Materials Systems for the Future of Fusion Energy: Poster Session
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, Baylor University

Tuesday 5:30 PM
March 1, 2022
Room: Exhibit Hall C
Location: Anaheim Convention Center


N-21: Dislocation Loop Formation in Self-ion Irradiated Ultra-high Purity Fe-Cr Alloys: Yao Li1; Arunodaya Bhattacharya2; Yajie Zhao1; Ling Wang2; Steven Zinkle1; 1University of Tennessee Knoxville; 2Oak Ridge National Laboratory
    Understanding the loop formation in Fe-Cr alloys as a function of irradiation temperature, Cr concentration, dose, and dose rate is an important aspect of radiation resistance. 8 MeV Fe ion irradiations were performed at 10-5 - 10-4 dpa/s, 350-450°C on ultra-high purity Fe and Fe-Cr alloys (3-18wt.%Cr) up to ~0.35 and 3.5 midrange dpa. Using TEM, we observed loop size reduction and density increase with increasing Cr, an increasing ½ <111> fraction with higher Cr concentration, and petal-shaped loops on {100} planes at 450°C. The fraction of ½ <111> loops reduced from ~20% to ~2% with increasing temperature. The <100> fraction decreased with increasing damage depth. 80% of the observed loops were ½ <111> in beyond-damage region in pure Fe at 350°C. Our observations contradict the glide and interaction models as the main mechanism for <100> loop production. Cr segregation onto dislocation loops was quantified for Cr < 10wt.%.

N-23: Heavy Ion Irradiation Studies on an Additively Manufactured 316LN Stainless Steel at Elevated Temperatures: Zhongxia Shang1; Cuncai Fan1; Yinmin Wang2; Lin Shao3; Haiyan Wang1; Xinghang Zhang1; 1Purdue University; 2University of California, Los Angeles; 3Texas A&M University
    Additive manufacturing provides ample opportunities to fabricate structural materials and nuclear reactor components with complex geometries and excellent mechanical properties. However, irradiation response of additively manufactured steels is still less well understood. Here, an additively manufactured 316LN stainless steel was studied via in-situ and ex-situ heavy ion irradiation at 400 to 450˚C with a focus on the roles of solidification cellular structures. It was found that dislocations trapped along cell walls exhibit an enhanced irradiation stability compared with those induced by severe plastic deformation. Moreover, the cellular walls with trapped dislocations can serve as effective defect sinks, thus reduce Frank loop density, and mitigate the magnitude of solute segregations along the cellular walls comparing with high angle grain boundary. These findings were compared with previous neutron irradiation studies on wrought 316 counterparts. The current work advances the understandings on high-temperature irradiation response of additively manufactured steels for nuclear reactor applications.

N-24: Microstructure Deformation and Possible Densification of Tungsten in High Heat Flux Conditions: Minsuk Seo1; Ke Wang1; John Echols2; Leigh Winfrey1; 1The Pennsylvania State University; 2Oak Ridge National Laboratory
    Tungsten is a candidate material for the divertor component in future tokamak reactors. However, high heat flux during disturbances (1-10GW/m2) can melt the surface and cause severe surface damage. In this study, high heat fluxes (12.5-46.3GW/m2) were applied using electrothermal plasma to simulate and investigate surface damage. TEM microphotographs showed porous resolidification and highly deformed microstructures. A total of 16 nanoindentations were taken from the matrix to get an average nanohardness. Dislocation density is estimated to be ~1016/m2. The hardening and possible densification are due to the high dislocation density environment. Due to the FIB process, Ga-focused ion beam induced amorphous artifacts formed at the pores. The pores are likely formed from the stress concentration as dislocations increased from as machined to ~1017/m2. We suggest that ductile fracture occurred at the pore during the high heat event.

