Materials and Fuels for the Current and Advanced Nuclear Reactors VI: Poster Session
Sponsored by: TMS Structural Materials Division, TMS: Corrosion and Environmental Effects Committee, TMS: Nuclear Materials Committee, TMS: Mechanical Behavior of Materials Committee
Program Organizers: Ramprashad Prabhakaran, Pacific Northwest National Laboratory; Dennis Keiser, Idaho National Laboratory; Raul Rebak, GE Global Research; Clarissa Yablinsky, Los Alamos National Laboratory

Monday 6:00 PM
February 27, 2017
Room: Hall B1
Location: San Diego Convention Ctr

G-1: Effects of Added Molybdenum on Corrosion of 3l6L Stainless Steel: Tahsin Rahman1; J. E. Indacochea1; W. L. Ebert2; 1University of Illinois at Chicago; 2Argonne National Laboratory
    Molybdenum is a rnajor constituent of metallic fuel wastes to be immobilized with steel cladding in an alloy waste form. The effects of added molybdenum (5-20 wt%) on the corrosion behavior of 316l-based alloys in an aggressive acid-brine environment (pH 4) were measured using potentiodynamic and potentiostatic methods. Different amounts of two Mo-rich intermetallic phases formed as Mo was added. Sigma phases having modest molybdenum contents (21-29 wt% Mo) formed in all the alloys and Laves phases with higher molybdenum contents (33-40 wt% Mo) formed in alloys made with more than 5 wt% Mo. Alloys with 15 wt% Mo or more showed poor corrosion resistance that is attributed to the abundant Laves phases. Detailed analyses of the corroded materials using SEM/EDS showed preferential corrosion of the Sigma and Laves phases (probably galvanic conosion). Test results and their relevance to formulation and performance modeling of waste forms will be discussed.

G-2: Fabrication and Microstructures of Burnable Absorber-cored Oxide Pellets for Advanced Nuclear Fuel: Qusai Mistarihi1; Yong Kim1; Ho Ryu1; 1Korea Advanced Institute of Science and Technology
    Gd2O3 is used as a burnable absorber for UO2 fuel. As more effective reactivity control can be achieved by lumping Gd2O3 into UO2, Gd2O3-cored oxides pellets were fabricated using 8 mol. % Yttria Stabilized Zirconia (8YSZ) as a surrogate for UO2. The fuel pellets were compacted by using uni-axial pressing and Cold Isotactic Pressing (CIP) followed by microwave sintering at temperatures up to 1600oC. The microstructures of fuel pellets were observed to characterize densification and interdiffusion of the composite oxide pellets.

G-3: Diffusion Studies in the Development of an FCCI Barrier for High-Burnup Metallic Nuclear Fuel: Daniel Eichel1; James Vollmer1; 1TerraPower, LLC
    TerraPower, LLC, is developing a high-burnup metallic fuel design for use in its Travelling Wave Reactor design. Many innovative features have been applied to extend the burnup capability of the fuel, including the use of barriers between the fuel and HT9 steel cladding to mitigate fuel/cladding chemical interaction (FCCI). Different barrier permutations, of various composition and morphology, have been tested in diffusion couple experiments, and the resulting diffusion interfaces were characterized using scanning electron microscopy and energy dispersive spectroscopy. Diffusion couple experiments included control samples in which HT9 steel was placed directly in contact with neodymium, an aggressive, representative surrogate fission product. To varying degrees, FCCI barriers were shown to greatly reduce the degree to which the cladding was attacked chemically.

G-4: Irradiated Materials Characterization Laboratory at Idaho National Laboratory: Lingfeng He1; Brandon Miller1; Dean Blanton1; Karen Wright1; Assel Aitkaliyeva1; Jian Gan1; 1Idaho National Laboratory
    Irradiated Materials Characterization Laboratory (IMCL) is the newest nuclear energy research facility at Idaho National Laboratory’s Materials and Fuels Complex. IMCL incorporates many features designed to allow researchers to safely and efficiently prepare irradiated fuel and material samples for microstructural–level investigations. The key current capabilities include a shielded hot cell, glovebox and hood for radioactive sample preparation, a shielded Electron Probe MicroAnalyzer (EPMA), Ga and Xe Dual Beam Focused Ion Beam (FIB) systems, and a Titan Themis 200 Transmission Electron Microscope (TEM) with Super-X Energy Dispersive Spectroscopy (EDS) system. Recent activities of irradiated fuel and materials studies at IMCL using advanced microscopies, including FIB, EPMA and TEM are highlighted.

G-5: Quantification of the Stress-Stabilization of Tetragonal ZrO2: Mitra Taheri1; Wayne Harlow1; 1Drexel University
    Zirconium-based alloys are commonly used for nuclear fuel rod cladding material because of their oxidation resistance and low neutron cross section. The oxidation process has been well studied, however questions remain as to how the tetragonal ZrO2 phase is stabilized. Since it is widely thought that stress may be the stabilizing factor for ZrO2 in this system, it is important to ascertain exactly how much stress present in the sample may affect the percentage of tetragonal ZrO2 present an oxidized zircaloy material. Here we present a fundamental study of stabilization and in particular, an attempt to pinpoint the onset of relaxation resulting from reduction of stress in the sample. Using a combination of precession electron diffraction-based strain measurements and electron energy loss spectroscopy, the fraction of tetragonal ZrO2 present as a function of sample thickness was quantified. The results provide additional insight into the zircaloy cladding oxidation process.

