Late News Poster Session: Nuclear Materials
Program Organizers: TMS Administration
Tuesday 5:30 PM
March 1, 2022
Room: Exhibit Hall C
Location: Anaheim Convention Center
N-39: Assessing Void Swelling Resistance of Internal Reactor Materials with an Ultrarapid Photoacoustic Technique: Nouf Almousa1; Benjamin Dacus1; Kevin Woller1; Ji Ho Shin2; Changheui Jang3; Michael Short1; 1Massachusetts Institute of Technology; 2Korea Advanced Institute for Science and Technology ; 3Korea Advanced Institute for Science and Technology
A non-contact photoacoustic technique in the form of transient grating spectroscopy (TGS) is applied to assess relative void swelling resistance of multiple structural materials. Statistically significant changes in the frequency of probed surface acoustic waves (SAWs) suggest that newly developed alloys containing uniformly distributed nanosized NbC precipitates show higher resistance to irradiation-induced void swelling when compared to their simpler, commercial analogues. The higher reduction in SAW frequency seen in the simpler steels, proportional to porosity, indicates more void formation which is shown by TEM imaging. This finding illustrates the minimum set of targeted TGS studies required to rapidly rank materials by relative void swelling resistance, and hence, accelerate irradiation defects studies.
N-41: Effect of the Surface Roughness of Tungsten on the Sputtering Yield Under Helium Irradiation: A Molecular Dynamics Study: Hyeonho Kim1; Kunok Chang1; 1Kyung Hee University
The divertor is one of the armor materials in the nuclear fusion reactor that keeps plasma purity and temperature. Sputtering is a phenomenon in which components of the divertor are bounced off when helium ions are irradiated to the divertor. Research has been conducted to reduce sputtering because sputtering reduces the purity and temperature of the plasma. Previous experiment studies have been regulated sputtering yields by controlling the roughness of the divertor surface. In this study, we used molecular dynamics to simulate the sputtering of tungsten-based divertors. We quantitatively evaluated the effect of surface roughness on sputtering phenomena, compared the simulation results, and discussed differences from experimental results.
N-42: Fabrication and Phase Analysis of U3Si2 and UB2 Fuel Composites with Al and Al2O3 Strengthening Additions: Geronimo Robles1; Joshua White2; Elizabeth Sooby1; 1University of Texas at San Antonio; 2Los Alamos National Laboratory
High uranium density fuels (HDF) composites candidates such as U3Si2 and UB2 have been investigated as drop-in replacement of UO2 in LWR’s for their superior uranium density and thermal conductivity. These properties benefit fuel structural stability, economy, and safety. But the energetic pulverization of U3Si2 in high temperature water bearing atmospheres and the low radiation tolerance of UB2 have slowed their advancement. However, as a composite, the oxidation onset temperature of U3Si2 has been delayed significantly with the addition of UB2 at concentrations <10wt%.The work presented investigates the microstructure and phase distribution of U3Si2-UB2 HDF composites with Al and Al2O3 additions. Al is selected for properties shown to improve oxidation resistance and Al2O3 is employed as an ODS component improving toughness. Fabricability challenges for each composition is then compared and evaluated. Each composition is characterized via SEM/EDS and p-XRD establishing a baseline for microstructural evolution in future oxidation testing.
N-43: Nanoindentation Creep Testing on Austenitic Alloys: Tianyi Chen1; 1Oregon State University
The evaluation of creep in nuclear reactors is a major concern for materials qualification and reactor safety during operation. Conventional creep testing is time consuming and expensive. Naonindentation creep testing provides a rapid-turnaround and low-cost screening tool to probe the creep properties. In addition, aanoindentation techniques allow mechanical testing on ion-irradiated samples and smaller neutron-irradiated samples. We found that nanoindentation creep testing is highly-sensitive to radiation damage by probing the changes in dislocation interactions with the damage features. Using data from neutron-irradiated austenitic alloys, we will showcase the correlations between deformation microstructures and the measured nanoindentation creep parameter changes. A new data-analysis method will also be introduced to explain the significant variance in nanoindentation creep exponents between different indents when the data is processed by the existing data-analysis method. This work advances the understandings of deformation mechanisms at small scales to close the gap between different testing length scales.
N-44: Oxidation and Microstructural Characterization of Tristructural Isotropic Particles (TRISO) in High Temperature Mixed Gas Atmospheres: Katherine Montoya1; Brian Brigham1; Tyler Gerczak2; Elizabeth Sooby2; 1University of Texas at San Antonio; 2Oak Ridge National Laboratory
TRISO particles are composed of 4 layers surrounding a fuel kernel suspended in graphite matrix material to form a fuel compact for high temperature gas cooled reactors. Off-normal reactor conditions can introduce steam to the fuel causing matrix oxidation with volatile gas production (CO and H2). The degraded matrix would expose the SiC layer in the TRISO to mixed gas atmosphere. Damage to this layer can cause particle failure and fission product release. To investigate potential failure modes of SiC in a complex atmosphere, oxidation testing of surrogate SiC exposed TRISO particles was conducted with a mixture of steam (<0.2 atm H2O), CO, and H2 (<0.02 atm) at high temperatures (1300ºC <T<1500ºC) in thermogravimetric analyzer. This study provides a mapping of microstructures associated with SiC oxidation regimes under matrix oxidation. Advanced characterization techniques include x-ray diffraction, focused ion beam and scanning electron microscopy to characterize the oxidized SiO2-SiC interface.
N-46: ZrO2 Corrosion Layers and Their Grain Boundary Networks: Aaron Chote1; 1Imperial College, London
Zircaloy-4 is the principal alloy employed as the cladding material in the UK’s only pressurised water reactor (PWR). Lithium is an additive in the coolant water of PWRs but the effect of lithium on the microstructural evolution and texture of the growing ZrO2 layer is not well-understood. Lithiated and non-lithiated ZrO2 layers from autoclaved Zircaloy 4 samples (oxidised for 105 days at 573K, 140MPa in 70-ppm lithiated water) were probed using precession electron diffraction, providing insight into the grain boundary plane distribution and grain boundary character distribution. Atom probe tomography was used to understand lithium’s segregation behaviour in ZrO2 layers. A change in thickness of the ZrO2 layer related to the underlying Zircaloy-4 grain orientation was observed, along with the co-segregation of lithium and iron to grain boundaries. The effect that lithium has on the prominence of certain grain boundary planes and the texture evolution in ZrO2 layers is discussed.