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
Pyrochlore-Fluorite Transitions in Y2Sn2-xZrxO7: Implications for Stability.
Massey de los Reyes 1 , Karl R Whittle 1 , Robert G Elliman 2 , Nestor J Zaluzec 3 , Sharon E Ashbrook 4 , Martin R Mitchell 4 , Gregory R Lumpkin 1 Show Abstract
1 Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 2 Department of Electronic Materials Engineering, Australian National University, Canberra, Australian Capital Territory, Australia, 3 Materials Science Division, Argonne National Laboratory, Chicago, Illinois, United States, 4 School of Chemistry, University of St. Andrews, St. Andrews, Fife, United Kingdom
The Y2Sn2-xZrxO7 pyrochlore series undergoes a phase transformation from a cubic pyrochlore structure-type (Fd3m) to defect fluorite (Fm3m) actuated by the increase in Zr content, coupled with thermal annealing above 1500°C. The pyrochlore-fluorite transition is an important factor in determining how materials behave under irradiation whether as a waste form or other nuclear material (e.g., inert matrix fuel, transmutation target, oxygen dispersion strengthened ODS additives). X-ray diffraction analysis reveals the onset of a pyrochlore to defect fluorite transition at Y2Sn1.6Zr0.4O7 with the loss of long range ordering. This is confirmed further by selected area diffraction, illustrating shorter range ordering in the defect fluorite phase incommensurate with unit cell size. However, this transformation occurs at a much higher Zr content than that predicted by classical radius ratio models. The diffuse scattering features observed in electron diffraction patterns of defect fluorite phases indicate some form of longer range ordering involving compositional-displacive structural modulation. The behavior of these materials during irradiation will be discussed and linked with the observed structural parameters (diffuse scattering, unit cell size).
4:45 PM - A2.7
The Role of Sn, Zr and Hf in the Radiation Damage in II,III, IV, and V Pyrochlores.
Karl Whittle 1 , Massey de los Reyes 1 , Yan Gao 1 , Mark Blackford 1 , Nestor Zaluzec 2 , Gregory Lumpkin 1 Show Abstract
1 Materials Engineering, ANSTO, Sydney, New South Wales, Australia, 2 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States
Ceramics based on the general formulation CaLnZrNbO7 (Ln = La, Nd, Sm, Gd and Ho) have been studied as model four-component oxide systems, utilising the observation that they contain 2+, 3+, 4+ and 5+ cations. X-ray and neutron diffraction results show the materials to adopt pyrochlore across the series. Using a combined structural refinement the cation disorder across the A/B cation sites has been determined. The samples have subsequently been irradiated using the IVEM-TANDEM facility at Argonne National Laboratory (ANL), with 1 MeV Kr2+ ions at various temperatures. The results show a decrease in the critical temperature for amorphisation (Tc) from ~ 680 K for CaLaZrNbO7 to < 50K for CaHoZrNbO7. The change in Tc is discussed with references to disorder across cation sites, changes within the structure, and how these affect the radiation damage response. The results are also used to expand further the reliability of predicting those systems which are more tolerant of radiation damage, and how it can be used to develop new waste forms and other nuclear materials. Complementary systems containing Sn and Hf will also be discussed, in both how they respond and how they compare with CaLnZrNbO7 materials.
5:00 PM - A2.8
Structural Features in Fluorite Compounds Relevant for Nuclear Applications.
Gianguido Baldinozzi 1 2 , David Simeone 2 1 , Dominique Gosset 2 1 , Laurence Luneville 2 1 , Lionel Desgranges 3 Show Abstract
1 SPMS, MFE, CNRS, Chatenay-Malabry France, 2 DEN, DANS, DMN, SRMA, MFE, CEA, Gif-sur-Yvette France, 3 DEN, DEC, SESC, CEA, St Paul lez Durance France
Oxides with fluorite (or fluorite related) structures form a large class of compounds with a high radiation tolerance, somewhat related to their peculiar ability to accommodate a variety of defects and to form nonstoichiometric compounds with a large homogeneity range. Structural modifications are generally observed when the departure from the ideal composition is large. We would like to discuss these structural features using an approach based on the crystal symmetry analysis and to address some of the phase transition mechanisms in compounds relevant for nuclear applications.
