Composite Materials for Nuclear Applications II: Composite Metallic Systems
Sponsored by: TMS Structural Materials Division, TMS: Nuclear Materials Committee, TMS: Composite Materials Committee, TMS: Mechanical Behavior of Materials Committee, TMS: Advanced Characterization, Testing, and Simulation Committee
Program Organizers: Anne Campbell, Oak Ridge National Laboratory; Dong Liu, University of Oxford; Rick Ubic, Boise State University; Lauren Garrison, Commonwealth Fusion Systems; Peng Xu, Idaho National Laboratory; Johann Riesch, Max-Planck-Insitut Fuer Plasmaphysik
Tuesday 8:00 AM
March 21, 2023
Room: 24B
Location: SDCC
Session Chair: Johann Riesch, Max-Planck-Institut für Plasmaphysik; Anne Campbell, Oak Ridge National Laboratory
8:00 AM Invited
An Innovative Additive Manufacturing Route for Metal Matrix Composites for Nuclear Applications: Taegyu Lee1; Wonjong Jeong1; SeunhHyeok Chung1; Ho Jin Ryu1; 1KAIST
As metal additive manufacturing has been actively applied in many industries, much research has focused on the development of additive manufacturing for nuclear applications. However, the additive manufacturing of metal matrix composites for nuclear applications is still in the emerging stage because printable raw materials are not sufficient and the solidification of reinforcement-bearing metals during the additive manufacturing processes has not been fully established. In this study, the progress of additive manufacturing of various metal matrix composites by using a newly proposed surface modification and reinforcement transplantation technology is introduced. The manufactured metal matrix composites include Y2O3 dispersed alloys for radiation resistance, B4C dispersed Al composites for neutron shielding, and TiC dispersed Ni-base superalloy for high-temperature applications. Both the strengthening of the reinforcements and better solidification behavior improve the mechanical properties of the additive manufactured metal matrix composites, which is beneficial for structural and functional applications under extreme conditions.
8:30 AM
Effect of Interfacial Features on the Strengthening Behavior of B4C/Al Composites: Juyeon Han1; Hansol Son1; Yoonjung Won1; Kisub Cho1; Hyunjoo Choi1; 1Kookmin university
Aluminum matrix composites reinforced by B4C are of interest as radiation-shielding material. B4C has the advantages of a high melting point and hardness as well as neutron absorption and shielding properties from the nuclear reaction with 10B. Hence, aluminum matrix composites containing B4C exhibit high specific strength, hardness, and radiation-shielding ability. However, weak interfaces due to poor wettability between B4C and aluminum deteriorates the performance of B4C/Al composites.In this study, aluminum composites containing 1.5, 3, and 5 vol.% B4C were produced by hot rolling mechanically milled composite powder and heat treatment to partially decompose the B4C particles. This allowed B and C atoms to intercalate into and then distort the aluminum lattices. We shall discuss the effect of the interfacial features, varied by the heat treatment durations, on the strengthening behavior of B4C/Al composites.
8:50 AM
Development and Additive Manufacturing of ODS IN-718 Alloys for Nuclear Applications: Eda Aydogan1; Yesim Yalcin1; Bora Derin2; Bahattin Koc3; 1Middle East Technical University; 2Istanbul Technical University; 3Sabanci University
Oxide dispersion strengthened (ODS) ferritic and Ni-based alloys having sub-micron grain size with a high density of nano-oxides (NOs) (<10 nm) are considered to be good candidates for structural components in Generation IV nuclear systems. In this study, three different grades of ODS IN-718 alloys have been designed using CALPHAD-based thermochemical modeling approach and produced by Selective Laser Melting (SLM) method with various power and velocity values to determine the best process parameters. Besides, heat treatment studies have been conducted to maximize the NO formation. As a result, IN-718 alloys with ~1023 m-3 NO concentration have been obtained. Tensile tests demonstrate that the strength and ductility of these samples are much improved compared to the standard IN-718 samples, especially at elevated temperatures due to the fine and uniform microstructure.
