Advanced Characterization of Materials for Nuclear, Radiation, and Extreme Environments: Poster Session
Sponsored by: TMS Nuclear Materials Committee
Program Organizers: Samuel Briggs, Oregon State University; Christopher Barr, Department Of Energy; Emily Aradi, University of Huddersfield; Michael Short, Massachusetts Institute of Technology; Janelle Wharry, Purdue University; Cheng Sun, Clemson University; Dong Liu, University of Oxford; Khalid Hattar, University of Tennessee Knoxville
Tuesday 4:45 PM
November 3, 2020
Room: Poster Hall
Location: MS&T Virtual
Effects of Post-Processing Variability on Radiation Response of Additively-Manufactured HT9: Pengyuan Xiu1; Niyanth Sridharan2; Kevin Field1; 1University Of Michigan; 2Lincoln Electric
In this study, the radiation response of an additively manufactured (AM) HT9 alloy using laser powder blown directed energy deposition (DED) was studied. Three HT9 variants were investigated, one directly after DED (as-built, or ASB) and two post-build heat-treated following DED (called ACO3 and FCRD). The unirradiated microstructures varied significantly with the ASB specimen being highly defective (≥1014/m2 estimated) while the ACO3 and FCRD alloys exhibited dislocation densities of 2.26×1014/m2 and 3.58×1014/m2 respectively. The variants were dual-ion irradiated to the damage level of 16.6 dpa and 4 He appm/dpa at 445°C to study the radiation response. Radiation induced cavity densities were 3.1×1022/m2, 7.0×1021/m2, and 4.9×1021/m2, with the size of 4.3±1.7nm, 6.4±6.7nm, and 4.3±2.5nm for ASB, ACO3 and FCRD respectively. Nanometric Ni/Si-rich clusters existed only in the ACO3 and FCRD variants with densities around 8×1021/m2 after irradiation. The implications of the AM-based sink strengths on the radiation tolerance will be discussed.
Investigation of Uranium Silicide Fuel Form Additions through Rietveld Refinement and Internal Standard P-XRD: Zachary Acosta1; Cole Moczygemba1; Elizabeth Sooby Wood1; 1The University of Texas at San Antonio
Uranium oxides have historically been utilized to power nuclear reactors due to their resistance to further oxidation and ease of large scale fabrication. However, other fuel forms may make more ideal candidates. U3Si2 has a higher thermal conductivity in addition to having a higher uranium density, making it more economical and less prone to cracking during thermal cycling. Interstitial additions may be introduced into the uranium silicide fuel form to boost lacking properties. Zirconium has been hypothesized as an ideal interstitial material due to its low neutron cross section and resistance to corrosion. Zirconium additions to U3Si2 were fabricated at 2%, 7%, and 14% by volume percent. These additions were characterized by powder-XRD, pre and post thermal solutionization, using the internal standard method and Rietveld refinement. The selected method allows for structural information to be derived about the atomic lattice, identification of secondary phases, and the quantification of interstitial mixing.