5:15 PM - A2.9
Swift Heavy Ion Induced Amorphization of ZrO2.
Fengyuan Lu 1 , Maik Lang 2 , Jianwei Wang 2 , Fereydoon Namavar 3 , Christina Trautmann 4 , Rodney Ewing 2 , Jie Lian 1 Show Abstract
1 Mechanical, Aerospace and Nuclear Engineering, RPI, Troy, New York, United States, 2 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 3 , University of Nebraska Medical Center, Omaha, Nebraska, United States, 4 , GSI Helmholtzzentrum Schwerionenforsch, Darmstadt Germany
ZrO2 is an important engineering material as fuel matrix and nuclear waste forms, and the behavior of ZrO2 upon displacive and ionizing radiations is both scientifically and technologically crucial. Bulk monoclinic ZrO2 displays an excellent radiation tolerance and cannot be amorphized by displacive irradiation. However, we found that the extreme ionizing radiation by swift heavy ion of 1.33 GeV U-238 can induce amorphization of monoclinic ZrO2 with a grain size of ~ 50 nm. A similar amorphization trend was observed in 1.33 GeV U-238 irradiated tetragonal ZrO2. A computational simulation based on the thermal spike model demonstrates that with a very high electronic energy loss of 52.2 KeV/nm, the 1.33 GeV U-238 irradiation causes high transient temperatures in ZrO2 lattice beyond the melting point, resulting in the amorphization of the monoclinic ZrO2. An electronic energy loss threshold can be implied above which the radiation induced amorphization can occur in ZrO2. This work also highlights the potential of controlling ZrO2 phase by varying radiation conditions.
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
A3: Nuclear Fuels II
Tuesday AM, November 29, 2011
Independence W (Sheraton)
9:15 AM - A3.1
Computational Modeling of Iso- and Aliovalently Doped ThO2 and UO2.
Vitali Alexandrov 1 2 , Niels Gronbech-Jensen 3 , Alexandra Navrotsky 1 4 , Mark Asta 2 1 Show Abstract
1 Department of Chemical Engineering and Materials Science and NEAT ORU, University of California, Davis, Davis, California, United States, 2 Department of Materials Science and Engineering, University of California, Berkeley, Berkeley, California, United States, 3 Department of Applied Science, University of California, Davis, Davis, California, United States, 4 Peter A. Rock Thermochemistry Laboratory and NEAT ORU, University of California, Davis, Davis, California, United States
The thermodynamic and kinetic properties of ThO2 and UO2 based solid solutions are of direct relevance for nuclear-fuel applications. In this talk we present computational results obtained by density-functional-theory (DFT) calculations coupled with cluster expansion and Monte-Carlo simulations. We examine the nature of the defect clusters and the consequences of such defect clustering tendencies for the composition and temperature dependencies of thermodynamic properties. Oxygen stoichiometric solid solutions of both ThO2 and UO2 with isovalent solute additions are shown to have positive mixing enthalpies, with the cation size mismatch being the dominant energy contribution, in quantitative agreement with continuum elasticity theory. ThO2-based solid solutions containing trivalent dopants (Sc, In, Y, Nd, Ce, La) are calculated to have positive enthalpies of formation with respect to constituent oxides and show a tendency to decrease in magnitude as the size and electronegativity of the trivalent dopant decrease. Monte Carlo simulations at elevated temperatures establish that the solid solutions display a significant degree of short-ranged defect clustering. These results are contrasted with those obtained for aliovalently doped UO2.
9:30 AM - A3.2
Electrochemistry of Defects in Irradiated UO2.
