Materials in Nuclear Energy Systems (MiNES) 2021: Advanced and Novel Materials- Session IV
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory
Thursday 1:10 PM
November 11, 2021
Room: Urban
Location: Omni William Penn Hotel
Session Chair: Kevin Field, University of Michigan
1:10 PM Invited
Novel Nickel-based Alloys for Molten Salt Fast Reactor Structural Applications: Vijay Vasudevan1; 1University of Cincinnati
Molten chloride and fluoride salt fast reactors (MSR) are under active development because they offer several operational and safety advantages over other types of reactors. Modern designs require structural materials with superior corrosion, creep, thermomechanical fatigue, irradiation damage and Helium bubble and tellurium-induced grain boundary embrittlement resistance to high temperatures of 750-950°C. Nickel-base alloys generally perform better than all other alloys studied to date, though none yet have the set of property requirements to meet these demanding conditions. In this talk, I will report on the development of the next generation of nickel-base alloys for MSRs utilizing an ICME approach combined with a detailed experimental processing, testing, characterization and modeling program. Results of the alloy design strategy, phase equilibria and transformations and evolution of microstructure and high temperature mechanical properties will be presented and discussed. The irradiation and corrosion behavior in molten chloride salt of selected alloys will also be presented. Future directions will be discussed.
1:50 PM
Contextualizing Dispersoid Evolution within Friction Stir Welded and Ion Irradiated MA956: Elizabeth Getto1; Nicholas Nathan1; Jack McMahan1; Brad Baker1; Stephen Taller2; 1United States Naval Academy; 2Oak Ridge National Laboratory
Understanding the co-evolution of dispersoids with the dislocation loops and network is critical for a comprehensive understanding of the response of friction stir welded (FSW) and oxide-dispersion-strengthened (ODS) steels to radiation. Ion irradiations were performed on FSW and ODS Fe-Cr-Al steel MA956 with 5 MeV Fe++ ions from 400 to 500°C at doses ranging from 50 to 200 dpa. Characterization was performed primarily with scanning transmission electron microscopy and energy-dispersive x-ray spectroscopy to investigate the Y-Al-O dispersoids, voids and dislocations. The co-evolution of the microstructure was explained as a function of the evolving defect kinetics, utilizing rate theory to calculate point defect concentrations, determine defect partitioning among sinks, and the increasing diffusivity of vacancies. Regardless of temperature, the dispersoids increased in diameter and decreased in number density, which was attributed to an Ostwald coarsening mechanism supported by calculations of the radiation enhanced diffusion and ballistic dissolution.
2:10 PM
Temperature-controlled Friction Stir Welding: A Potential Crack Repair Technology for 304L Stainless Steel Spent Nuclear Fuel-dry Storage Canisters (SNF-DSC): Saumyadeep Jana1; Indrajit Charit2; Krishnan Raja2; Madhumanti Bhattacharyya3; Anirban Naskar2; 1Pacific Northwest National Laboratory; 2University of Idaho; 3Indian Institute of Technology-Dhanbad
Dry storage canisters for storing spent nuclear fuel rods are usually fabricated from 304L austenitic stainless steels, since it offers excellent resistance to atmospheric corrosion. However, 304L SS is susceptible to pitting corrosion when exposed to chloride containing environments. Chloride-induced stress corrosion cracking (CI-SCC) can lead to early crack nucleation around heat affected zones and fusion weld seams in 304L SS. Thus, it is critical to eliminate any such cracks for a SNF-DSC to perform reliably. In the present study, the feasibility of Friction Stir Welding (FSW), as a low-temperature crack repair method has been explored. Multiple temperature-controlled FSW runs were carried out on 0.5” thick 304L SS plates in a gantry-type machine using a PCBN tool. Weld parameters were optimized to achieve weld temperatures that range from ~700°C to 900°C. The presentation will provide a summary of various characterization details (microstructural, mechanical, electrochemical etc.) of these temperature-controlled welds.
2:30 PM
Thermal Annealing and Irradiation Behavior of Ultrafine-grained and Nanocrystalline FeCrAl Alloys: Haiming Wen1; Maalavan Arivu1; Rinat Islamgaliev2; 1Missouri University of Science and Technology; 2Ufa State Aviation Technical University
FeCrAl alloys are leading candidate materials for cladding of accident tolerant fuels in light water reactors replacing Zircaloy owing to their high temperature strength and corrosion resistance in steam environments. However, FeCrAl alloys suffer from embrittlement after long-time aging at ~500 oC and lower, due to α’ Cr precipitation. This phenomenon is detrimental to the structural performance and corrosion resistance. Another major issue for FeCrAl alloys in nuclear environments is the irradiation-enhanced α′ precipitation, leading to irradiation-induced hardening and embrittlement. Bulk ultrafine-grained and nanocrystalline metals possess drastically higher strength than their conventional coarse-grained counterparts, and are anticipated to have significantly enhanced irradiation tolerance. In this study, ultrafine-grained and nanocrystalline FeCrAl alloys were manufactured by equal-channel angular pressing and high-pressure torsion, respectively. The thermal annealing and irradiation behavior of these materials were carefully studied. Results indicated that reducing grain size can hinder both thermally induced and irradiation enhanced α’ Cr precipitation.
2:50 PM
Finding a Balance in FeCrAl Alloys: Optimization of Alloy Chemistry for Balanced Properties: Andrew Hoffman1; Vipul Gupta1; Fabiola Cappia2; Raul Rebak1; 1GE Research; 2Idaho National Laboratory
FeCrAl alloys have shown great promise for use as accident tolerant fuel cladding due to their excellent high temperature steam oxidation resistance, good hydrothermal corrosion resistance, and desirable mechanical properties. One concern, however, is the formation of α’, a Cr enriched phase which precipitates in Cr bearing ferritic alloys at temperatures around or below 500°C. This presentation will give an overview of GE’s development of FeCrAl alloys and future plans to combine experiments and regression analysis using machine learning to optimize alloy chemistry. Considerations of microstructure and chemistry on corrosion, mechanical properties, and phase stability will be discussed. Attention will also be given to the effects of radiation on such properties.
3:10 PM Break