9:10 AM
Ion Beam Synthesis of Nano-Oxides in FeCr: Towards an Understanding of Precipitation in Oxide Dispersion Strengthened Steels: Stephanie Jublot-Leclerc1; Martin Owusu-Mensah2; Aurélie Gentils1; 1Université Paris-Saclay, CNRS/IN2P3, IJCLab; 2Kwame Nkrumah University of Science and Technology
The properties of oxide dispersion strengthened steels, conventionally produced by powder milling, are highly dependent on the nature and size distribution of their constituting nano-oxide precipitates. A fine control of the processes of synthesis would enable the optimization of pertinent properties for use in various energy systems. This control however requires the knowledge of the mechanisms of precipitation, which are still a matter of debate. In the present study, nano-oxide precipitates were produced by implantation of Y, Ti and O ions in a Fe-10%Cr matrix subsequently thermally annealed. The results show that the oxides that precipitate are not necessarily favoured thermodynamically, but rather result from complex kinetics aspects related to the interaction between implanted elements and induced defects. In specific conditions, the formation of nanoprecipitates with precise characteristics similar to those in conventionally produced ODS steels is evidenced. Key elements in understanding the early stages of precipitation were obtained.
9:30 AM Break
9:50 AM
Characterization of the Effects of Intermediate Temperature Neutron Irradiation on Model Fe-Cr Alloys: Dhriti Bhattacharyya1; Alan Xu1; Takuya Yamamoto2; G. Robert Odette2; 1Australian Nuclear Science and Technology Organization; 2UCSB
Ferritic and ferritic-martensitic steels with high Cr and stable oxide precipitates (e.g. ODS steels) show improved resistance to radiation damage and creep. However, irradiation enhanced diffusion is known to cause precipitation of Cr in the form of α' clusters. Moreover, irradiation by neutrons causes the formation of various defects, e.g., loops and dislocations.Here, we present TEM studies on a series of Fe-Cr alloys (Fe-3%Cr - Fe-18%Cr), irradiated in the Advanced Test Reactor at INL. The samples were irradiated at a temperature of ~450°C to 6.7 dpa. The results show the variation in size and number density of the Cr rich clusters and displacement defects as a function of Cr content, and help in understanding the behaviour of these alloys under neutron irradiation in different conditions by providing data for comparison with corresponding results from a previous series of the same alloys irradiated at ~320°C to ~1.8 dpa.
10:10 AM
Effect of Copper Fiber in RAFM Steel Composite on Improving the Thermal Conductivity: Yong Hwan Cho1; Hyun Joon Yang1; Chang-Hoon Lee2; Woong-Ryeol Yu1; Heung Nam Han1; 1Seoul National Univ; 2Korea Institute of Material Science
Reduced activation ferritic/martensitic (RAFM) steel is one of the primary candidates for blanket modules in a nuclear fusion reactor. Increasing the thermal conductivity of the blanket material contributes to improve the efficiency of the fusion reactor. This study aimed to fabricate composite materials that have higher thermal conductivity than conventional RAFM steel by adding copper fibers in the steel. To compare the effect of fiber types on the thermal conductivity, copper powder and 3D woven copper mesh were used. The composite materials were fabricated by utilizing spark plasma sintering. To calculate the thermal conductivity of the specimens, their thermal diffusivity and heat capacity were measured by using laser flash apparatus and differential scanning calorimetry. The mechanical properties were also measured by miniature tensile test. 3D copper mesh can effectively improve the thermal conductivity of the composites, but the optimization between their thermal and mechanical properties should be conducted.
10:30 AM
Microstructure and Thermophysical Property Characterization of U-ZrHx Fuel Fabricated by Powder Metallurgy: Tyler Smith1; Caitlin Taylor1; Michael Hahn1; Erik Luther1; Thomas Nizolek1; Aditya Shivprasad1; 1Los Alamos National Laboratory
Uranium-zirconium hydride (U-ZrHx) fuel is under consideration for microreactor applications. This type of fuel combines both fuel and moderator into a single material, and has proven to be a useful fuel type for TRIGA reactors. However, the ability to retain hydrogen and mitigate swelling at elevated temperatures and burnup becomes essential for fuel success in these applications. LANL has developed a powder metallurgy method for fabricating high density U-ZrHx fuel pellets. As-fabricated pellets were characterized using x-ray diffraction, microscopy, and thermophysical property measurements. This talk will overview these results and specifically address the thermal expansion, thermal conductivity, thermal diffusivity, and hydrogen retention of this fuel type at elevated temperatures.