Abdel-Rahman Hassan 1 , Thomas Hochrainer 2 , Jianguo Yu 3 , Xianming Bai 3 , Todd Allen 4 , Anter El-Azab 2 Show Abstract
1 Materials Science and Engineering Program, Florida State University , Tallahassee, Florida, United States, 2 Department of Scientific Computing , Florida State University, Tallahassee, Florida, United States, 3 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 4 , University of Wisconsin, Madison, Wisconsin, United States
Irradiation alters the local stoichiometry of oxides significantly. The resulting stoichiometric changes play a critical role in the dynamics of defects and microstructure evolution in oxides under irradiation. Stoichiometry in oxides is also sensitive to the surrounding oxygen environment. In general, the levels of point defects and electronic charge carriers in an oxide are sensitive to the oxygen partial pressure in contact with the oxide at hand. We investigate the electrochemistry of defects in UO2 under irradiation, where both the atomic displacements by energetic collision cascades and the exchange of oxygen with the ambient drive stoichiometric changes in the material. The problem is cast in the form of balance laws of lattice and electronic defects under defect generation and diffusion, with boundary conditions dictated by the oxygen partial pressure at the free surface. Inherent to this problem is the electrostatic field resulting from the segregation of charged lattice and electronic defects in the material. Using this model, the scenario of dynamic stoichiometric changes in a UO2 film under ion irradiation will be illustrated in detail. This research was supported as a part of the Energy Frontier Research Center on Materials Science of Nuclear Fuel funded by the U.S. Department of Energy, Office of Basic Energy Sciences under subcontract # 00091538 from INL to Florida State University.
9:45 AM - A3.3
First-Principles Modeling of Defects Behavior in Ceramic Fuels.
Ying Chen 1 , Hua Y. Geng 2 , Yasunori Kaneta 3 , Jia C. Shang 4 , Motoyasu Kinoshita 5 , Shuichi Iwata 3 Show Abstract
1 Department of Nanomechanics, Tohoku University, Sendai, Miyagi, Japan, 2 , Institute of Fluid Physics, Mianyang, Sichuan, China, 3 , The University of Tokyo, Tokyo Japan, 4 , Nuclear Power Institute of China, Chengdu, Sichun, China, 5 , Central Research Institute of Electric Power Industry, Tokyo Japan
Uranium dioxide is a most important fuel material used in nuclear reactor, its performance quite relates to the defects behavior under irradiation which arises the deviation from stoichiometric compounds. To investigate the formation, stability mechanism and relevant physical properties of the nonstoichiometric uranium dioxide, comprehensive first principles calculations have been performed using PAW-LSDA+U method for various complex defects clusters of oxygen atoms in UO2. Calculations revealed the stability of the cuboctahedron embedded into the crystal UO2, clarified the ambiguousness remaining for long in structure of nonstoichiometric UO2+x. By incorporating the temperature effect, a pseudo phase diagram of temperature and the oxygen concentration has been constructed, and a new physical model of thermodynamic competition between cuboctahedron and point oxygen interstitials is proposed. The interplay of one main fission products, Xe, and the defect clusters in ceramics fuels has been also investigated.
 Ying Chen, Hua Y. Geng, Shuichi Iwata, et al., Comp. Mater. Sci. 49 (2010), S364
 H. Y. Geng, Ying Chen, Y. Kaneta, M. Kinoshita and Q. Wu, Phys. Rev. B 82 (2010), 094106
 H. Y. Geng, Y. Chen, Y. M. Kinoshita, et al., Applied Physics Lett. 93 (2008),201903
 H. Y. Geng, Ying Chen, Y. Kaneta and M. Kinoshita, Phys. Rev. B77, 180101 (2008)
. H. Y. Geng, Ying Chen, Y. Kaneta and M. Kinoshita, Phys. Rev. B77, 104120 (2008)
10:00 AM - A3.4
Phase-Field Simulation of Intergranular Bubble Growth and Percolation.
Paul Millett 1 , Michael Tonks 1 , Bulent Biner 1 Show Abstract
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
The production of fission gas products, namely xenon and krypton, in irradiated nuclear fuel elements leads to a variety of phenomena that directly influence fuel performance. Central to the retention and release of fission gases is the evolution of bubbles existing on grain boundaries and grain triple junctions. Here, three-dimensional phase-field simulations of the growth and coalescence of intergranular Xe bubbles in UO2 bicrystal grain geometries will be presented. We investigate the dependency of bubble percolation on three factors: the initial bubble density, the Xe grain boundary diffusivity, and the bubble shape, which is governed by the ratio of the grain boundary energy over the surface energy. The simulations show that variations of each of these factors can lead to large discrepancies in the bubble coalescence rate, and eventual percolation, which may partially explain this observed occurrence in experimental investigations. This research was supported by the NEAMS program within DOE-NE.
10:15 AM - A3.5
Near-Surface Stoichiometry in UO2: A DFT Study.
