Materials Systems for the Future of Fusion Energy: Radiation Effects in FeCr Alloys and ODS Steels
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Additive Manufacturing Committee, TMS: Computational Materials Science and Engineering Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Jason Trelewicz, Stony Brook University; Kevin Field, University of Michigan; Takaaki Koyanagi, Oak Ridge National Laboratory; Yuanyuan Zhu, University Of Connecticut; Dalong Zhang, Pacific Northwest National Laboratory
Wednesday 2:00 PM
March 2, 2022
Room: 203A
Location: Anaheim Convention Center
Session Chair: Gary Was, University of Michigan; William Cunningham, University of California Santa Barbara
2:00 PM
NOW ON-DEMAND ONLY - Post-irradiation Examination of High-dose Ion Irradiated MA956 ODS Alloy: Yu Lu1; Yaqiao Wu1; Ramprashad Prabhakaran2; Megha Dubey1; Lin Shao3; Jing Wang2; Dalong Zhang2; 1Boise State University/Center for Advanced Energy Studies; 2Pacific Northwest National Laboratory; 3Texas A&M University
Oxide dispersion strengthened (ODS) alloys are candidate cladding materials for the next generation advanced nuclear reactors due to demonstrated excellent resistance to irradiation damage, high-temperature creep and superior mechanical properties under high temperature and radiation environment. MA956 ODS alloys have been being studied, however, only very limited data was reported regarding to its microstructural evolution under irradiation to date, which makes it difficult to understand its performance fundamentally. In this study, the MA956 ODS alloys are ion irradiated to different doses (2.5, 50 and 100 dpa) under different temperatures (190°C and 320°C). Post-irradiation characterizations are performed by using scanning transmission electron microscopy (STEM), atom probe tomography (APT) and nanoindentation techniques. The irradiation induced defects and the effects of irradiation dose and temperature on the evolution of the dispersoid are illustrated and discussed in details.
2:20 PM Invited
NOW ON-DEMAND ONLY - Low Temperature Hardening-embrittlement in Neutron Irradiated ODS Steels: Arunodaya Bhattacharya1; Samara Levine2; Xiang Chen1; Takashi Nozawa3; Steven Zinkle2; Yutai Katoh1; 1Oak Ridge National Laboratory; 2University of Tennessee; 3National Institutes for Quantum and Radiological Science and Technology (QST)
A major challenge for the fusion reactor first-wall/blanket is the well-known issue of irradiation-induced low temperature hardening-embrittlement (LTHE) of structural steels, resulting in loss of fracture toughness. While LTHE in RAFM steels significantly improves for irradiation temperatures ≥ 350 °C, the susceptibility of ODS steels to LTHE is not well-evaluated. Using post-irradiation mechanical property testing, we reveal compelling evidence of LTHE in various ODS steels after neutron irradiations in the high flux isotope reactor between 300-500 °C and doses up to ~20 dpa. The LTHE scenario in ODS is compared with RAFM steels to identify and rationalize potential design challenges such emerging results may pose. Research sponsored the U.S. Department of Energy, Office of Fusion Energy Sciences under contact DE-AC05-00OR22725 with UT-Battelle LLC.
2:50 PM Invited
Recent Progress in Understanding Fundamental Radiation Degradation Processes: Steven Zinkle1; Yan-Ru Lin1; Yajie Zhao1; Samara Levine1; Yao Li1; Zehui Qi1; Arunodaya Bhattacharya1; Shradha Agarwal1; 1University of Tennessee
The intense neutron fluxes and He generation rates in the first wall and blanket regions of proposed DT fusion reactors generally accelerate the five key radiation damage degradation mechanisms in structural materials. TEM analysis of dual ion irradiated Fe-Cr and steel specimens irradiated at 0.1, 10 and 50 appm He/dpa and 400-550oC reveals a transition from monomodal to bimodal cavity size distributions with increasing He/dpa and peak swelling near fusion-relevant 10 appm He/dpa, along with a shift in peak cavity swelling to higher temperatures at higher He/dpa. A modified precipitate stability model has been developed to quantify the competing effects of ballistic dissolution and radiation enhanced precipitate regrowth in structural alloys. Improved insight on mechanisms for formation and conversion of 1/2<111> and <100> loops in Fe alloys are revealed by introducing novel nanostructured heterogeneous sink structures. The interrelationship between high temperature He embrittlement, stress, He/time and creep mechanisms are explored.
3:20 PM Break
3:45 PM
Effect of Cr and He on Cavity Swelling in Dual-Ion Irradiated High Purity Fe-Cr Alloys: Yan-Ru Lin1; Arunodaya Bhattacharya2; Jean Henry3; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory; 3CEA
Simultaneous dual-beam irradiations on ultra-high purity Fe and Fe-Cr alloys (3-14 wt% Cr) at 400-550°C were performed to improve understanding of cavity formation in ferritic alloys. 8 MeV Ni ions (relatively wide safe analysis zone with a midrange dose ~30 dpa) and helium production rates of 0.1, 10, and 50 appm He/dpa were selected. Cavities were observed by TEM in all irradiated samples. A bimodal cavity size distribution was observed in the 10 and 50 He appm/dpa samples, but not for 0.1 He appm/dpa. The peak swelling temperature was noticeably higher for the Fe-Cr alloys compared to pure Fe. Higher He/dpa content resulted in a higher peak swelling temperature for pure Fe. Peak cavity swelling occurs at intermediate He/dpa (fusion relevant condition). Swelling as a function of the Cr content is non-monotonic and could be controlled by solute trapping of defects and α’ precipitates leading to increased recombination.
4:05 PM Invited
Synergies between H and He in Radiation-induced Swelling of Candidate Fusion Blanket Materials: Logan Clowers1; Zhijie Jiao1; Gary Was1; 1University of Michigan
The synergistic effect of H and He on radiation-induced swelling was studied on alloys F82H, Fe8Cr2W - a high purity analog of F82H, and CNA3, a castable nanostructured alloy. Single beam (Fe2+), dual beam (Fe2+/He2+ and Fe2+/H+), and triple beam (Fe2+/He2+/H+) irradiations were conducted over the temperature range 400-550°C to 50 dpa at a rate of 1 x 10-3 dpa/s with He and H injection rates of 10 and 40 appm/dpa, respectively. Post-irradiation characterization via bright field TEM and high-angle annular dark-field STEM was performed on all irradiated conditions to characterize the cavity size distribution and determine the effects of H/He injection on cavity microstructure. In all three alloys, hydrogen co-injection with helium led to an increase in swelling over that from helium injection alone, and to varying degrees depending on the alloy. Results will be presented in light of the role of hydrogen in increasing cavity growth and swelling.