G-6: Steam Oxidation Resistance of Silicide and Aluminide-coated Refractory Metals: Woojin Lim1; Hyun Gil Kim2; Hojin Ryu1; 1Korea Advanced Institute of Science and Technology; 2Korea Atomic Energy Research Institute
    The high melting points and excellent high-temperature strengths are considered advantages of refractory metals as candidate accident tolerant fuel claddings. In order to enhance the oxidation resistance of refractory metals at an elevated temperature, protective coatings have been applied to Mo and Nb metallic substrates by pack cementation. Silicide and aluminide layers with various thicknesses were coated on Mo and Nb samples with the annealing time for chemical reactions. When the bare and coated specimens were exposed to an elevated temperature up to 1200oC to analyze the steam oxidation resistance, silicide and aluminide coated specimens showed quite better oxidation resistance performance with the small amount of weight change when compared to bare specimen oxidation results. The comparison of the oxidation behavior of the silicides and aluminides of Mo and Nb were performed by using microstructural analyses.

G-7: Advanced Electron Microscopy of Fission Products in Irradiated TRISO Fuel: Rachel Seibert1; Chad Parish2; Kurt Terrani2; Jeff Terry1; 1Illinois Institute of Technology; 2Oak Ridge National Laboratory
    We have examined the SiC containment layer of fissioned TRISO nuclear fuel particles from the AGR-1 program using advanced microscopy techniques. Post-irradiation, some of the TRISO particles were annealed at higher temperatures to simulate hypothesized accident scenarios. TEM/STEM samples were prepared at the inner, middle, and outer regions of the SiC particle fragments. SiC microstructure and the migration behavior of fission products were studied using TEM/STEM and energy dispersive X-ray spectroscopy (EDS) for elemental mapping and identification. Pd and U have been found dispersed throughout the SiC matrix. A large inventory of fission products is observed at the IPyc interface, including Ag and Eu. The results have provided information on the fission product transport behavior through irradiated SiC at varying temperatures. Experimental data such as this is necessary to constrain the current models used to predict the long-term behavior of fission products in TRISO fuels for next generation nuclear reactors.

G-8: Phase Field Modeling of Fission Gas Behavior in Metallic Nuclear Fuel: San-Qiang Shi1; Pengchuang Liu1; Xin Wang2; Pengcheng Zhang2; 1The Hong Kong Polytechnic University; 2Institute of Materials
    As metallic nuclear fuels, U-Mo and U-Zr have advantages in irradiation stability, better mechanical strength, corrosion resistance and neutronic economy. However, the formation and evolution of fission gases in irradiated fuels will bring severe impacts on their thermo-mechanical properties and then affect the fuel performance. Microscale and mesoscale theoretical investigations are urgent needed for deeper understanding of the influences brought by fission gases on these fuels. Phase field modeling method is used in this study. This method does not need to track individual atom or individual interfaces at the gas bubbles, and it predicts the microstructural evolution by solving Cahn-Hilliard and Allen-Cahn kinetic equations. In this work, we study the formation and evolution of fission gas bubbles with the effect of elastic interaction, and investigate how morphological changes of gas bubbles can influence on the thermal conductivity of nuclear fuels. The theoretical predictions will be compared to experimental results.

G-9: Asymptotic Expansion Homogenization of Thermal Conductivity and Elasticity of Irradiated Hafnium-Aluminum Composite Performed on Reconstructed and Synthetic Microstructures: William Harris1; Donna Guillen2; 1North Carolina State University; 2Idaho National Laboratory
    A thermal neutron-absorbing metal matrix composite (Hf-Al), comprised of HfAl3 intermetallic particles in an Al matrix, has been fabricated with 20.0, 28.4, 36.5, and 100 vol% HfAl3 and irradiated for up to 4 dpa in a light water reactor. DSC and LFA were performed before and after high-temperature annealing and the samples’ bulk thermal conductivities were calculated as functions of temperature, HfAl3 volume fraction, extent of irradiation, and state of annealment. SEM, EBSD, and EDS are performed to assess the effects of irradiation on the material’s microstructure, and a solid texture synthesis algorithm is used to generate 3D synthetic microstructures for the pre- and post-irradiation 28.4 vol% HfAl3 samples. Finite element simulations are performed on reconstructed and synthetic volumes to assess the applicability of the asymptotic expansion homogenization (AEH) method in predicting the bulk thermal conductivity and elasticity of the unirradiated and irradiated composite.