Jianguo Yu 1 , Xian-Ming Bai 1 , Anter El-Azab 2 , Todd Allen 3 1 Show Abstract
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Florida State University, Tallahassee, Florida, United States, 3 , University of Wisconsin, Madison, Wisconsin, United States
Near-surface stoichiometry in uranium dioxide (UO2) is an important issue in microstructural evolution of nuclear fuel under radiation. The mechanisms of oxygen release and uptake by UO2 crystals containing high density of point defects are important for understanding the dynamics of defects and microstructure in these crystals. Few or no previous studies have been performed to understand the surface effects on stoichiometry. In this work, density functional theory (DFT) calculations and temperature-accelerated dynamics (TAD) and thermodynamic analysis are used to investigate the transition of oxygen from the surface into the bulk and vice versa. This investigation will enable the modeling of the oxygen kinetics in irradiated UO2 as a function of temperature and oxygen partial pressure. In an attempt to pin down the mechanisms of oxygen exchange between the gas environment and defected UO2, the results of this work will be compared to available experimental data. This work is supported by the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number FWP 1356.
10:30 AM - A3.6
DFT+U Investigation of Oxygen, Uranium and Xenon Transport in Uranium Dioxide Using Electronic Occupancy Control.
Marjorie Bertolus 1 , Boris Dorado 1 2 , David Andersson 2 , Philippe Garcia 1 , Michel Freyss 1 , Blas Uberuaga 2 , Christopher Stanek 2 Show Abstract
1 , CEA, DEN, Saint-Paul-lez-Durance France, 2 MST Division , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Uranium dioxide (UO2) attracts much interest due to its technological value as the standard nuclear fuel for pressurized water reactors. Although it has been extensively studied theoretically, its description by first principles remains challenging. The main difficulty lies in the description of the strong correlations between the 5f electrons of uranium atoms, which requires approximation beyond the standard density functional theory (DFT), such as the DFT+U. The use of the latter approximation, however, induces the presence of numerous metastable states which makes it difficult to converge to the ground state . Electronic occupancy control (EOC) has been extensively applied to UO2, as well as to other actinide bearing compounds, and has been shown to tackle effectively the problem of metastable states [2,3,4].We report here results of DFT+U calculations on oxygen, uranium and xenon transport in UO2 using EOC. We have investigated several elementary migration mechanisms and have calculated the associated migration energies. Results for oxygen transport have been compared to a comprehensive range of experimental data: interstitial migration and formation energies, vacancy migration, and Frenkel pair formation energies. A very encouraging correspondence is observed between experimentally determined formation and migration energies and those calculated using the DFT+U approximation with EOC . These results open up the prospect of using first-principles DFT+U calculations as part of a predictive approach to determining transport properties in other actinide compounds.ReferencesP. Larson, W. R. L. Lambrecht, A. Chantis, and M. van Schilfgaarde, Phys. Rev. B 75, 045114 (2007)B. Dorado, B. Amadon, M. Freyss, M. Bertolus, Phys. Rev. B 79, 235125 (2009)B. Amadon, F. Jollet, and M. Torrent, Phys. Rev. B 77, 155104 (2008)B. Dorado, M. Freyss, G. Jomard, M. Bertolus, Phys. Rev. B 82, 035114 (2010)B. Dorado, P. Garcia, M. Freyss, G. Carlot, M. Fraczkiewicz, B. Pasquet, G. Baldinozzi, D. Simeone, M. Freyss, M. Bertolus, Phys. Rev. B 83, 035126 (2011)
10:45 AM - A3.7
Simulations of Xe Redistribution in UO2: From Atomistics to Continuum.
