Ceramic Materials for Nuclear Energy Research and Applications: Radiation Effects and Mass Transport
Sponsored by: TMS Extraction and Processing Division, TMS Structural Materials Division, TMS Light Metals Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nuclear Materials Committee
Program Organizers: Xian-Ming Bai, Virginia Polytechnic Institute and State University; Yongfeng Zhang, University of Wisconsin; Larry Aagesen, Idaho National Laboratory; Vincenzo Rondinella, Jrc-Ec
Thursday 8:30 AM
March 18, 2021
Room: RM 51
Location: TMS2021 Virtual
Session Chair: Simon Pimblott, Idaho National Laboratory; Walter Luscher, Pacific Northwest National Laboratory
8:30 AM Invited
Irradiation Effects on Zirconium Alloy Oxides and Their Impacts on In-reactor Corrosion Rates: Adrien Couet1; Zefeng Yu1; Taeho Kim1; Hongliang Zhang1; Mukesh Bachhav2; Lingfeng He2; 1University of Wisconsin-Madison; 2Idaho National Laboratory
It is well-known that irradiation affects the in-reactor corrosion rates of zirconium alloy fuel cladding via irradiation effects in the substrate, the growing oxide and the water environment. In this study we will focus on the photon, neutron and ion irradiation effects in zirconium oxides. UV irradiation has been performed in-situ on zirconium alloys and the irradiated region shows a lower oxide thickness, and photo-electrochemically induced Fe oxide deposition. Proton and electron irradiation of zirconium oxides induces specific defects, likely F centers, characterized by Raman spectroscopy, which affect post-irradiation oxide growth. Finally, oxides grown in-reactor up to 4 cycles on ZIRLO®, M5® and X2® alloys have been characterized by STEM and APT to determine the effect of irradiation on the redistribution of alloying elements in the oxide. These observations are rationalized to propose an overall understanding of irradiation effects on zirconium oxide and their impacts on corrosion rate.
9:00 AM
Effect of UV and Gamma Irradiation on the Hydrothermal Corrosion of Ion-irradiated SiC: Arunkumar Seshadri1; Koroush Shirvan1; Taeho Kim2; Adrien Couet2; 1Massachusetts Institute of Technology; 2University of Wisconsin-Madison
SiC is an advanced nuclear material proposed as accident tolerant fuel claddings for light water reactors. To evaluate their viability as nuclear materials, understanding the kinetics of SiC corrosion under irradiation and high-temperature, high-pressure aqueous environment is critical. Previously, hydrothermal corrosion of the composites in a high temperature water had been investigated. Radiation effect on hydrothermal corrosion is complex as both radiolysis and neutron induced material defects contributes to the corrosion kinetics. In the present work, separate effect studies on Si ion-irradiated samples is performed under hydrothermal conditions. Further, effect of radiolysis is studied based on low temperature gamma irradiation and high temperature Ultraviolet-Ozone (UVO) irradiation on both non-irradiated and ion-irradiated samples. Surface morphology and chemistry studies accompanied by electrochemical potential measured revealed several interesting insights on the corrosion mechanism in SiC under different conditions. Both radiolysis and material defects significantly enhanced corrosion in SiC surfaces.
9:20 AM Invited
In-situ Measurement of Tritium Release from Lithium Aluminate Under Neutron Irradiation: Walter Luscher1; David Senor1; Gary Hoggard2; 1Pacific Northwest National Laboratory; 2Idaho National Laboratory
Gamma-lithium aluminate (γ-LiAlO2) pellets with engineered microstructures were irradiated in the Advanced Test Reactor between September 2016 and January 2019 for a total of 350 EFPD at 23 MWth. In-situ tritium measurements were obtained from capsules containing individual pellets. Each capsule was designed to provide both active temperature control and a sweep gas to carry any tritium released from the pellet to an ex-reactor measurement system. Results from these measurements were used to assess tritium release from γ-LiAlO2 as a function of burnup, burnup rate, and microstructure. Pellet microstructures were tailored to examine the effects of grain size (2-11 μm) and porosity (1-15%) on tritium release. In addition to these engineered microstructures, an alternate pellet design consisting of ~30μm γ-LiAlO2 granules dispersed in a zirconium matrix was also evaluated. An overview of the test train, capsule, and ex-reactor measurements systems will be provided in addition to the tritium release measurements.
9:50 AM
Influence of Dose Rate and Temperature on Mass Transport in Hematite: Kayla Yano1; Sandra Taylor1; Tiffany Kaspar1; Danny Edwards1; Daniel Schreiber1; 1Pacific Northwest National Laboratory
In this study, mass transport pathways are directly visualized in an irradiated model iron oxide using novel isotopic tracer and characterization techniques. In nuclear reactor environments, irradiation-induced defects and elevated temperatures interact to influence mass transport. These two forces can be difficult to separate but we attempt to do so here. Isotopic tracers (e.g. 57Fe and 18O) are incorporated via molecular beam epitaxy. Irradiation dose rate and annealing temperature are varied to elucidate their role on atomic and nanostructural evolution. Complementary scanning/transmission electron microscopy and 3D atom probe tomography are used to document the microstructural changes and the diffusion of isotopic tracers in bulk and defected crystal structures. Comparisons are made of the isotopic redistribution upon either thermal annealing or argon irradiation to reveal the role of non-equilibrium point defect populations on self-diffusion coefficients. This study provides fundamental insight into mass transport mechanisms impacting material performance and degradation.
