Materials in Nuclear Energy Systems (MiNES) 2021: Advanced and Novel Materials- Session III
Program Organizers: Todd Allen, University of Michigan; Clarissa Yablinsky, Los Alamos National Laboratory; Anne Campbell, Oak Ridge National Laboratory

Thursday 10:20 AM
November 11, 2021
Room: Urban
Location: Omni William Penn Hotel

Session Chair: Andrew Hoffman, Catalyst Science Solutions


10:20 AM  Invited
Opportunities for Advanced Concepts in Nuclear Fuel Development: Andrew Nelson1; 1Oak Ridge National Laboratory
    Advances in the science of nuclear materials have been largely centered on structural materials in recent decades. However, properties and performance of the fissile material (the fuel) ultimately dictate the power output, cycle length, safety margin, and beyond design basis behavior of reactors. Recent advances in manufacturing of ceramic materials, composite structures, incorporation of in situ fabrication diagnostics have the potential to impact how nuclear fuels are designed, fabricated, qualified, and accepted from a quality assurance standpoint. The challenge for nuclear fuels researchers in the coming decade will be efficiently developing new concepts within the broader understanding of nuclear fuel performance and material response to irradiation. This talk will present a range of contemporary efforts in this area including near term opportunities. Finally, the crucial role that accelerated irradiation testing methods including modeling and simulation will play in understanding irradiation performance of these fuel concepts will be summarized.

11:00 AM  
Metal Hydride Moderator Development at Los Alamos National Laboratory: Tarik Saleh1; Aditya Shivprasad1; Caitlin Taylor1; Thomas Nizolek1; Joshua White1; Erik Luther1; 1Los Alamos National Laboratory
    The use of moderators in advanced reactors can allow for less fuel and result in smaller reactors, which is particularly useful when there are space or weight constraints for the ultimate reactor application or location. Metal hydrides are strong neutron moderator candidates because they contain a high hydrogen density and are phase stable at much higher temperatures than water. This talk will discuss research at Los Alamos National Laboratory studying fabrication techniques and property measurements in a variety of metal hydride moderator materials, focusing on Zirconium and Yttrium Hydrides. Stoichiometry, phase, and fabrication methods and their effect on thermophysical and thermomechanical properties will be presented.

11:20 AM  
Radiation Tolerance of Capacitive Discharge Resistance Welded 14YWT: Calvin Lear1; Benjamin Eftink1; Hyosim Kim1; Matthew Schneider1; Todd Steckley1; Yongqiang Wang1; Thomas Lienert1; Stuart Maloy1; 1Los Alamos National Laboratory
    Advantages in radiation tolerance, creep resistance, and high temperature strength make oxide dispersion strengthened (ODS) ferritic steels promising materials for extreme conditions. Unfortunately, excess heat and localized melting from traditional fusion welding degrades the dispersed oxide particles responsible for these attributes – making ODS components difficult to join. Recent work with solid-state capacitive discharge resistance welding (CDRW) has produced 14YWT-14YWT joints (caps to thin-walled cladding tubes) without changing dispersoids or microstructure near the weld line. Material from these joints was subjected to self-ion (5.0 MeV Fe2+, 600 dpa, 450 °C) and proton (1.5 MeV H+, 0.7 dpa, 300 °C) irradiations, with pre- and post-irradiation microstructure characterized using electron microscopy and mechanically tested using nano-indentation. The accumulation of radiation-induced defects (dislocation loops, void swelling) and the stability of the post-CDRW microstructure (grain structure, dispersoids) were evaluated to ensure that the CDRW process does not degrade pre-existing resistance to radiation-induced microstructural evolution.

11:40 AM  Cancelled
In-situ Nanomechanical Characterization of Neutron-irradiated HT-9 Steel: Tanvi Ajantiwalay1; Assel Aitkaliyeva1; Megha Dubey2; Yaqiao Wu2; 1University of Florida; 2Boise State University
    For a safe operation of HT-9 steels as structural material in advanced reactors, a relationship between its irradiated microstructure and mechanical behavior needs to be established. In-situ nanocompression testing inside a transmission electron microscope (TEM) involves a combined analysis of defect microstructure and mechanical properties in real time. Under the application of a uniaxial compressive load, the electron-transparent nano-pillars reveal the morphology and movement of defects. In this study, nano-pillars of different dimensions were fabricated using focused ion beam (FIB) from a HT-9 specimen neutron-irradiated to 4.29 dpa at 469 ºC. All these pillars were compressed in a displacement-controlled mode using the PI-95 Picoindenter. The results obtained from the stress-strain curves show that the average yield stress varies with the cross-sectional area of the pillars. The in-situ TEM observation shows several dislocations burst events taking place after yielding as depicted by the sharp load drops in the curves.