David Andersson 1 , Blas Uberuaga 1 , Michael Tonks 2 , Paul Millett 2 , Chris Stanek 1 Show Abstract
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Redistribution of fission gases such as Xe is closely coupled to nuclear fuel performance. For example, the formation and retention of fission gas bubbles induce fuel swelling, which leads to mechanical interaction with the clad thereby increasing the probability for clad breach. Fission gas bubbles also decrease the thermal conductivity of the fuel. Alternatively, fission gas can be released from the fuel to the plenum, which increases the pressure on the clad walls. Most fission gases have low solubility in the fuel matrix, specifically the insolubility is most pronounced for large fission gas atoms such as Xe, and as a result there is a significant driving force for segregation of gas atoms to grain boundaries or dislocations and subsequently for nucleation of gas bubbles at these sinks. The first step of the fission gas redistribution is diffusion of individual gas atoms through the fuel matrix to existing sinks, which is governed by the activation energy for bulk diffusion. Fission gas bubbles are then formed by either separate nucleation events or by filling voids that were nucleated at a prior stage; in both cases the formation and later on growth are coupled to vacancy dynamics and thus linked to the production of vacancies via irradiation or thermal events. In order to better understand bulk Xe diffusion mechanisms in UO2±x we first calculate the relevant activation energies using density functional theory (DFT) techniques. By analyzing a combination of Xe solution thermodynamics, migration barriers and the interaction of dissolved Xe atoms with U vacancies, we demonstrate that Xe diffusion predominantly occurs via a vacancy-mediated mechanism, though other alternatives may exist for high irradiation fields. Since Xe transport is closely related to diffusion of U vacancies, we have also studied the activation energy for this process. Experimental data for the Xe and U activation energies are best reproduced if the active charge-compensation mechanism for intrinsic defects in UO2±x is considered. Due to the high thermodynamic cost of reducing U4+ ions, any defect formation occurring at a fixed composition, i.e. no change in UO2±x stoichiometry, always avoids such reactions, which, for example, implies that the ground-state configuration of an O Frenkel pair in UO2 does not involve any explicit local reduction (oxidation) of U ions at the O vacancy (interstitial). Next a continuum transport model for Xe and U is formulated based on the diffusion mechanisms established from DFT. After combining this model with descriptions of the interaction between Xe and UO2 grain boundaries derived from separate atomistic calculations, we simulate Xe redistribution for a few simple microstructures using finite element methods (FEM), as implemented in the MOOSE framework from Idaho National Laboratory.
11:30 AM - A3.8
On the Deformation of UO2: A Molecular Dynamics Study.
Paul Fossati 1 , Remi Dingreville 3 4 , Laurent Van Brutzel 1 , Jean-Paul Crocombette 2 , Timothy Bartel 4 Show Abstract
1 DEN/DPC/SCP, CEA Saclay, Gif-sur-Yvette France, 3 Department of Mechanical and Aerospace Engineering, Polytechnic Institute of New York University, Brooklym, New York, United States, 4 Advanced Nuclear Fuel Cycle Tech., Sandia National Laboratories, Albuquerque, New Mexico, United States, 2 DEN/DMN/SRMP, CEA Saclay, Gif-sur-Yvette France
Exemplified by the recent events in Fukushima Japan, increasing concerns about energy security and the environmental impact of energy use have led to intensive interest in nuclear power. In order to safely extend the lifetime of existing nuclear reactors and develop fuels for the next generation of reactors, we need have a fundamental understanding of the mechanical behavior of reactor fuel, uranium dioxide (UO2). Although UO2 pellets are widely used as nuclear fuel, uncertainty still remains about their thermo-mechanical behavior. Most of our knowledge regarding UO2 mechanical behavior is obtained by experiments on unirradiated fuel, or post-mortem analysis on spent fuel. Atomistic models give us a good grasp on what is the behavior of the fuel in conditions inaccessible to current experiments, and the related effects at a larger length-scale.The present investigation considers recent studies on the mechanical properties of UO2 by means of atomistic simulations using empirical potentials. For this study four different rigid ion potentials have been assessed. UO2 elasticity and plastic behavior are of primary concerns in this study. Firstly, the elastic constants of different polymorphs (fluorite, rutile, α-PbO2) have been studied for a wide range of temperatures. Temperature and orientation dependence will be discussed. Secondly, investigations and reflection on the stability and motion mechanisms of dislocations in UO2 will be discussed. Finally, crack initiation and propagation in UO2 single-crystal will be considered in light of the aforementioned results. All of these studies constitute a coherent set of parameters that can be used in mesoscale and continuum models to investigate creep and plastic deformation of nuclear fuel.
11:45 AM - A3.9
Temperature Accelerated Dynamics Simulations of Defect Clustering in UO2.