10:10 AM
Radiation Tolerance of Nanoporous Gadolinium Titanate: Nathan Madden1; Matthew Janish2; James Valdez2; Blas Uberuaga2; Jessica Krogstad1; 1University of Illinois at Urbana-Champaign; 2Los Alamos National Laboratory
Defect sinks are critical for improving a material’s radiation tolerance, needed to meet the demands of future nuclear applications. However, it has been demonstrated previously that not all defect sinks are equally effective. To determine the effectiveness of different defect sinks, grain boundaries and pores (i.e. free surfaces) are tested directly. A gadolinium titanate sample was created with three distinct regions in the same TEM lamella, each containing only one of the primary sinks of interest. The TEM lamella was irradiated using 1 MeV Kr in-situ irradiation within the TEM. Nanobeam electron diffraction was used after specific intervals of fluence to assess the crystallinity of each region. The results at room temperature and 600 C show that a free surface defect sink is more effective than grain boundaries at capturing radiation-induced defects. These results allow for a more informed microstructural design to meet the needs of future nuclear applications.
10:30 AM
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces: Simon Pimblott1; Jay LaVerne2; 1Idaho National Laboratory; 2University of Notre Dame
Carbides have great potential for application in the nuclear industry; however, various properties require detailed understanding for the materials to be properly utilized in situations where they will be exposed to extreme temperatures and mixed radiation fields. The behavior of SiC and ZrC as well as water slurries have been systematically studied using gamma and alpha radiation. Damage to the carbides was determined using analytical techniques including TGA, ICP-OES and SEM-EDS. Gamma irradiation of ZrC in air results in a change in Zr:C ratio with SEM examination showing oxidation of the surface. SiC is relatively stable under gamma irradiation except for some conversion of β to the α phase: α-ray radiolysis leads to the formation of SiO2. Aqueous slurry irradiations yield a large increase in the radiolytic yield of hydrogen compared to water, an effect that could have significant deleterious consequences if the material is deployed in nuclear energy environs.
10:50 AM
Molecular Dynamics Investigations of AlN-based Piezoelectric Ceramics under Irradiation: Michael Kempner1; Jesse Sestito1; Eva Zarkadoula2; Yan Wang1; 1Georgia Institute of Technology; 2Oak Ridge National Laboratory
Material degradation due to irradiation is a challenge in nuclear applications, including structural materials and materials used as transducers for sensing applications. Therefore, materials for such applications require good radiation response, while maintaining their piezoresponse. Aluminum nitride (AlN) has good piezoelectric properties which make it useful in sensor and actuator applications. Recently, scandium (Sc) doping was shown to increase the piezoelectric coefficient of AlN. However, the piezoelectric properties of Sc-doped aluminum nitride (Al1-xScxN) under irradiation have yet to be investigated at an atomistic level through molecular dynamics (MD) simulations. In this research, an Al1-xScxN force field is generated by including Sc in an existing AlN force field. The new force field is calibrated based on both the modulus of elasticity and piezoelectric coefficient by using a multi-objective Bayesian optimization approach. The defect production and the effects of irradiation on the piezoresponse of AlN and Al1-xScxN are examined using MD simulations.
11:10 AM
Irradiation Damage in High-entropy Carbide Ceramics: Fei Wang1; Xueliang Yan1; Tianyao Wang2; Yaqiao Wu3; Lin Shao2; Michael Nastasi2; Yongfeng Lu1; Bai Cui1; 1University of Nebraska-Lincoln; 2Texas A&M University; 3Boise State University
Novel high-entropy carbide ceramics (HECC) have been developed as a promising structural material for advanced reactor designs. (Zr0.25Ta0.25Nb0.25Ti0.25)C HECC with a single-phase rock-salt structure was synthesized by spark plasma sintering, which was irradiated by 3 MeV Zr ions to 20 dpa at 25, 300, and 500 ºC. X-ray diffraction analysis showed that (Zr0.25Ta0.25Nb0.25Ti0.25)C maintained a high phase stability without phase transformation after irradiation. About 0.2% lattice parameter expansion was revealed. The irradiation-induced microstructures were comprised of defect clusters with diameters of several nanometers, without void formation or radiation-induced segregation. The defect clusters were characterized by transmission electron microscopy as two types of dislocation loops, including perfect loops and faulted Frank loops. The growth of dislocation loops may be suppressed by the strong local lattice distortion. Nanoindentation tests showed irradiation-induced hardness increase, which was possibly caused by dislocation loops and lattice strain.