Ceramic Materials for Nuclear Energy Research and Applications: Microstructural Evolution under Irradiation in Oxide Ceramics
Sponsored by: TMS Structural Materials Division, TMS: Advanced Characterization, Testing, and Simulation Committee, TMS: Energy Committee, TMS: Nuclear Materials Committee
Program Organizers: Xian-Ming Bai, Virginia Tech; Yongfeng Zhang, Idaho National Laboratory; Maria Okuniewski, Purdue University; Donna Guillen, Idaho National Laboratory; Marat Khafizov, Ohio State University; Thierry Wiss, European Commission- JRC -Institute of Transuranium Elements – Germany
Monday 8:30 AM
February 27, 2017
Location: Marriott Marquis Hotel
Session Chair: Yongfeng Zhang, Idaho National Laboratory; Thierry Wiss, EC - JRC - Institute for Transuranium Elements
8:30 AM Invited
Ceramic Materials for Nuclear Energy Research and Applications: Kurt Sickafus1; 1University of Tennessee
With the exception of uranium dioxide (UO2), uses of ceramics in nuclear energy applications have (to date) been extremely limited. Nevertheless, the demonstrated success of UO2 as a robust nuclear fuel is helping to foster research into new applications for ceramics as nuclear materials. Most notable among candidate nuclear ceramics is silicon carbide (SiC), a refractory ceramic possessing high oxidation resistance and thermal conductivity. It is less obvious what other ceramic materials will be implemented in nuclear energy applications in the near-term. Most probably, oxide ceramic compounds will be used to immobilize long-lived radionuclides such as surplus actinides and other by-products of nuclear fission. In this presentation, we will review opportunities for oxide ceramics as advanced nuclear materials. We will consider research on both monolithic and composite ceramic systems. We will emphasize discoveries that have promoted the efficacy of oxide ceramics as potentially robust nuclear fuel and waste forms.
Alpha-damage Formation in Mixed Americium-uranium Compounds: Thierry Wiss1; Oliver Dieste1; Rudy Konings1; Ondrej Benes1; Jean-Yves Colle1; Joaquina Zappey1; Florent Lebreton2; Thibaud Delahaye2; Enrica Epifano2; Philippe Martin2; Christine Guéneau2; Damien Prieur1; Joe Somers1; 1European Commission; 2CEA
Americium contributes extensively to the radiotoxicity of spent fuel during the long term disposal of spent fuel. In a closed fuel cycle, however, americium can be separated and transmuted in fast neutron reactors, thereby limiting its long term radiological impact. A first such transmutation experiment took place in the mid-eighties in the SUPERFACT irradiation programme, where UO2 based-fuels with different content of americium (and other actinides) were irradiated in the Phénix fast reactor. Some of the irradiated SUPERFACT fuels have been investigated recently by transmission electron microscopy (TEM) showing a rather remarkable behaviour against of radiation damage. Archive materials from the SUPERFACT fuels have also been investigated as well as material with higher amount of americium (up to 50 at%). Electron energy loss spectroscopy (EELS), helium desorption spectroscopy and EXAFS was performed with the aim of determining the evolution of the structural damage and helium behaviour in these materials.
Microstructural Characterization of the Processes, Stability, and End-of-Range Effects in Heavily Irradiated Pyrochlores: Terry Holesinger1; James Valdez1; Cortney Kreller1; Yongqiang Wang1; Blas Uberuaga1; 1Los Alamos National Laboratory
Pyrochlores have been investigated for a number of technological applications including use for nuclear waste forms. The level to which these materials can tolerate damage accumulation is often probed with ion irradiation. In this talk, we will discuss heavy-ion irradiated (Kr+, 400 keV, 5.36E16 cm-2, room temperature, SRIM est. peak damage 129 dpa) of Gd2Ti2O7 (GTO) and Gd2Zr2Oy (GZO) pyrochlore bulk pellets and single-crystal films. GTO readily amorphizes while GZO converts to the fluorite structure. Conventional (S)TEM before and after irradiation was used to track the structural changes in each material, including the end-of-range region. In situ (S)TEM was used to examine the evolution of the as-irradiated microstructures at elevated temperatures. This combination of post mortem and in situ approaches was used to identify and understand the differences in the irradiation processes of these two pyrochlores that governed damage accumulation, defect stability/kinetics, and overall microstructural changes.
