Accelerated Discovery and Qualification of Nuclear Materials for Energy Applications: Innovative Design and Development of Nuclear Materials
Sponsored by: TMS Structural Materials Division, TMS Materials Processing and Manufacturing Division, TMS: Integrated Computational Materials Engineering Committee, TMS: Nuclear Materials Committee, TMS: Additive Manufacturing Committee
Program Organizers: Yongfeng Zhang, University of Wisconsin; Adrien Couet, University of Wisconsin-Madison; Michael Tonks, University of Florida; Jeffery Aguiar, Lockheed Martin; Andrea Jokisaari, Idaho National Laboratory; Karim Ahmed, Texas A&M University
Thursday 8:30 AM
March 18, 2021
Room: RM 48
Location: TMS2021 Virtual
Session Chair: Julie Tucker, Oregon State University ; Andrea Jokisaari, Idaho National Laboratory
8:30 AM
Compositionally Graded Bulk Specimen: A High-throughput Approach for Nuclear Alloy Development and Qualification: Xiaoyuan Lou1; Jingfan Yang1; Xiang Liu2; Miao Song3; Lingfeng He2; Yongfeng Zhang4; Daniel Schwen2; 1Auburn University; 2Idaho National Lab; 3University of Michigan; 4University of Wisconsin-Madison
This study demonstrated the use of compositionally graded specimens, fabricated by laser additive manufacturing (AM) and post-AM treatment, to accelerate the evaluations of irradiated and unirradiated materials in nuclear environments. The low-level doping (<1 wt%) of minor refractory elements in austenitic stainless steel was selected as the topic of interest, not only because of the difficulty of controlling accurate composition gradient at this level but also due to its scientific significance related to the effects of atomic level heterogeneity. In conjunction with the composition gradient, the study confirmed wrought-like microstructure could be achieved across the sample. The grain structure, composition, and hardness as a function of processing conditions will be reported. The case studies for nuclear specific properties include radiation hardening, oxidation, swelling, and irradiation assisted stress corrosion cracking (IASCC). The advantages and challenges of using compositionally graded design for high-throughput nuclear alloy development and qualification will be discussed.
8:50 AM
A Superb Void Swelling Resistant Type 316L Stainless Steel Developed by Additive Manufacturing Enabled High Throughput Microalloying: Miao Song1; Jingfan Yang2; Xiang Liu3; Xiaoyuan Lou2; Yongfeng Zhang4; Lingfeng He3; Daniel Schwen3; 1University of Michigan; 2Auburn University; 3Idaho National Laboratory; 4University of Wiscousin
Additive manufacturing (AM) has drawn increasing attention in the nuclear industry. However, data are still limited relevant to radiation damage in AM alloys. Previous results indicate that the void swelling resistance of AM 316L stainless steel (SS) was inferior compared to hot-isotropic pressing counterparts in high-temperature self-ion irradiation. Here, we show that the swelling resistance of AM alloys can be significantly improved by microalloying. Gradient 316L SS with 5 different Hf concentrations was printed in a small volume and then irradiated with 5 MeV Fe2+ ions under identical conditions to 50 dpa at 500-600℃. Void swelling decreases monotonously as the concentration of Hf increases to 1 wt.% at all three temperatures. With 1wt.% Hf, the void swelling of microalloyed AM 316L was one-two orders of magnitude lower compared to undoped AM 316L. This investigation demonstrates that the AM enabled high throughput microalloying is promising for nuclear materials development.
9:10 AM
Improving Irradiation Resistance of Alloys by Controlling Defect Diffusion: A Modeling Perspective: Yongfeng Zhang1; Miao Song2; Xiang Liu3; Lingfeng He4; Daniel Schwen4; Xiaoyuan Lou5; 1University of Wisconsin-Madison; 2University of Michigan ; 3Idaho National Laboratory ; 4Idaho National Laboratory; 5Auburn University
Irradiation damages materials by producing lattice defects and inducing undesirable component segregation/precipitation. The amount and the rate of damage accumulation depend critically on the diffusion of defects. Taking austenitic FeNiCr alloy as an example, the feasibility of improving the irradiation resistance of traditional alloys by minor alloying is studied using lattice kinetic Monte Carlo. It’s found that, additives that trap point defects can in general reduce radiation induced segregation (RIS) of Cr and void swelling. However, the strong trapping itself can cause RIS and precipitation of additives. As such, the mitigation effect disappears at high doses when the additive concentrations are low. In alloys with relatively high additive concentrations, the mitigation effect may be sustained at high doses due to the formation of high density of nanosized precipitates. The modeling findings are consistent with experimental observations in the literature and from recent proton irradiation experiments.