N-26: Oxide-dispersion-strengthened Steel Processing by Additive Manufacturing of Gas Atomization Reaction Synthesis (GARS) Powders: Matthew Dejong1; Ryan Schoell1; Sourabh Saptarshi1; Sarah Timmins1; Emma White2; Iver Anderson2; Djamel Kaoumi1; Christopher Rock1; Timothy Horn1; 1North Carolina State University; 2Iowa State University
    Oxide-Dispersion-Strengthened steels have properties such as a high thermal conductivity, low thermal expansion coefficient, low void swelling during exposure to neutron irradiation, and high elevated-temperature strength that make them desirable for nuclear engineering structural applications [1]. A high density of Ti-Y oxide particles dispersed in the F/M matrix is at the source of many of these properties. Fe-15%Cr powders with additions of yttrium, titanium, and oxygen were obtained by gas atomization reaction synthesis (GARS), and printed via Laser-Powder-Bed-Fusion (L-PBF) Additive Manufacturing with controlled oxygen environments of Argon, 1% oxygen, 5% oxygen, and Air in layers. The resulting microstructure showed the formation of the oxide particles. Transmission Electron Microscopy (TEM) characterization was conducted on focused ion beam (FIB) lift-outs from each region. The impact of oxygen environment during L-PBF on the particle size distribution and overall density are discussed.

N-27: Polycrystal Homogenization Modelling Accounting for Channeling in Irradiated Metals and Alloys: Diogo Gonçalves1; Maxime Sauzay1; Laurent Dupuy1; 1Université Paris-Saclay, Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA)
    The effects of irradiation on the microstructure of metallic materials have been widely reported in the literature. Irradiation defects lead to a meaningful increase in yield stress and ultimate strength, a decrease in the work-hardening rate and a decrease in ductility. Finally, the intense plastic slip localization within thin slip bands called "channels" is often reported. A polycrystalline homogenization model accounting for the production of channels inside grains is proposed, as observed experimentally in irradiated ferritic and tempered ferritic-martensitic steels. All material parameters are issued from experimental observations (SEM, TEM), dislocation dynamic calculations and single crystal tensile curves, avoiding any inverse fitting using experimental polycrystalline curves. Due to the low channel aspect ratio, weak internal stresses inside irradiated polycrystals are obtained, and the predictions are in good agreement with experimental observations (very low strain-hardening and low number of activated slip systems).

N-28: Promoting Radiation Resistance in Metallic Solid Solutions via the Use of Multiple Synergistic Solutes: Soumyajit Jana1; Thomas Schuler2; Pascal Bellon1; Robert Averback1; 1University of Illinois Urbana Champaign; 2Cea Saclay
    One strategy to increase radiation resistance of solid solutions relies on the addition of solutes that bind strongly to point defects, in particular to vacancies, to slow down their migration and promote recombination. In most alloy systems, however, vacancy-binding solutes tend to have high mobility and to experience solute drag, which progressively removes solute from grain interiors to the sinks. A novel approach to overcome this limitation is considered by using two synergistic solutes, B and C in an A matrix, where the B solute binds to vacancies while the C solute is a slow diffuser that binds to B. Kinetic Monte Carlo simulations and KineCluE transport calculations are employed to screen for promising solutes, such as Sn and Ni in Cu, by quantifying solute efficiency in promoting recombination over solute drag. Radiation resistance is shown to be significantly improved when B and C solutes form clusters prior to irradiation.

N-29: Thermal and Mechanical Characterization of W-Cu Composites for Next Generation Fusion Devices: Elena Tejado Garrido1; Alexander Müller2; Jeong-Ha You2; J.Y. Pastor1; 1Universidad Politécnica de Madrid; 2Max-Planck-Institut für Plasmaphysik
     W–Cu metal matrix composites (MMC) offer a unique combination of thermophysical and mechanical properties which are of great interest for the development of the heat-exhaust system (divertor) of the European demonstrational fusion reactor (DEMO). The high tensile strength and low thermal expansion coefficient of W are combined with a coherent Cu-based matrix that ensures high thermal conductivity and ductility of the system. The excellent wetting and bonding characteristics between them is another beneficial feature that is also valid for the joining between the W armour and the heat sink.In this contribution, the thermomechanical properties of novel W-Cu MMCs are presented. Focus is put on the microstructure and tensile strength up to 800 °C under a high vacuum atmosphere. Although strong degradation is observed at the highest temperature tested, the mechanical properties at operational temperatures, i.e., below 350 °C, remain rather high and even better than composites materials reported previously.