G-11: A Composite Waste Form for Electrochemical Processing Wastes: Xin Chen1; J. Ernesto Indacochea1; William Ebert2; 1University of Illinois at Chicago; 2Argonne National Laboratory
    Metallic/ceramic composite waste forms are being developed for combined metallic and oxide waste streams generated during electrochemical refining of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. The Nd2O3 and some of the uranium reacted with Zr to form zirconates and the remaining uranium was reduced and incorporated in Fe-Zr-U intermetallics. Two Fe-Cr-Mo intermetallics also formed, which are expected to host Tc. The results of microstructure characterizations of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors will be presented. Test results suggest composite waste forms will provide flexibility for immobilizing complex waste streams by accommodating both metallic and oxidized waste streams in durable host phases while lowering waste form production and disposal costs.

G-12: A Proposed Mechanism of Corrosion of Nickel by Tellurium in Molten Salt: Natasha Skowronski1; 1MIT
    Generation VI liquid-fueled molten salt reactors (MSRs) require workable, corrosion-resistant materials to contain fuel salts. A notable corrosive agent in operating MSRs is the fission product tellurium (Te), known to Oak Ridge National Laboratory (ORNL) to have led to intergranular stress corrosion cracking (IGSCC) of the vessel during the Molten Salt Reactor Experiment (MSRE) in the 1970s. Though a renewed interest in MSRs has led to some increased knowledge of how to reduce intergranular corrosion by Te through alloy selection and redox potential control, the specific mechanism of intergranular corrosion by Te during reactor operation remains poorly understood. To determine a possible mechanism of corrosion, commercially pure nickel-201 was held submerged in FLiNaK salt containing dissolved Te for 100 hours, 250 hours, and 500 hours. Resulting images taken with a scanning electron microscope (SEM) and a corresponding proposed mechanism of corrosion are presented.

G-14: Fission Gas Release and Swelling Model of Uranium Nitride Based on the Rate Ttheory: Jing Liu1; Yedong Gao1; Yang Du1; Bo Zhang1; Di Yun1; 1Xi’an Jiaotong University
    A mechanistic model for fission gas release and fuel swelling of uranium nitride has been developed based on the classical rate theory. The model considered intragranular and intergranular gas behaviors and described gas atoms and gas bubbles diffusion, gas bubbles nucleation, resolution of intragranular gas bubbles and growth and coalescence of gas bubbles on the grain boundary. The model calculated fission gas release and fuel swelling with respect to burnup and the results were compared with published experimental data to determine important model parameters. These parameters were then compared to those obtained in the literatures. The sensitivities of key model parameters on uranium nitride fuel fission gas release and fuel swelling have been obtained through a parametric study. The feasibility of applying rate theory mechanistic model to simulate uranium nitride fuel fission gas release and fuel swelling behaviors has been verified and the validity of the model has been demonstrated.

G-16: Simulation of Constituent Redistribution and Fuel Restructuring in MOX Fuel: Yang Du1; 1Xi'an Jiaotong University
    Fuel performance is one of the most important issues in fuel behavior analysis, constituent redistribution and fuel restructuring are critical to the prediction of fuel behavior. These behaviors can be simulated by solving self-defined equations with COMSOL, a finite-element based multi-physics platform. In our work, redistribution phenomena such as O, Pu and Cs transportation are considered. Restructuring phenomena such as porosity migration, formation of central hole, formation of regions with different grain types, growth of equiaxed grains, fuel densification and cracking are also included. In addition, we took into account the growth of Joint Oxide Gain over time and its effects on heat transfer. The results of simulation can describe the above mentioned phenomena reasonably well. This work demonstrated the interplay mechanisms between heat transfer, fuel restructuring, chemical constituent redistribution and solid mechanics on a finite-element platform and established a basis for future comprehensive fuel performance analysis of MOX fuel.

G-17: Thermodynamic Properties of Strontium-Bismuth Alloys for Electrochemical Separation of Strontium: Nathan Smith1; Timothy Lichtenstein1; Jarrod Gesualdi1; Hojong Kim1; 1Pennsylvania State University
    Thermodynamic properties of Sr-Bi alloys were determined by electromotive force (emf) measurements to evaluate the viability of liquid metals as media for separating alkali/alkaline-earth fission products (e.g. Sr, Ba, and Cs) from molten salt electrolytes in the pyrochemical processes for used nuclear fuel reprocessing. An electrochemical cell of Sr(s)|CaF2-SrF2|Sr(in Bi) was used to measure emf values of Sr-Bi alloys (0.05 < xSr(in Bi) < 0.80) at 475-725C. Activity values of Sr in Bi are reported (e.g. aSr(in Bi) = 3.6 10-12 for xSr = 0.05 at 625C) as are the partial molar entropy and enthalpy for each Sr-Bi alloy. The Sr-Bi system was also characterized by differential scanning calorimetry and x-ray diffraction to elucidate their phase behavior. Based on the strong chemical interactions between Sr and Bi, bismuth metal shows promise as a liquid electrode for electrochemical separation of alkali/alkaline-earth fission products from molten salt electrolytes.