|About this Abstract
||2020 TMS Annual Meeting & Exhibition
||Accelerated Materials Evaluation for Nuclear Applications Utilizing Irradiation and Integrated Modeling
||H-10: Radiation Response of HT9 Ferritic/Martensitic Alloys as a Function of Interstitial Content
||Eda Aydogan, Jonathan Gigax, Scott Parker, Benjamin Eftink, Yongqiang Wang, Stuart Maloy
|On-Site Speaker (Planned)
HT-9 ferritic/martensitic steels are one of the best candidates for structural materials in nuclear applications such as the fuel cladding and fuel ducts for fast reactors. Effect of interstitial elements on swelling, radiation induced segregation and damage has been investigated extensively back in 70s and 80s. However, it is still unclear how it effects hardening and void swelling. In this study, HT9 alloys having various nitrogen contents have been ion irradiated up to ~23 dpa at 300 °C. Transmission electron microscopy (TEM), nano-indentation and atom probe tomography (APT) analyses have shown that the dislocation loop size and density, radiation induced hardening and second phase precipitation mechanisms are different in low and high nitrogen alloys. This study sheds light on any discrepancies in the literature and paves the way to improvements in both modelling and structural material development efforts for next generation reactors.