Xian-Ming Bai 1 , Jianguo Yu 1 , Anter El-Azab 2 , Todd Allen 3 1 Show Abstract
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Florida State University, Tallahassee, Florida, United States, 3 , University of Wisconsin, Madison, Wisconsin, United States
The aggregation of radiation-induced point defects in uranium dioxide (UO2) is a critical step in microstructural evolution, and consequently can have significant effects on material properties such as thermal and mass transport and mechanical properties. Therefore, understanding the kinetic evolution of point defects in UO2 is important for predicting nuclear fuel performance. Here we use temperature accelerated dynamics (TAD) simulations with the Basak potential to investigate the elementary atomistic mechanisms involved in defect clustering processes across different timescales at both 300 K and 1000 K. We investigate the binding and migration energies of different cluster sizes and configurations. Several types of defect clusters are observed including the cuboctahedral oxygen interstitial cluster. The comparison between our TAD simulation results with density functional theory calculations is also discussed. These results give new insight into the initial stages of microstructural evolution necessary for predicting fuel performance. This work is supported by the Center for Materials Science of Nuclear Fuel, an Energy Frontier Research Center (EFRC) funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number FWP 1356.
12:00 PM - **A3.10
Properties of Vacancy Defects Induced in UO2 by Irradiation and Probed by Using Positron Annihilation Spectroscopy.
Marie-France Barthe 1 , Tayeb Belhabib 1 , Stéphanie Leclerc 1 , Laszlo Liszkay 1 , Hicham Labrim 1 , Virginie Moineau 1 , Pierre Desgardin 1 , Gaelle Carlot 2 , Philippe Garcia 2 Show Abstract
1 CEMHTI, CNRS, Orleans France, 2 DEN/DEC/SESC, CEA, Saint Paul lez Durance France
The understanding of the behavior of fission nuclear fuel under irradiation is of first importance to foresee the state of the fuel in reactors and also its evolution after use in storage conditions. The study of this behavior can be carried out by the characterization of the in pile irradiated fuel. This work requires heavy installations (hotlabs..) and long term experiments (decay …) and doesn’t always allow to discriminate between interdependent phenomena which can occur in the material. These studies have to be completed with separate effects experiments associated with multi-scale modeling. This approach allows a more detailed knowledge of the different phenomena and their interconnections. The complete modeling of the behavior of material under irradiation begins with the first stages of damage and requires the detailed knowledge of fundamental data especially concerning the point defects properties. These data can be obtained by calculations and/or from experiments performed in separate effects conditions. Uranium dioxide, as the major component, is used as the model material of the nuclear fission fuel . The damage induced in UO2 by irradiation has been extensively studied by using different techniques such as Channeling Rutherford Backscattering, RX diffraction, Transmission Electron Microscopy and so on. Very few studies have been focused on the direct observation of the point defects and the determination of their properties in the UO2 matrix. In this work, we have used positron annihilation spectroscopy (PAS) to characterize the vacancy defects induced in UO2 by irradiation in different conditions (different particles, energies and fluencies). Both 22Na based Positron Annihilation Lifetime Spectroscopy (PALS) and coincidence Doppler annihilation-ray Broadening Spectrometry (CDBS) and Slow Positron Beam coupled with Doppler Broadening Spectrometry (SPBDBS) have been used to probe these vacancy defects and study their evolution as a function of temperature. In some cases it can be very useful to follow the behavior of gas such as Helium by using Nuclear Reaction Analysis and to identify the interaction between both identities He and vacancy. This work was funded in part by F-Bridge Project which is part of the European Union’s Framework Programme 7 and by the National Research Group MATINEX of the French PACEN Programme.
12:30 PM - **A3.11
Atomistic Simulation of Nuclear Fuels.
