Karl R. Whittle Australian Nuclear Science and Technology Organisation
Marjorie Bertolus CEA, DEN, DEC/SESC/LLCC
Blas Uberuaga Los Alamos National Laboratory
Robin W. Grimes Imperial College London
A1: Nuclear Fuels I
Monday AM, November 28, 2011
Independence W (Sheraton)
9:30 AM - **A1.1
Radiation Resistance of UO2 under Severe Damaging Conditions.
Thierry Wiss 1 , Arne Janssen 1 , Hartmut Thiele 1 , Bert Cremer 1 , Jean-Yves Colle 1 , Dragos Staicu 1 , Vincenzo Rondinella 1 , Rudy Konings 1 Show Abstract
1 , European Commission - JRC - ITU, Karlsruhe Germany
The most commonly used nuclear fuel, UO2, is subjected to radiation damage not only during its in-pile irradiation, but also during cooling and storage. Magnitude, rate and conditions of the damage accumulation are different for reactor irradiation and for (long time) storage conditions, but to some extent the damage pattern is very similar. During irradiation in nuclear reactor, each atom in the fuel experiences several thousand displacements from its initial lattice position. A large amount of energy, mainly generated by the fission, is dissipated in the lattice and causes the formation of defects. Driven by power and temperature gradients and as a consequence of radiation damage the properties of the fuel change significantly with increasing burnup. Defects generated in the fuel structure (point and extended defects, micro- and macro-bubbles, solute and segregating impurities) will alter key properties, like e.g. thermal conductivity, density and mechanical properties, which determine the performance and ultimately the safety of the fuel. Future reactor concepts envisage the use of fuel (and materials) up to higher burnup and more severe irradiation conditions; moreoverthey are often characterized by higher Pu- and minor actinide-content, which results in higher alpha-decay damage extent. The fuel after irradiation and during storage is still very radioactive. The long timescale considered for storage in many countries requires understanding of the damage mechanisms and developing suitable tools to predict the fuel evolution. Properties relevant for safe handling/processing of high specific alpha-activity fuels are strongly affected by the build-up of alpha-decay damage and helium. This is the object of a campaign of studies carried out at JRC-ITU, which covers in particular the evolution of thermal transport and mechanical properties as a function of accumulated radiation/decay damage and He.In order to simulate alpha-damage accumulation in UO2 spent fuels aged for periods corresponding up to a few thousand years, samples doped with short-lived alpha-emitters (e.g. 238Pu) have been fabricated and characterized. The alpha-damage accumulation affects many properties of UO2 like thermal diffusivity, lattice parameter, heat capacity, showing a rapidly saturating behaviour. Irradiated fuels have been characterized by different techniques including transmission electron microscopy, X-ray diffractometry, in combination with thermal annealing methods. Comparative analysis of spent fuel and alpha-doped materials allows assessing superimposition of alpha-decay effects after fuel discharge onto radiation damage occurred in-pile. It was shown that the damaged microstructure of irradiated fuel and of UO2 doped with alpha emitters is very similar despite the large difference in the conditions under which the damage occurred. UO2 shows a remarkable ability to maintain its original fluorite structure even under severe irradiation conditions.
10:00 AM - A1.2
Sesquioxide Effect on Thermal Diffusion Processes in UO2.
Simon Middleburgh 1 2 , Robin Grimes 1 , Paul Blair 2 , Karin Oldberg 2 Show Abstract
1 Department of Materials, Imperial College London, London United Kingdom, 2 Materials and Fuel Rod Design, Westinghouse Electric Sweden, Vasteras Sweden
The effects of trivalent cation solution on fission gas release has been studied by calculation of a number of transition states for diffusion processes within UO2. Reduction in the energy required for a vacancy migration to take place has been observed with solution of all trivalent cations, the larger reductions occurring with the smaller cations aluminium and chromium, both suggested fuel dopants. A qualitative comparison of the diffusion co-efficient for chromium doped fuel with undoped fuel has been made, which suggests that higher Cr concentrations will be associated with higher xenon diffusivity (due to enhanced vacancy migration).
10:15 AM - A1.3
Effect of Lanthanide and Actinide Substitution in UO2 Using Atomic Level Simulations.
