Mechanical Behavior and Degradation of Advanced Nuclear Fuel and Structural Materials: Novel Nuclear Materials & Characterization I
Sponsored by: TMS Structural Materials Division, TMS: Mechanical Behavior of Materials Committee, TMS: Nuclear Materials Committee
Program Organizers: Dong Liu, University of Bristol; Peng Xu, Idaho National Laboratory; Simon Middleburgh, Bangor University; Christian Deck, General Atomics; Erofili Kardoulaki, Los Alamos National Laboratory; Robert Ritchie, University of California, Berkeley

Monday 2:00 PM
February 28, 2022
Room: 204A
Location: Anaheim Convention Center

Session Chair: Alex Leide, University of Bristol; Jie Lian, Rensselaer Polytechnic Institute


2:00 PM  Invited
Characterizing and Testing High Dose Neutron Irradiated Materials for Cladding Applications: Stuart Maloy1; Ben Eftink1; Tarik Saleh1; Mychailo Toloczko2; Dave Hoelzer3; T. S Byun3; 1Los Alamos National Laboratory; 2Pacific Northwest National Laboratory; 3Oak Ridge National Laboratory
    The Nuclear Technology R&D program has significant experience at qualifying metallic fuels for fast reactor applications. In this process, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. New ferritic/martensitic and ferritic Oxide Dispersion Strengthened (ODS) alloys have been developed and tested with improved radiation tolerance after high dose neutron exposures. Research includes developing ferritic/martensitic and ferritic ODS alloys in plate and tube form for future nuclear applications. Recent progress in high dose irradiated materials testing and materials development will be presented including recent plans for testing high dose irradiated materials.

2:30 PM  
Microstructure Characterization and Micro-mechanical Properties of 14YWT Tubing after Proton Irradiation: Cayla Harvey1; Osman El-Atwani2; Stuart Maloy2; Siddhartha Pathak3; 1University of Nevada, Reno; 2Los Alamos National Laboratory; 3Iowa State University
    14YWT nanostructured ferritic alloy (NFA) is a leading structural material candidate for fast nuclear reactors due to its exceptional irradiation tolerance and high strength. We examined This work examines alloys after different processing paths of hydrostatic extrusion and pilger processing with varying annealing temperatures. It is targeted for application as fuel cladding. We conducted electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM) based micro-structural characterization along with corresponding micro-mechanical testing on 14YWT alloys before and after proton irradiation. This work examines alloys after different processing paths of hydrostatic extrusion and pilger processing with varying annealing temperatures. Electron backscatter diffraction was used to identify the principle texture and microstructure components. Transmission electron microscopy was used to characterize the morphology, dislocation density, and oxide particles. Spherical indentation shows the compressive stress-strain response, while micro-tensile tests reveal the change in ductility and tensile strength after proton irradiation.

2:50 PM  
Behavior of High Entropy Alloy in Molten Salt Environments Under Biaxial Stresses: Wylie Simpson1; Xinyi Wang1; James Earthman1; 1University of California Irvine
    Structural performance in molten salt environments is important for materials used for novel energy generation such as thermal solar and next generation nuclear power systems. Characterization of these materials under these operative conditions is critical for understanding their ability to resist corrosion damage. The present research aims to characterize corrosion mechanisms and mechanical degradation of an oxide-oxide ceramic matrix composite, and high entropy alloy in NaSO4-NaCl and FLiNaK molten salts under biaxial loading conditions in an inert gas environment. Corrosion penetration depth will be measured as a function of time for these materials at 1023K unloaded as well as under biaxial tensile stress imposed using a novel disc bend configuration. The results will be compared with alloy 718 and the high-entropy alloy Al0.1CrCoFeNi under these conditions.

3:10 PM  
Ion Irradiation Effects on Microstructure Evolution and Mechanical Properties of Silicon Oxycarbide: Kathy Lu1; Sanjay Kumar Singh1; Kaustubh Bawane2; 1Virginia Polytechnic Institute and State University; 2Idaho National Laboratory
    Polymer derived ceramic is a promising nuclear material due to its irradiation resistant and high temperature stable nature. In this study, silicon oxycarbide (SiOC) was fabricated by pyrolysis of a polysiloxane precursor polymer at 1000C and 1500C in an Ar atmosphere. The 1000C pyrolyzed SiOC is fully amorphous while the 1500C sample involves crystalline β-SiC nanodomains, turbostratic carbon network, and amorphous SiOC. After 100 keV He ion irradiation, there is no detectable microstructural changes for the 1000C pyrolyzed SiOC. However, the 1500C pyrolyzed SiOC shows amorphization. Neither sample has He bubbles, elemental segregation or voids as a result of irradiation. Raman studies showed effect of irradiation on Si-O-C bonding. Irradiation temperature has no perceivable effect on microstructure and mechanical properties of SiOC. This work offers important guidance for development of radiation-tolerant polymer derived SiOC materials with applications in advanced nuclear reactors.

