Corrosion Behavior of Nuclear Materials in Oxalic Acid for Reductive Decontamination Applications

This study investigated the corrosion behavior of four structural materials of nuclear reactors, austenitic stainless steel (SUS304), Incoloy 800H, zirconium alloy (Zr-2.5Nb), and carbon steel (SAB106) in 20 mM oxalic acid (OA) at 95°C under static immersion conditions. OA is commonly used in chemical oxidation reduction decontamination for nuclear decommissioning due to its reductive chelation capability.The results indicated that OA induced moderate dissolution, primarily through chelation with Fe2+/Fe3+ ions, while exhibiting surface passivation tendencies that restrained uncontrolled attack. A weight loss analysis showed the highest mass reduction for SAB106, followed by SUS304, while Incoloy 800H and Zr-2.5Nb exhibited minimal changes. The highest Fe concentration, 258.87 ppm after 9 h, was observed in SAB106.Elemental analysis data confirmed significant oxygen enrichment on SAB106 (39.67%), indicating aggressive oxide formation, whereas SUS304, Incoloy 800H, and Zr-2.5Nb remained comparatively stable. Scanning electron microscopy images revealed pronounced pore formation and intergranular attack in SAB106, while SUS304 retained a smooth morphology with shallow pits. Incoloy 800H and Zr-2.5Nb showed only fine micropores due to its Ni-rich matrix.These results demonstrated that OA induces severe localized corrosion in Fe-dominant alloys, but exerts only mild effects on Ni-Cr alloys, highlighting the compositional dependence of corrosion resistance. The findings provide insights into alloy compatibility during OA-based reductive decontamination processes, particularly for primary coolant components in pressurized heavy water reactor and pressurized water reactor systems.