Rakesh Behera 1 , Chaitanya Deo 1 Show Abstract
1 Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering , Georgia Institute of Technology, Atlanta, Georgia, United States
Uranium-based fuels are the most common fuel used for commercial nuclear energy generation. The complete fuel cycle based on UO2 fuels generates a large number of transuranic nuclides (Pu, Am, Np, Cm). These fission products influence a variety of properties. While the nuclear fuel cycle is well characterized, the understanding of the physical and chemical properties of the actinides is still limited. This study focuses on characterizing the effect of dilute concentrations of Lanthanides and Actinides on bulk properties of UO2. In particular, the results will include the effect of elastic and electrostatic effects due to the substitution of +4e- (Am, Pu, Ce, Np, U, Th) and +3e- (Gd, Eu, Sm, Am, Nd, Pu, U) ions in the UO2 lattice. The discussions will be based on the experimentally observed concentrations of Lanthanides and Actinides in urania using atomic level simulations.
10:30 AM - A1.4
Effect of Cr Segregation to UO2 Grain Boundaries.
Minki Hong 1 , Simon Phillpot 1 , Blas Uberuaga 2 , Chris Stanek 2 , Susan Sinnott 1 Show Abstract
1 MSE, University of Florida, Gainesville, Florida, United States, 2 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
The UO2 fuel pellet has a polycrystalline microstructure and the density and the size of each grain are the key to control the fuel performance particularly by modifying its thermal conductivity. A significant amount of research has been conducted to improve these properties by doping sintering additives and Cr has been suggested as one of the elements that have the capability of grain enlarging especially during the sintering process of UO2 pellet. However the mechanism of the grain enlarging and the effect of Cr on grain boundary behavior under actual operating condition are not well understood. Here, atomic-level simulation methods using empirical interatomic potentials are used to examine segregation of Cr to UO2 grain boundaries and understand its grain enlarging mechanism. In addition, the quantitative energetics of Cr near the grain boundary and its chemical or bonding environment are examined using density functional theory calculations with the Hubbard U approximation. The results indicate that Cr is mostly insoluble in UO2 unless it substitutes uranium under hyper-stoichiometric condition and the segregation energy of Cr to the Σ5 tilt boundary with (310)/(001) plane is about 2.8 eV.
10:45 AM - A1.5
Crack Tip Plasticity in Single Crystal UO2: Atomistic Study.
Yongfeng Zhang 1 , Xiangyang Liu 2 , Bulent Biner 1 , Paul Millett 1 , Michael Tonks 1 , David Andersson 2 Show Abstract
1 Fuel Modeling and Simulation, Idaho National Lab, Idaho Falls , Idaho, United States, 2 Structure/Property Relations, Los Alamos National Lab., Los Alamos, New Mexico, United States
The room temperature fracture behavior of single crystal UO2 is studied using molecular dynamics (MD) simulations with the Basak potential. The cracks are introduced on two low-index charge neutral planes, the (111) and (110), and the mode-I loading is applied normal to the crack planes. At the onset of growth of the cracks, plastic deformations such as dislocation emission and phase transformations are observed at the crack tips. The dislocations are characterized as ½<110> full dislocation gliding on the (001) plane. Two metastable phases are identified as Rutile and Scrutinyite structures, and their formation is confirmed by separate density-functional-theory calculations. The cracks residing on the (111) plane propagate along the high-energy incoherent boundaries between the ground Fluorite and the newly formed metastable phases. In the case of cracks located on the (110) plane, the new phases form coherent boundaries. As a result, the stress at the crack tips is largely reduced; and no crack extension is observed.
11:30 AM - A1.6
Multiscale Fuel Performance Simulation of Metallic Reactor Fuels.
Michael Tonks 1 , Paul Millett 1 , Bulent Biner 1 Show Abstract
1 Fuels Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States
Metallic fuel is a popular option for Generation IV nuclear reactors. However, the nuclear industry lacks the years of operational experience with metallic fuel that they have with UO2 fuels. Thus, an accurate science-based model of metal fuel performance could be a powerful tool for investigating metal fuel performance in typical and accident conditions. A predictive fuel performance model must account for microstructure evolution, and to develop such a model requires input at the atomistic, meso- and engineering-scales. In this research, atomistic simulation is used to develop important parameters, such as point defect mobilities, and to identify critical mechanisms. Mesoscale phase field models then use this information to predict microstructure evolution due to external conditions, such as loading and radiation damage. The mesoscale model then determines the effect of the microstructure evolution on various bulk material properties, including thermal conductivity and density. These mesoscale-informed properties are used in the engineering-scale fuel performance simulation to predict the thermal and mechanical behavior of metallic fuel during its lifetime in the reactor.
