Abstract
Ferritic/martensitic steels are one of the best candidates for structural materials for high dose applications in next generation nuclear reactors. The composition of structural materials must be optimized for reliable service during irradiation. This study reports the effect of interstitial nitrogen on radiation response in 12Cr ferritic/martensitic HT9 steels having a controlled nitrogen concentration. Results show that a high amount of ‘free’ nitrogen in the matrix stabilizes the interstitial clusters which leads to (i) larger loop sizes (ii) lower loop density and (iii) slightly reduced radiation induced hardening. It also affects diffusion mechanism of Ni and formation of Ni/Si-rich precipitates.
| Original language | English |
|---|---|
| Pages (from-to) | 145-150 |
| Number of pages | 6 |
| Journal | Scripta Materialia |
| Volume | 189 |
| DOIs | |
| State | Published - Dec 2020 |
Funding
This research was partially supported by the DOE-NE Fuel Cycle Research and Development Program under the Contract number DE-AC52-06NA25396. Work was performed at Los Alamos National Laboratory . Los Alamos National Laboratory is an affirmative action/equal opportunity employer, and is operated by Triad National Security, LLC for the National Nuclear Security Administration of U.S. Department of Energy under contract 89233218CNA000001 . APT was conducted at ORNL's Center for Nanophase Materials Sciences (CNMS), which is a U.S. DOE Office of Science User Facility. The authors would like to thank James Burns for help with APT sample preparation and experimentation. This manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy . The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. This research was partially supported by the DOE-NE Fuel Cycle Research and Development Program under the Contract number DE-AC52-06NA25396. Work was performed at Los Alamos National Laboratory. Los Alamos National Laboratory is an affirmative action/equal opportunity employer, and is operated by Triad National Security, LLC for the National Nuclear Security Administration of U.S. Department of Energy under contract 89233218CNA000001. APT was conducted at ORNL's Center for Nanophase Materials Sciences (CNMS), which is a U.S. DOE Office of Science User Facility. The authors would like to thank James Burns for help with APT sample preparation and experimentation. This manuscript has been authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.
| Funders | Funder number |
|---|---|
| DOE-NE Fuel Cycle Research and Development Program | DE-AC52-06NA25396 |
| United States Government | |
| U.S. Department of Energy | 89233218CNA000001 |
| Office of Science | DE-AC05-00OR22725 |
| National Nuclear Security Administration | |
| Los Alamos National Laboratory |
Keywords
- Dislocation loops
- Ferritic/martensitic alloys
- G-phase formation
- Interstitial effect
- Radiation induced hardening