NSUF 23-1837: Evolution of Dispersoid in Austenitic Fe-Cr-Ni Oxide Dispersion Strengthened Alloy in Ion Irradiation
This project is proposed by a collaborative team established by Boise State University (BSU), Idaho National Laboratory (INL), and Texas A&M University (TAMU) to study the evolution of dispersoids in oxide dispersion strengthened (ODS) alloy under irradiation. Oxide dispersion strengthened (ODS) alloys are designed for extreme environments and are candidates for Gen IV nuclear reactor structural and cladding materials. The oxide particles in ODS alloys contribute outstanding radiation resistivity and high-temperature mechanical properties. In addition, dispersoids can hinder the movement of dislocations and restrain creep. The particles can also stabilize grain/sub-grain structure to prevent growth. Moreover, the boundary between the dispersoids and matrix can serve as defect-trapping sites to enhance radiation resistivity. Therefore, the stability of oxide particles is critical for ODS alloys to maintain their integrity. However, understanding the oxide particle evolution mechanism under irradiation is not completed and is limited to certain conditions and structures. This proposal aims to comprehensively investigate the dispersoid radiation response of face-centered cubic (FCC) Fe-Cr-Ni ODS alloys from 40-400 dpa from 300-700°C. The irradiation matrix is designed to capture oxide particle evolution from low to high temperatures and doses. The low-temperature irradiation(300 °C) limits the thermal-driven diffusion. Therefore, radiation-driven diffusion will be the dominant mechanism. The intermediate temperature (500 °C) is close to the maximum swelling temperature for ferritic and austenitic stainless steels under heavy ion irradiation at 475 °C and 575 °C, respectively [6, 7]. High-temperature irradiation (700 °C) is also required to study high-temperature radiation responses of the dispersoids. In addition, since the dpa rate changes with depth during heavy ion irradiations, a single irradiation will allow for the observation of multiple doses. The peak dpa selection of 100 dpa and 400 dpa allows the continuous investigation of oxide particle behavior from approximately 40-400 dpa. The evolution of dispersoids under irradiation is a competitive process between dissolution from radiation damage cascade and thermal diffusion. The process must consider multiple factors, including temperature, interface type, damage rate, and structure. Interface coherency could also alter during the size change. The complex factors and the small size of the oxide particles make the characterization work difficult and challenging to provide a comprehensive investigation. Previous studies of dispersoid behavior under radiation focused more on ferritic ODS alloys, and understanding the mechanism in austenitic ODS alloys is still lacking. Due to the significantly different swelling resistance of BCC versus FCC, the effect of point defect flux on dispersion morphology changes and the interaction between voids and dispersion needs to be systematically studied at a different stage of void growth, which is the key interest of the project. Obtaining data on both dpa dependence and temperature dependence of dispersion evolution is essential to develop predictable modeling capability. After this project, the combined MOOSE-based rate theory and phase-field modeling will be pursued as our following effort.
Additional Info
| Field | Value |
|---|---|
| Abstract | This project is proposed by a collaborative team established by Boise State University (BSU), Idaho National Laboratory (INL), and Texas A&M University (TAMU) to study the evolution of dispersoids in oxide dispersion strengthened (ODS) alloy under irradiation. Oxide dispersion strengthened (ODS) alloys are designed for extreme environments and are candidates for Gen IV nuclear reactor structural and cladding materials. The oxide particles in ODS alloys contribute outstanding radiation resistivity and high-temperature mechanical properties. In addition, dispersoids can hinder the movement of dislocations and restrain creep. The particles can also stabilize grain/sub-grain structure to prevent growth. Moreover, the boundary between the dispersoids and matrix can serve as defect-trapping sites to enhance radiation resistivity. Therefore, the stability of oxide particles is critical for ODS alloys to maintain their integrity. However, understanding the oxide particle evolution mechanism under irradiation is not completed and is limited to certain conditions and structures. This proposal aims to comprehensively investigate the dispersoid radiation response of face-centered cubic (FCC) Fe-Cr-Ni ODS alloys from 40-400 dpa from 300-700°C. The irradiation matrix is designed to capture oxide particle evolution from low to high temperatures and doses. The low-temperature irradiation(300 °C) limits the thermal-driven diffusion. Therefore, radiation-driven diffusion will be the dominant mechanism. The intermediate temperature (500 °C) is close to the maximum swelling temperature for ferritic and austenitic stainless steels under heavy ion irradiation at 475 °C and 575 °C, respectively [6, 7]. High-temperature irradiation (700 °C) is also required to study high-temperature radiation responses of the dispersoids. In addition, since the dpa rate changes with depth during heavy ion irradiations, a single irradiation will allow for the observation of multiple doses. The peak dpa selection of 100 dpa and 400 dpa allows the continuous investigation of oxide particle behavior from approximately 40-400 dpa. The evolution of dispersoids under irradiation is a competitive process between dissolution from radiation damage cascade and thermal diffusion. The process must consider multiple factors, including temperature, interface type, damage rate, and structure. Interface coherency could also alter during the size change. The complex factors and the small size of the oxide particles make the characterization work difficult and challenging to provide a comprehensive investigation. Previous studies of dispersoid behavior under radiation focused more on ferritic ODS alloys, and understanding the mechanism in austenitic ODS alloys is still lacking. Due to the significantly different swelling resistance of BCC versus FCC, the effect of point defect flux on dispersion morphology changes and the interaction between voids and dispersion needs to be systematically studied at a different stage of void growth, which is the key interest of the project. Obtaining data on both dpa dependence and temperature dependence of dispersion evolution is essential to develop predictable modeling capability. After this project, the combined MOOSE-based rate theory and phase-field modeling will be pursued as our following effort. |
| Award Announced Date | 2023-02-08T10:46:58.053 |
| Awarded Institution | None |
| Facility | None |
| Facility Tech Lead | Lin Shao, Yaqiao Wu |
| Irradiation Facility | Accelerator Laboratory |
| PI | Ching-Heng Shiau |
| PI Email | [email protected] |
| Project Type | RTE |
| RTE Number | 4508 |