11:45 AM - A1.7
Chemical Behavior of Oxide Nuclear Fuel: Recycle and High Burn-up.
Theodore Besmann 1 , Stewart Voit 1 , Dongwon Shin 1 , Evan Noon 1 , Robert Austin 1 Show Abstract
1 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Thermochemical models of oxide nuclear fuel systems containing transuranic and fission product elements are being developed. Specifically, subsystems of major actinides with fission products are being represented by solid solution models such as the subregular model for the 5-metal white phase and the compound energy formalism sublattice approach for variable stoichiometric oxides such as the fluorite-structure fuel phase. Current work has emphasized the behavior of actinides with rare earths as these are important for both fuel recycle where rare earth elements in significant concentrations remain with the actinides, and in-reactor where they influence stoichiometry and oxygen potential. This report will discuss recent experimental and modeling efforts related to the behavior of rare earths in the fuel phase, and the overall complexity and importance of oxygen behavior in fuel.This work was supported by the US Department of Energy Office of Nuclear Energy, Fuel Cycle Research and Development Program.
12:00 PM - A1.8
Spark Plasma Sintering of Fuel Cermets for Nuclear Reactor Applications.
Yang Zhong 1 2 , Robert O'Brien 1 , Steve Howe 1 , Nathan Jerred 1 , Kristopher Schwinn 1 , Amy Kaczmarowski 1 , Joshua Hundley 1 , Laura Sudderth 1 Show Abstract
1 , Center for Space Nuclear Research, Idaho National Lab, Idaho Falls, Idaho, United States, 2 Department of Chemical, Materials and Biomolecular Engineering, University of Connecticut, Storrs, Connecticut, United States
The recent events with the reactors in Fukishima, Japan revealed a need for a high temperature fuel form that will not melt down from decay heat after a loss of coolant accident. Furthermore, the fuel material should contain the fission products from dispersion during a combination of accidental high temperature excursions and steam/hydrogen explosions. Such requirements will necessitate a new robust fuel encapsulation matrix. The Center for Space Nuclear Research has been developing a new fuel form (fuel cermets) for nuclear reactors to be used for space exploration. Fuel Cermets consist of a tungsten-rhenium (W/Re) encapsulating matrix and a ceramic compound (a nuclear fuel such as uranium in its oxide form). Owing to the good thermal conductivity, mechanical strength, hardness, and high melting point of W/Re alloys, as well as their ability to contain fission products, tungsten cermet fuels are highly attractive for applications where enhanced nuclear reactor safety and proliferation resistance is essential. In this study, an analysis of cermet fuels produced via Spark Plasma Sintering (SPS) is provided. SPS processing can greatly reduce the average sintering temperature and minimize the grain growth during production in comparison to traditional sintering techniques. In the examples presented, CeO2 is used to simulate fuel loadings of UO2 or Mixed-Oxide (MOX) fuels within tungsten-based cermets due to the similar kinetic properties of these materials, in particular their respective melting points and Gibbs free energies. The densification of the cermets with respect to the volumetric ratios between metal and ceramic, the grain size and the morphology of the starting powder is systematically studied in this work. The sintering kinetics is also investigated using master sintering curve.
12:15 PM - A1.9
Effect of Intergranular Gas Bubbles on Thermal Conductivity.
Karthik Chockalingam 1 , Paul Millett 1 , Michael Tonks 1 , Bulent Biner 1 , Liangzhe Zhang 1 , Yangfeng Zhang 1 Show Abstract
1 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho National Laboratory, Idaho Falls, Idaho, United States
Model microstructures obtained from phase-field simulations are used to study the effective heat transfer across systems with stationary grain boundary bubble populations. From the analysis it is found that grain boundary bubbles represent a larger impediment to thermal transfer than traditional ‘rule-of-mixture’ theories predict. Additionally, we find that the grain boundary coverage, irrespective of the intergranular bubble radii, is the most relevant parameter to the actually thermal resistance, which we use to derive ‘effective’ Kapitza resistances. Models have been proposed to predict thermal conductivity as a function of porosity, grain size and grain boundary bubble coverage.
