THE EFFECT OF ZR DOPING ON THE CORROSION BEHAVIOR OF FE-CR-AL-W-TI-ZR-Y?O? ODS FERRITIC-MARTENSITIC ALLOYS IN PB ENVIRONMENTS AT TEMPERATURE 600 °C

• Main Author: Wafda, Hakimul
• Format: Theses
• Language: Indonesia
• Online Access: https://digilib.itb.ac.id/gdl/view/86913
• Institution: Institut Teknologi Bandung

Nuclear energy is recognized as an environmentally friendly energy source with lower emissions compared to other energy sources. One of the latest innovations in nuclear technology is the Lead-Cooled Fast Reactor (LFR), which uses lead (Pb) as the primary coolant. Lead offers advantages such as high thermal conductivity and chemical stability at elevated temperatures. However, a significant challenge lies in material corrosion, which can reduce the lifespan and safety of the reactor. To address this challenge, Oxide Dispersion Strengthened (ODS) steel-based materials have emerged as promising candidates due to their superior resistance to high temperatures, corrosion, and oxidation. This study aims to investigate the effect of zirconium (Zr) addition on Fe-9Cr-5Al-1W-0.3Ti-xZr-0.35Y?O? material in terms of morphology, phase, elemental distribution, and corrosion resistance in a Pb environment. The material was synthesized using the mechanical alloying method with planetary ball milling (PBM) for 8 hours. After milling, the powder was pressed and sintered at 1200°C for 2 hours to form solid material. Characterizations were performed on both powder and solid samples. SEM-EDS was used to analyze morphology and elemental distribution, while XRD was employed to identify the phases formed. The analysis revealed that the powder material exhibited irregular morphology with a tendency for agglomeration. The average powder particle size varied among samples, with sizes of 8.468 µm for Sample 1 (0% Zr), 8.049 µm for Sample 2 (0.5% Zr), 10.103 µm for Sample 3 (1% Zr), and 9.133 µm for Sample 4 (1.5% Zr). Homogeneous elemental distribution in solid material was confirmed by SEM-EDS, indicating effective sintering. The initial phases identified included alpha-ferrite and martensite, contributing to high mechanical strength and corrosion resistance. Corrosion testing was conducted under two environmental conditions: closed (oxygen-deficient) and open (oxygensaturated). In the closed environment, Samples 1 (0% Zr) and 2 (0.5% Zr) did not form protective oxide layers, allowing Pb penetration to depths of 5.741 µm and 11.917 µm, respectively. In contrast, Samples 3 (1% Zr) and 4 (1.5% Zr) formed Fe oxide layers with thicknesses of 24.119 µm and 31.806 µm, respectively, effectively preventing Pb penetration. Sample 4 exhibited an additional Fe-Cr spinel layer with a thickness of 9.133 µm, serving as a supplementary protective layer. In the oxygen-saturated environment, the influence of Zr became more pronounced. Samples with low Zr content (0% and 0.5%) showed thinner oxide layers. Conversely, samples with higher Zr content (1% and 1.5%) developed thicker oxide layers. Sample 3 (1% Zr) demonstrated optimal oxide layer thickness accompanied by a broad diffusion zone of 46.368 µm. However, in Sample 4 (1.5% Zr), although the oxide layer was thicker, the diffusion zone was narrower (26.373 µm) due to Zr saturation, which hindered the diffusion mechanism. This saturation also caused Zr segregation at grain boundaries, limiting the migration of Fe and Cr elements to the material's surface. Overall, this study shows that Zr addition significantly affects the microstructure and corrosion resistance of ODS materials. A Zr concentration of 1% provides optimal conditions with a stable Fe-Cr spinel layer, an extensive diffusion zone, and the best resistance to Pb penetration. At 1.5% Zr, a reduction in the Fe oxide layer thickness and diffusion zone was observed, potentially affecting the material's long-term resistance, as evidenced by penetration depths of up to 37.230 µm. This research is expected to contribute to understanding the influence of Zr on the corrosion resistance of ODS materials in Pb environments and serve as a foundation for developing corrosion-resistant materials for Generation IV nuclear reactors, particularly the Lead-Cooled Fast Reactor (LFR).