Edition. Elsevier, 218–241.
Chen, L., M. Wang, V. Tsisar, C. Schroer, and Z. Zhou (2020). Investigation of Microstructure and Liquid Pb Corrosion Behavior of a Fe-18Ni-16Cr-4Al Base Alumina-Forming Austenitic Stainless Steel. Materials Research Express, 7(2), 026533.
Del Giacco, M., A. Weisenburger, P. Spieler, F. Zimmermann, F. Lang, A. Jianu, and G. Mueller (2012). Experimental Equipment for Fretting Corrosion Simulation in Heavy Liquid Metals for Nuclear Applications. Wear, 280–281, 46–53.
Dömstedt, P., M. Lundberg, and P. Szakalos (2019). Corrosion Studies of Low-Alloyed FeCrAl Steels in Liquid Pb at 750 °C. Oxidation of Metals, 91(3), 511–524.
Giuranno, D., S. Amore, R. Novakovic, C. Tomasi, and E. Ricci (2021). Wetting Property of Liquid Pb on Different Steel Candidates as Structural Materials for the Generation IV Nuclear Reactors. High Temperatures - High Pressures, 50(1), 49–61.
Gong, X., M. P. Short, T. Auger, E. Charalampopoulou, and K. Lambrinou (2022). Environmental Degradation of Structural Materials in Liquid Pb- and Pb-Bi Eutectic-Cooled Reactors. Progress in Materials Science, 126, 100920.
Henry, J. and S. A. Maloy (2017). Irradiation-Resistant Ferritic and Martensitic Steels as Core Materials for Generation IV Nuclear Reactors. In Structural Materials for Generation IV Nuclear Reactors, 329–355.
Jiang, H., X. Zhao, D. Wang, Q. Zhu, T. Li, and Y. Lei (2022). Effects of Y2O3 Addition on the Microstructure and Static Pb-Bi Eutectic Thermal Corrosion Behaviors of FeCrAlTiCxY2O3 Laser Clade Coatings. Coatings, 12(11), 1759.
Kimura, A., W. Han, H. Je, K. Yabuuchi, and R. Kasada (2016). Oxide Dispersion Strengthened Steels for Advanced Blanket Systems. Plasma and Fusion Research, 11(1), 2505090–2505090.
Lorusso, P., S. Bassini, A. Del Nevo, I. Di Piazza, F. Giannetti, M. Tarantino, and M. Utili (2018). GEN-IV LFR Development: Status & Perspectives. Progress in Nuclear Energy, 105, 318–331.
Martinelli, L., K. Ginestar, V. Botton, C. Delisle, and F. Balbaud-Célérier (2020). Corrosion of T91 and Pure Iron in Flowing and Static Pb-Bi Alloy Between 450 °C and 540 °C: Experiments, Modelling and Mechanism. Corrosion Science, 176, 108897.
Navas, M. and R. Hernández (2018). Compatibility of Structural Materials with Lead and Lead Bismuth Eutectic for CSP Applications. In AIP Conference Proceedings, volume 2033. American Institute of Physics Inc.
Parida, P. K., A. Dasgupta, S. Bajpai, T. Sakthivel, and R. Mythili (2023). Synthesis and Characterization of 9Cr ODS F/M Steel with Optimized Y2O3 and Ti Content and Its Comparison with P91 Steel. Powder Technology, 425, 118564.
Ricci, E., D. Giuranno, G. Canu, S. Amore, and R. Novakovic (2018). Corrosion Behaviour of Oxide Dispersion Strengthened Iron-Chromium Steels in Liquid Pb at 973 K. Materials and Corrosion, 69(11), 1584–1596.
Tsisar, V. and O. Yeliseyeva (2007). Oxidation of Armco-Fe and Steels in Oxygen-Saturated Liquid Pb. Materials at High Temperatures, 24(2), 93–101.
Wafda, H., D. H. Prajitno, E. A. Basuki, A. Syafiq, N. Widiawati, and A. P. A. Mustari (2024). High Temperature Oxidation Behavior of ODS Ferritic Stainless Steel Fe-16Cr4Al-1Ni-0.4Y2O3. Indonesian Journal of Chemistry, 24(5), 1456–1469.
Wafda, H., F. Roswita, D. H. Prajitno, and A. P. A. Mustari (2023). New Alloy Synthesis of Oxide Dispersion Strengthened (ODS) Ferritic Steel Fe-Cr-Al-W-Ti-Zr with Y2O3 Dispersoid as a Candidate Fuel Cladding Material for Nuclear Reactors Using Mechanosynthesis Method. Jurnal Iptek Nuklir Ganendra, 26(2), 72–81.
Wang, G., Z. Wang, and D. Yun (2023). Cladding Failure Modelling for Lead-Based Fast Reactors: A Review and Prospects. Metals, 13(9), 1524.
Wang, H., J. Xiao, H. Wang, Y. Chen, X. Yin, and N. Guo (2021a). Corrosion Behavior and Surface Treatment of Cladding Materials Used in High-Temperature Lead-Bismuth Eutectic Alloy: A Review. Coatings, 11(3), 1524.
Wang, J., S. Liu, B. Xu, J. Zhang, M. Sun, and D. Li (2021b). Research Progress on Preparation Technology of Oxide Dispersion Strengthened Steel for Nuclear Energy. International Journal of Extreme Manufacturing, 3(3), 032001.
Xiao, Z., J. Liu, Z. Jiang, and L. Luo (2022). Corrosion Behavior of Refractory Metals in Liquid Pb at 1000 °C for 1000 h. Nuclear Engineering and Technology, 54(6), 1954–1961.
Yaskiv, O. I. and V. M. Fedirko (2014). Oxidation/Corrosion Behaviour of ODS Ferritic/Martensitic Steels in Pb Melt at Elevated Temperature. International Journal of Nuclear Energy, 2014(1), 657689.
Zhu, C., W. Zhang, M. Zhou, L. Lv, Y. Zhong, Q. Li, and J. Yang (2023). Effect of Y2O3 Content on Microstructure and Corrosion Behavior of ZrO2 Coatings in Liquid Lead-Bismuth Eutectic. Materials Chemistry and Physics, 309, 128392.
Zinkle, S. J., J. L. Boutard, D. T. Hoelzer, A. Kimura, R. Lindau, G. R. Odette, and H. Tanigawa (2017). Development of Next Generation Tempered and ODS Reduced Activation Ferritic/Martensitic Steels for Fusion Energy Applications. Nuclear Fusion, 57(9), 092005.
Şahin, S. and Y. Wu (2018). 3.14 Fission Energy Production. In Comprehensive Nuclear Materials.
Authors
Setyo Hadi, D., Wafda, H., Pramutadi Andi Mustari, A., Trisnawan, V., Widiawati, N., Miftasani, F., & Hadi Prajitno, D. (2025). Corrosion Behavior of Modified F/M Steel with Ti and Dispersed Oxides: Y2O3 and ZrO2 Under High Temperature in Static Liquid Lead. Science and Technology Indonesia, 10(3), 877–888. https://doi.org/10.26554/sti.2025.10.3.877-888
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