12:30 PM - A1.10
Phase-Field Modeling of Pore Migration in Nuclear Fuels Due to a Temperature Gradient.
Liangzhe Zhang 1 , Michael Tonks 1 , Paul Millett 1 , Bulent Biner 1 , Yongfeng Zhang 1 , Karthikeyan Chockalingam 1 Show Abstract
1 Fuels Modeling and Simulation, Idaho National Laboratory, Idaho Falls, Idaho, United States
Sintered UO2 nuclear fuel materials undergo a unique microstructural evolution process during the course of the burn-ups. The evolved microstructure is usually characterized by the columnar grains surrounding a large central void, which mainly results from the migration of the initial pores towards the high temperature regions. A quantitative description of the pore migration process is therefore desirable for better understanding and accurate predictions of the fuel performance. For this purpose, a phase-field model is developed; in which the kinetics of the migration due to both bulk and surface diffusion is formulated by utilizing fourth order Cahn-Hillard (CH) equations. The results indicate that the porosities migrate towards the high temperature region owing to the temperature gradient as the driving force, which are consistent with the experimental observations. Furthermore, it is also seen that a pore can also changes its shape due to the small variations of temperature profile at its surrounding regions.
12:45 PM - A1.11
HIP Bonded U-10Mo Monolithic Fuel Plates with a Modified Fuel to Cladding Interface.
Jan-Fong Jue 1 , Blair Park 2 , Cynthia Breckenridge 1 , Jeffery Hess 1 , Glenn Moore 2 , Dennis Keiser 1 , Daniel Wachs 1 Show Abstract
1 Fuel Performance & Design, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Fuel Fabrication, Idaho National Laboratory, Idaho Falls, Idaho, United States
Under the RERTR (reduced enrichment for research and test reactors) program, the nuclear fuels used in research and test reactors are being converted from highly enriched to proliferation-resistant low-enriched uranium (LEU), defined as less than 20% U235 enrichment. One of the fuel designs under development is a monolithic fuel type where the fuel is in the form of a single U-10Mo (uranium - 10 wt% molybdenum) alloy foil. With monolithic fuel, a uranium fuel density of more than 10 g/cm3 can be achieved. The post irradiation results from previous RERTR irradiation experiments indicate that the addition of silicon to the fuel-to-cladding interface resulted in reduced fuel/matrix chemical interaction and increased the stability of the interaction layer during irradiation. This paper provides an update of the developmental effort on the fabrication of monolithic fuel type by the HIP (hot isostatic press) bonding process with a silicon enriched fuel/cladding interface.
A2: Radiation Damage - Ceramics
Monday PM, November 28, 2011
Independence W (Sheraton)
2:30 PM - A2.1
Self-Healing Response of Ionic Crystals to Irradiation: Can Damage be Good?
Dilpuneet Aidhy 1 , Dieter Wolf 1 Show Abstract
1 , ANL, Argonne, Illinois, United States
Molecular dynamics simulations of irradiated CeO2 (often considered a surrogate for UO2, the most widely used nuclear fuel) reveal the formation of charge-neutral interstitial dislocation loops identical to ones observed recently in experiments. Focusing on the kinetic phase that follows the initial damage cascade, our simulations of the cluster-formation mechanism reveal a self-healing response of the perfect crystal to the radiation-induced defects. Remarkably, the lattice responds to point defects created during irradiation with the spontaneous creation of new point defects. We demonstrate that these new ‘structural defects’, with a negative energy of formation, neutralize the cluster by screening its long-range Coulomb potential, thereby lowering the overall energy and localizing the damage. A similar lattice response was recently identified also in simulations of MgO, although very different types of clusters were formed, suggesting that this self-healing screening response may be an intrinsic reaction of all ionic crystals to irradiation.
2:45 PM - A2.2
Dynamic Recovery in Silicate Apatite Structures under Irradiation and Implications for Long-Term Performance Modeling.