3:30 PM Break

3:50 PM  
Compositionally Graded Specimen: A High-throughput Approach for Nuclear Material Development : Jingfan Yang1; Laura Hawkins2; Miao Song3; Lingfeng He2; Zhijie Jiao3; Yongfeng Zhang4; Daniel Schwen2; Xiaoyuan Lou1; 1Auburn University; 2Idaho National Laboratory; 3University of Michigan; 4University of Wisconsin
    We demonstrate the compositionally graded specimen fabricated by laser additive manufacturing (AM) can accelerate the nuclear alloy synthesis and testing by 5-10 times. The study evaluates the effects of low-level refractory alloying elements on the proton-radiation induced damages and degradation in austenitic stainless steel, including swelling and loops, radiation induced segregation (RIS), radiation hardening, irradiation assisted stress corrosion cracking (IASCC). The results were benchmarked to the same materials produced by conventional forging. Under as-built condition, the materials produced by laser AM exhibited higher resistance to radiation damages and IASCC than wrought materials. Through post-AM thermomechanical treatment, recrystallization produced wrought-like equiaxed coarse grains across the graded specimen, which yielded similar radiation and IASCC resistance as its wrought counterparts. Although the demonstration was conducted with proton-irradiated specimens, we emphasize this high-throughput approach may be employed for neutron irradiation, and significantly reduces the cost of structural alloy development and qualification for nuclear applications.

4:10 PM  
Chemical Redistribution of Alloying Elements through Oxide/Metal Interface of Irradiated ZrNbFe Alloys and Its Implication on Corrosion Behavior: Zefeng Yu1; Elizabeth Kautz2; Hongliang Zhang1; Anton Schneider1; Yongfeng Zhang1; Sten Lambeets2; Arun Devaraj2; Adrien Couet1; 1University of Wisconsin-Madison; 2Pacific Northwest National Laboratory
    Transport of species through zirconium oxide is affected by oxide doping, resulting in complex corrosion mechanism. In this study, we focus on elemental redistributions across the oxide/metal interface. In-situ atom-probe-tomography experiments, where the needle is pre-oxidized in O2 at 260C, have been performed on unirradiated and 1 dpa proton irradiated ZrNbFe model alloys to characterize chemical redistribution across the oxide/metal interface. The results show that solute Nb redistributes across the oxide/metal interface, reducing the Nb/Zr ratio in the metal relative to the oxide, while the Fe does not redistribute. High-Resolution Transmission-Electron-Microscopy is performed to measure oxide thickness and determine its microstructure under similar corrosion conditions. Density functional theory calculations on solute thermodynamics are performed in metal and oxide matrices to tentatively explain, and compare to, the experimental observations. The implications in terms of oxide doping and transport of oxidizing species through the protective oxide under irradiation are discussed.

4:30 PM  
NOW ON-DEMAND ONLY - Deformation Behavior of Helium Irradiated Nano-pillars Containing a Helium Gas Bubble Superlattice: Andrew Scott1; Yujun Xie2; Peter Hosemann1; 1University of California Berkeley; 2Lawrence Berkeley National Laboratory
     The Helium Gas Bubble Superlattice (He-GBS), one example of defect self-ordering phenomenon encountered in structural materials in fission and proposed fusion systems, is a subject of ongoing interest in the nuclear materials community. While many studies have investigated the structure and formation of such systems, the resultant He-GBS induced changes of mechanical properties and deformation behavior of materials remain uncertain.In this study, nano pillars are milled from high purity single crystal Vanadium and Titanium using a focused ion beam (FIB) and irradiated using a Helium Ion Microscope (HIM). Mechanical properties are measured using in situ SEM compression testing, allowing for quantification and visualization of hardening and embrittlement in irradiated structures. Additional characterization is performed utilizing Transmission Electron Microscopy (TEM) to observe changes in the He-GBS structures before and after compression. We believe our results establish a more comprehensive understanding of the interactions between defect and host lattices during deformation.