9:30 AM Invited
Role of Composition and Thermal Aging on Corrosion Behavior of Duplex Stainless Steels in Pressurized Water Reactors: Julie Tucker1; Pratik Murkute1; Kofi Oware Sarfo1; Isak McGieson1; Melissa Santala1; Yongfeng Zhang2; Liney Arnadottir1; Burkan Isgor1; 1Oregon State University; 2University of Wisconsin - Madison
Dual-phase stainless steels are prevalent in nuclear power system welds and castings throughout the plant. Phase separation in the ferrite phase, due to thermal aging, degrades the mechanical properties of these alloys and also impacts the corrosion resistance. In this project, five duplex stainless steels (2101, 2003, 2205, 2101-weld, and 2209-weld) were aged at 427°C for up to 10,000 hours and characterized for phase separation and corrosion performance in a range of pressurized water reactor chemistries containing LiOH and H3BO3. First-principles studies and experimental results are integrated into a phase-field model to simulate the effects of aging on the nucleation and growth of oxide films as a function of alloy composition. This model is also being used to test hypothesized mechanisms for chloride-induced depassivation, suggested by first-principles studies. This approach will allow us to predict performance for new alloys in these reactor environments.
10:00 AM
Development of Sintered High Strength and Thermal Conductivity Cu-Cr-Nb-Zr Alloy for Fusion Components: Bin Cheng1; Ling Wang2; David Sprouster1; Jason Trelewicz1; Weicheng Zhong3; Ying Yang3; Steven Zinkle2; Lance Snead1; 1Stony Brook University; 2University of Tennessee, Knoxville; 3Oak Ridge National Laboratory
High-strength, high-thermal conductivity, and creep-resistant Cu-Cr-Nb-Zr alloys have attracted significant interests for the application in fusion energy system as the substrate materials for plasma-facing components. In this work, a Cu-Cr-Nb-Zr alloy was fabricated using direct current sintering (spark plasma sintering) from a gas atomized feedstock powder. The microstructure of sintered Cu-Cr-Nb-Zr alloy were characterized by synchrotron X-ray diffraction, small-angle X-ray scattering, scanning electron microscopy, and transmission electron microscopy. A high-thermal and electrical conductivity were achieved on the sintered CCNZ alloy together with a high hardness. A multi-modal precipitate distribution is achieved via our approach, with both coarse Cr and medium-sized Cr2Nb precipitates located at grain boundaries, and a high-density of nano-scale Cr precipitates evenly distributed throughout the Cu matrix. Based on the quantified microstructure evolution, we outline a possible precipitation scheme, which demonstrates the role of high-sintering pressure and recrystallization in facilitating grain boundary pinning precipitates.
10:20 AM
Evaluation of Creep Deformation of Ferritic/Martensitic (FM) Grade 91 Steel Fabricated Using Wire Arc Additive Manufacturing (WAAM): Mahmoud Hawary1; K. Murty1; 1North Carolina State University
Additive Manufacturing (AM) is a revolutionary technology and a transformative approach in the industry to fabricate complex structures from user-specific materials targeting specific properties and microstructure making it a competitive technology for the nuclear industry. Modified 9Cr-1Mo steel (Grade 91) is one of the main nuclear materials that have different applications in different types of reactors. In this work, the creep deformation of Grade-91 fabricated using Wire Arc Additive Manufacturing (WAAM) is evaluated. Constant load uniaxial creep tests were performed in air while monitoring the axial strains using a linear variable differential transformer (LVDT) extensometer. The tests have been conducted at different stress levels (90-200 MPa) and different temperatures (550 – 750 oC) to obtain the underlying creep mechanisms. Samples with different shielding gas compositions and tempering treatments have been considered in order to study the impact of various processing and post-processing techniques on the creep behavior.