William Weber 1 2 , Yanwen Zhang 2 1 , Haiyan Xiao 1 , Lumin Wang 3 Show Abstract
1 Materials Science & Engineering, University of Tennesee, Knoxville, Tennessee, United States, 2 Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 3 Nuclear Engineering & Radiological Sciences, The University of Michigan, Ann Arbor, Michigan, United States
The irradiation responses of Ca2La8(SiO4)6O2 and Sr2Nd8(SiO4)6O2 with the apatite structure are investigated to predict their long-term behavior as host phases for immobilization of actinide elements from the nuclear fuel cycle. Different ions and energies are used to study the effects of dose, temperature, atomic displacement rate and ionization rate on irradiation-induced amorphization and recrystallization. The dose for amorphization increases with temperature in two stages, below and above 150 K. In the high temperature stage relevant to actinide immobilization, the increase of amorphization dose with temperature exhibits a strong dependence on the ratio of ionization rate to displacement rate for the different ions. Data analysis using a dynamic model for amorphization reveals that ionization-induced processes, with activation energy of 0.15 ± 0.02 eV, dominate dynamic recovery for ions from Ne through Xe. For heavier Au ions or for alpha-recoil nuclei emitted in alpha decay of actinides, ionization becomes less dominant and dynamic recovery is controlled primarily by thermally-driven processes. In post-irradiation annealing studies of amorphous samples, epitaxial thermal recrystallization is observed at 1123 K, and irradiation-enhanced nucleation of nanocrystallites is observed in situ under irradiation with heavier ions. The recrystallization temperature under irradiation decreases with increasing ion mass to a value of ~ 823 K, which also defines the thermally-driven critical temperature for amorphization under irradiation with heavy ions. Some partial recovery due to alpha particle irradiation at 300 K is observed that suggests a self-healing mechanism in apatite phases containing actinides. Based on the results and dynamic model, the temperature and time dependence of amorphization in apatite host phases for actinide immobilization are predicted.
3:00 PM - A2.3
Defects, Minor Phases and Microstructures in O+ and Zr+ Ion Co-implanted Strontium Titanate: A Model Nuclear Waste Form.
Weilin Jiang 1 , Mark Bowden 1 , Zihua Zhu 1 , Libor Kovarik 1 , Bruce Arey 1 , Przemyslaw Jozwik 2 3 , Jacek Jagielski 2 3 , Anna Stonert 3 Show Abstract
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , Institute of Electronic Materials Technology, Warsaw Poland, 3 , The Andrzej Soltan Institute for Nuclear Studies, Otwock Poland
Operations in the nuclear power industry generate spent fuels that contain highly radioactive materials. The fission products of 235U and 239Pu have a high percentage of 90Sr and 137Cs isotopes in the nuclear waste stream from reprocessing of the spent fuels. Long-term storage or permanent disposal of nuclear wastes requires stabilization of the highly radioactive materials in a solid form that degrades very slowly over time, thereby limiting the radionuclide release to the environment. Discovery and development of such forms for immobilization of nuclear wastes are critical for proper management of existing and future spent fuels.Single-crystal strontium titanate (SrTiO3 or STO) is used in this study as a model material to simulate a waste form for disposal of radionuclide 90Sr that decays to 90Y and subsequently to 90Zr. The self-irradiation from decay electrons, charge state change from 90Sr2+ to 90Zr4+ and substantial heat production can significantly affect the structural stability of the host material and potentially lead to phase transformation, phase separation, and/or formation of new phases. In this study, sequential implantation of 16O+ and 90Zr+ ions was performed for STO at 550 K to minimize the charge imbalance and to avoid full amorphization of STO. Each of the implant concentrations of up to 1.5 at% was achieved. Post thermal annealing was conducted in flowing Ar gas environments at temperatures up to 1423 K for 10 hours. Various experimental methods have been employed to characterize the implanted sample, including time-of-flight secondary-ion mass spectroscopy, multiaxial ion-channeling analysis, high-resolution transmission electron microscopy, and micro-beam x-ray diffraction. The results show that, in contrast to the observed mobile Sr interstitials in STO, the implanted Zr does not diffuse noticeably during the ion implantation or thermal annealing up to the highest applied temperature (1423 K) in this study. This behavior is attributed to the formation of strong chemical bonding of the implanted Zr in the structure. A defect concentration was generated in STO and nearly all the implanted Zr was not located exactly at the original lattice site. There are Zr-containing microstructures, precipitates and voids (or oxygen blisters) in the implanted layer. A minor phase with a tetragonal structure was also observed. It survived thermal annealing at temperatures up to 1423 K with only a small decrease in the lattice parameter. Discussion about the results and a general assessment of the model waste form will be provided in this presentation.