Probing Oxygen Defects in Ion Irradiated Actinide and Analogue Oxides Using Neutron Total Scattering: Raul Palomares1; Jacob Shamblin1; Cameron Tracy2; Christina Trautmann3; Maik Lang1; 1The University of Tennessee; 2Stanford University; 3GSI Helmholtzzentrum für Schwerionenforschung
The radiation response of CeO2, ThO2, and UO2 has been extensively studied due to the potential use of these materials for energy-related applications. Experimental characterization of irradiation-induced defects in these materials is typically performed using transmission electron microscopy (TEM) and conventional X-ray techniques. Although valuable, these methods do not yield bulk-averaged measurements (TEM) or sufficient detail about the oxygen sublattice (X-rays) where disorder primarily occurs. In this presentation, we demonstrate the ability of neutron total scattering measurements to elucidate oxygen defect morphology, and defect accumulation and annihilation mechanisms in swift heavy ion irradiated CeO2 and ThO2. The samples were irradiated with 2.2 GeV Au ions and measured at the Spallation Neutron Source as a function of fluence and isochronal annealing temperature. Pair distribution function analysis reveals that irradiation triggers the formation of peroxide-like defects, which actively participate in ionic conductivity above 200ºC. Preliminary research on UO2+x characterization is discussed.
10:00 AM Break
10:20 AM Invited
High Burn-up Nuclear Fuel, Impact of Fission Gases: Jean Noirot1; Philippe Bienvenu1; Isabelle Zacharie-Aubrun1; Karine Hanifi1; Laurent Fayette1; Aurelien Moy1; Yves Pontillon1; 1CEA
In oxide nuclear fuels, for typical power histories, the fission gas release rate tends to increase with the burn-up. In spite of this, the amount of fission gas stored in micron or sub-micron size bubbles also increases. These bubbles are not homogeneously distributed in the fuel pellets and their characteristics and formation modes are different, from hot central parts to rim high burn-up structure or Pu rich agglomerates of heterogeneous MOX fuels. Detailed analyses of commercial reactor irradiated fuels and of test reactor dedicated program fuels have been conducted, before and after experiments on accidental conditions. It was shown that during accidental situations, fuel pellet cracking and fragmentation can be related to these porous areas. Still, differences in bubble formation according to the fuel characteristics have been found. This gives a lead on where to look for improving the fuel resistance to bubble formation and to fuel fragmentation during accidents.
10:50 AM Invited
Irradiation Effects on Electrochemical Performance of TiO2 Anode: Janelle Wharry1; Kassiopeia Smith2; Hui Xiong2; Darryl Butt3; 1Purdue University; 2Boise State University; 3University of Utah
The objective of this study is to understand the effect of irradiation on electrochemical charge storage capacity in TiO2 electrodes. Nanostructured TiO2 is a promising anode material for lithium-ion batteries, and it has been shown that cation defects in metal oxides increase the extent of lithium intercalation, thereby increasing charge storage capacity. We theorize that irradiation-induced defect production is an effective means for improving TiO2 electrode performance. In this study, we first focus on separating the effects of irradiating species and crystallographic orientation on defect production by conducting room-temperature proton, O2+, and Nb2+ irradiations on , , and  rutile TiO2 single crystals. We then extend our study to nanostructured TiO2, irradiated with protons or O2+ ions. Though we observe reduction in charge/discharge capacity potentially due to proton irradiation, we observe an extension of the discharge capacity due to O2+ ion irradiation.
Role of Ion Species in Radiation Effects of Lu2Ti2O7: Dongyan Yang1; Yuhong Li1; 1Lanzhou University
In an attempt to investigate the role of ion species in the radiation effects of pyrochlores, polycrystalline Lu2Ti2O7 samples were irradiated with three different ion beams: 400 keV Ne2+, 2.7 MeV Ar11+ and 6.5 MeV Xe26+. Grazing incident X-ray diffraction technique was applied to characterize the damaged layers in the sample. All the three irradiations induce significant amorphization processes and lattice swelling. However, when the ion fluence is converted to a standard dose as in dpa, the radiation effects of Lu2Ti2O7 exhibit a great dependence on the implanted ion species. The threshold amorphization dose decreases with increasing ion mass and energy. Besides, the amorphization rate, as well as lattice swelling rate, increases with increasing ion mass and energy. These results are discussed in the framework of defect types and the density of collision cascades based on SRIM simulations.
In-Situ Tritium Measurements from γ-LiAlO2 Pellets Irradiated in TMIST-3A: Walter Luscher1; David Senor1; Kevin Clayton2; 1Pacific Northwest National Laboratory; 2Idaho National Laboratory
TMIST-3 is an instrumented lead test consisting of two separate test trains and a total of 41 individual capsules containing a variety of pellets with different microstructural features subjected to a variety of irradiation conditions. The test trains will be irradiated sequentially in the Advanced Test Reactor and in-situ tritium measurements from the nine flow-through capsules in the TMIST-3A test train are the subject of this presentation. Time, burnup, and burnup rate dependence of the total tritium release from the pellets will be evaluated and the effects of microstructure on tritium release will be evaluated by irradiating pellets with varying grain size, porosity, and pore morphology. Finally, comparisons between ceramic LiAlO2 pellets and cermet LiAlO2/Zr pellets will be made to better understand tritium retention and release mechanisms. Measurement results will be compared with a diagnostic model to support data interpretations and provide insight into the governing mechanisms for tritium release.