Accelerated Discovery and Qualification of Nuclear Materials for Energy Applications: Poster Session
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
Tuesday 5:30 PM
March 16, 2021
Room: RM 48
Location: TMS2021 Virtual
Anisotropic Biaxial Creep Behavior of Textured Nb-modified Zircaloy Cladding: Mahmoud Hawary1; K. Murty1; 1North Carolina State University
ZIRLOŽ, a niobium alloyed Zr-4, was developed to enhance the deformation and degradation characteristics of the radioactive nuclear fuel cladding tubes in light water reactors. These thin-walled tubes are used following cold work and stress relief. In this work, the effects of Nb addition to Zry-4 on anisotropic biaxial creep characteristics in the as-received (CWSR) and recrystallized (Rx) conditions have been evaluated. The crystallographic textures were evaluated using EBSD from which crystallite orientation distribution functions (CODF) were generated for quantitative description of the creep anisotropy based on crystal slip models. Biaxial creep tests were performed using internally pressurized tubing superimposed with axial load under varied hoop to axial stress ratios of 0 to 2. The anisotropy parameters (R and P) in Hill’s formulation for generalized stress for anisotropic materials are evaluated. In addition, the creep locus is derived at a constant energy of dissipation that deviated from isotropy.
Defect Cluster Mobilities and Preferred Configurations in α-zirconium: A Comparison of Two Interatomic Potentials: Jose March-Rico1; Brian Wirth1; 1University of Tennessee Knoxville
The accuracy of molecular dynamics (MD) data and observations directly depends on the accuracy of the selected interatomic potential. Here we compare the performance of two interatomic potentials for hcp α-Zr: the M07 (2007) and BMD19 (2020) potentials. The M07 potential inaccurately predicts stable void formation as the preferred vacancy cluster configuration while the BMD19 potential accurately predicts the double-layered vacancy a-loop. Other key predictions of the BMD19 potential consistent with reported experimental or ab initio observations include: 1) a range of habit planes between type I and type II prism planes for a-loops, 2) the stable formation of a basal pyramidal vacancy cluster, 3) the faulting structure of basal c-loops, and 4) the thermodynamic preference for prismatic vacancy a-loops at large loop sizes. Due to the excellent agreement of the BMD19 potential predictions for defect cluster characteristics, its use is strongly recommended over the M07 interatomic potential.
Helium Effect on Cavity Swelling in Dual-ion Irradiated Fe and Fe-Cr Alloys: Yan-Ru Lin1; Arunodaya Bhattacharya2; Da Chen3; Ji-Jung Kai3; Jean Henry4; Steven Zinkle1; 1University of Tennessee; 2Oak Ridge National Laboratory; 3City University of Hong Kong; 4CEA
Fe-Cr ferritic-martensitic steels are promising structural materials for nuclear applications, in part due to cavity swelling resistance. However, transmutant helium enhances cavity nucleation and may cause unacceptable swelling for DT fusion reactors. We have performed multi-temperature (400 to 550°C) simultaneous dual-ion irradiations on ultra-high purity bcc Fe and Fe-Cr alloys (3 to 14 wt.% Cr). 8 MeV Ni3+ ions were selected to provide a relatively wide mid-range region with a dose ~30dpa. Energy-degraded 3.5MeV He2+ ions were simultaneously implanted at rates of 0.1appm/dpa and 10appm/dpa that are relevant to the irradiation conditions of fission and fusion reactors, respectively. Void swelling occurred at all temperatures in all alloys at both He implant conditions. Transmission electron microscopy (TEM) characterization found that the cavity swelling of 10appmHe/dpa irradiated samples was larger than the 0.1appmHe/dpa irradiated samples. In addition, the bimodal distribution of cavities was only observed in the materials irradiated with 10appm He/dpa.
Manufacturing Process Optimization of High-density LEU Targets for Mo-99 Production: Kinam Kim1; Tae Won Cho1; Sunghwan Kim1; Kyuhong Lee1; Yong Jin Jeong1; 1Korea Atomic Energy Research Institute
Mo-99 producers have been attempting to replace conventional highly enriched uranium (HEU) targets with low enriched uranium (LEU) targets by international non-proliferation policies. As a result, it is necessary to develop high-uranium-density targets with LEU to improve the Mo-99 production efficiency of LEU targets. Korea Atomic Energy Research Institute (KAERI) has been developing high-density LEU targets using atomized U-Al alloy powder. We successfully mass-produced uranium alloy powder for high-density targets through centrifugal atomization and fabricated high-density targets with a uranium density of 3.2 and 4.1 gU/cm3. The improvement in thermomechanical treatment (hot rolling + heat treatment) process was however needed because defects resulted from the meat expansion during phase transformation occurred on the surfaces of high density targets with 4.1 gU/cm3. In this work, we fabricated high density targets under various process conditions and analyzed them through destructive and non-destructive examinations in order to optimize the manufacturing process.
Mesoscale Modeling of the Effect of Interfaces on Segregation of Point Defects and Solutes and the Patterning of Extended Defects: Karim Ahmed1; Abdurrahman Ozturk1; Merve Gencturk1; Lin Shao1; 1Texas A&M University
We present a new approach for modeling the effect of interfaces on the segregation of point defects and solute atoms and the resultant pattering of voids and dislocations loops. The new approach couples the spatially-resolved rate-theory (SRRT) and phase-field (PF) modeling techniques. First the sink strength of discrete interfaces is calculated from the SRRT models as function of temperature, production bias, grain/particle size, and diffusivities. The values of the sink strengths are then used to derive the reaction rates between the individual interfaces and point defects in the PF models of microstructure evolution. Lastly, coupled PF and SRRT models are constructed to account for the concurrent microstructure evolution, solute segregation, and voids and loops pattering in Fe and Ni. The model predictions demonstrate the shortcoming of the classical homogenous rate-theory approach and the limitations of using ion irradiation to mimic neutron irradiation.
Modeling and Analysis of the Effects of the Microstructure on U-10Mo Fuel Thickness Variation during Hot Rolling: Lei Li1; Vineet Joshi1; Ayoub Soulami1; 1Battle Pacific Northwest National Lab
Low-enriched uranium metal alloyed with 10wt% molybdenum (U-10Mo) is being developed by the NNSA to replace high-enriched uranium fuel to minimize the risk of nuclear proliferation while still meeting the neutron flux requirements in the U.S. high-performance research reactors. During hot rolling, cast U-10Mo ingots are sandwiched between two zirconium (Zr) sheets and then loaded into a steel can. Experimental observations of the hot-rolled samples indicate undesirable U-10Mo and Zr thickness variations in the fuel foil product after hot rolling. In this work, a microstructure-based finite element model was developed to investigate the effects of as-cast microstructural attributes on thickness variation in fuel foil and the resultant Zr interlayer. Parametric study on the effect of U-10Mo grain size, porosity, carbides, and hot roll can materials on the fuel and Zr interlayer variation was conducted. The potential for this modeling to guide the U-10Mo fabrication process of fuel foils is discussed.