3:15 PM - A2.4
Post Irradiation Examination of Neutron Irradiated Inert Matrix Ceramics.
Donald Moore 1 , Cynthia Papesch 2 , Brandon Miller 2 , Pavel Medvedev 2 , Juan Nino 1 Show Abstract
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
There is an increasing radiotoxic inventory of nuclear waste requiring safe, ecologically friendly, and economically sensible disposal. A promising approach for reducing waste from spent nuclear fuel and weapons programs is by utilizing an inert matrix fuel (IMF) for the transmutation of waste in light water reactors (LWRs). Implementation of an IMF requires a stable and radiation tolerant inert matrix material with similar thermophysical and neutronic properties as UO2.Several potential inert matrix materials and other ceramic materials including MgO, Nd2Zr2O7, MgO-Nd2Zr2O7 cercer composite, MgAl2O4, MgO1.5Al2O3, and Mg2SnO4 pellets were irradiated in-pile of the Advanced Test Reactor at Idaho National Laboratory. The effects of irradiation temperatures (~350 and ~700°C) and fast neutron fluencies (~1x10^25 and ~2x10^25 n/m2) on the materials properties are currently being investigated. Post irradiation examination includes thermal diffusivity, scanning electron microscopy, and transmission electron microscopy. The radiation induced thermophysical and structural evolution of MgO and MgAl2O4 will be presented. We will discuss the effects of irradiation damage on the thermal diffusivity of MgO and MgAl2O4. Comparison of the different irradiation conditions verse non-irradiated samples gives details of how defects affect the thermal diffusivity.
3:30 PM - **A2.5
Ion Irradiation Induced Defects in Non-Metallics.
Philip Edmondson 1 , Fereydoon Namavar 2 , Robert Birtcher 3 , Jonathan Hinks 4 , Stephen Donnelly 4 , William Weber 5 1 , Yanwen Zhang 1 5 Show Abstract
1 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 , University of Nebraska Medical Center, Omaha, Nebraska, United States, 3 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States, 4 School of Computing and Engineering, University of Huddersfield, Huddersfield, West Yorkshire, United Kingdom, 5 Department of Materials Science and Engineering, University of Tennessee, Knoxville, Tennessee, United States
The binary oxide ceramics CeO2 and ZrO2 are key engineering materials in nuclear systems; either as structural materials, as possible inert fuel matrices, or as nonradioactive surrogates in studies of nuclear fuel systems. Recently, the nanocrystalline phases of these materials have come under increased interest due to their enhanced properties, and the ability to tailor these properties with grain size. In the work presented here, thin films of both cubic ceria and zirconia have been irradiated with Au+ ions to doses up to 35 displacements per atom (dpa), over a range of temperatures from 160 to 400K. Subsequent examination was performed using a combination of Rutherford Backscattering spectroscopy (RBS), scanning transmission electron microscopy (S/TEM), atom probe tomography (APT) and x-ray diffraction (XRD). In both films, the cubic phase is retained despite the relatively high dpa levels. Grain growth was also observed in all cases and may be attributed to the production of high levels of defects near the grain boundaries. In the zirconia film, the XRD results showed a lattice variation during the irradiation – the extent of which was dependent on the irradiation temperature. This lattice deviation is attributed to the generation and saturation of oxygen vacancies of different charge states being formed during irradiation. The ceria film showed no such lattice variation, and no enhanced oxygen deficiency was observed under ion irradiation. Analysis of the symmetry of the grain boundaries (GBs) showed that the initial film was dominated by asymmetric GBs and that during the irradiation symmetric GBs begin to dominate, reaching a saturation at dpa values of ~5, indicating a reduction in energy of the film. Dark bands of contrast were also formed at the film/substrate interface. STEM-EDS results indicate that the band was an ion-beam induced, chemically-mixed Ce/Zr-Si phase. In addition, amorphization processes in elemental (Si) and ternary (CuInSe2 (CIS)) semiconductors, as studied by TEM during in situ ion irradiation will be discussed. It will be shown that the interstitial-vacancy pair may be the dominant defect formed in Si during ion irradiation. The CIS proved to be resistant to amorphization at temperatures above 200K under the irradiation conditions used. At and below 200 K, amorphization only occurred in the samples that were Cu deficient relative to perfectly stoichiometric CIS.
4:30 PM - A2.6