Matthias Krack 1 Show Abstract
1 , Paul Scherrer Institute, Villigen PSI Switzerland
The experimental investigation of actinide materials like nuclear fuels is difficult and usually very costly. Therefore a reliable multi-scale modeling of these often hazardous materials starting at the atomistic level is inevitable to gain further insight into this type of materials. The development of new, more advanced simulation methods accompanied by the rapid growth of the available computational resources provided by high-performance computing facilities, allows the modeling of such materials at a new quality level. Also the recent development of the CP2K program package (http://cp2k.berlios.de) has been partially focused on enabling state-of-the-art simulations of actinide materials using classical potential as well as electronic structure methods. The goal is to perform reliable molecular dynamics simulations for actinide materials including advanced simulation techniques like metadynamics. Metadynamics is an accelerated molecular dynamics method allowing for the fast exploration of a system's energy landscape, even if energy barriers are present which are large compared to the typical thermal fluctuations. In this way, rare events can be observed within the time periods accessible by standard molecular dynamics simulation runs. The CP2K program package and some of its first applications to actinide materials, especially uranium dioxide, are presented.
A4: Radiation Damage - Metals
Tuesday PM, November 29, 2011
Independence W (Sheraton)
2:45 PM - A4.2
Irradiation-Induced Creep in Dilute Nanostructured Cu-W Alloys.
Kaiping Tai 1 , Yinon Ashkenazy 2 , Robert S. Averback 1 , Pascal Bellon 1 Show Abstract
1 Department of Materials Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois, United States, 2 Racah Institute of Physics, Hebrew University of Jerusalem, Jerusalem Israel
The development of new radiation resistant materials focuses largely on creating high densities of neutral sinks for point defects, such as nanosized inclusions and grain boundaries. While this strategy is appears useful for eliminating point defects and/or trapping He gas atoms, very little is known about the creep properties of these materials under the extreme conditions of an irradiation environment. We report here, new in situ creep measurements on 1.8 MeV Kr+ irradiated Cu and dilute Cu-W nanostructured alloys films as a function of temperature using plane-strain bulge testing. Primary and secondary creeps were observed at all temperatures. The secondary creep rates in the Cu-W alloys were observed to increase with increasing temperature between 300 K and 473 K, and then become constant up to 573K. An activation enthalpy of 0.30±0.05 eV was obtained for Cu93.5W6.5 and Cu99W1 alloys. Subsequent (scanning) transmission electron microscopy analysis revealed a high density of small (2-3 nm) W-rich nanoparticles in these irradiated samples. The precipitates had a BCC structure and were (semi) coherent with the Cu matrix. The alignment of specific crystallographic planes in the W and Cu follow the K-S or N-W orientations. Analysis of kinetic behavior shows that the irradiation-enhanced creep in these materials derives from neutral point defect fluxes to the grain boundaries. A new model of radiation-enhanced creep in ultrafine grained metals, based on molecular dynamics simulations, is presented; it explains how neutral fluxes of point defects can lead to creep deformation.
3:00 PM - A4.3
An Atomic-Level Perspective into the Evolution of Interfaces in Purposed Radiation Tolerant Materials.
Jeffery Aguiar 1 2 , Nigel Browning 1 2 5 , Luke Hsiung 2 , Michael Fluss 2 , Peter Hosemann 3 4 , Sanchita Dey 1 Show Abstract
1 Chemical Engineering and Material Science Department, University of California Davis, Davis, California, United States, 2 Condensed Matter and Materials Division, Lawrence Livermore National Laboratory, Livermore, California, United States, 5 Molecular Cell Biology, University of California Davis, Davis, California, United States, 3 Nuclear Engineering, University of California Berkeley, Berkeley, California, United States, 4 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
In an attempt to address the energy crisis, next generation nuclear technology poses as a serious candidate. Conditionally next generation nuclear power also presents serious issues related to reactor safety and lifetimes due to unknown longevity of reactor containment materials due to heightened reactor environments. In order to address these concerns relating to containment materials, several materials issues are currently and remain to be investigated, ranging from corrosion to synergistic radiation induced effects in a series of purposed damage resistant materials, including oxide dispersion strengthened (ODS) alloys. To address the pressing need for analytical characterization of proposed radiation tolerant materials, we have used the latest the technologically advanced techniques in simultaneous imaging and spectral analysis using aberration corrected (scanning) transmission electron microscopy (S)TEM coupled with high resolution electron energy loss spectroscopy (HR-EELS) and energy dispersive x-ray (EDX) analysis to study a series of ODS alloys. In our work we are focused on the use the use of aberration corrected microscopy and spectroscopy to characterize three purposed radiation tolerant materials, ODS alloys MA-957, K3, and PM-2000, with and without the effects of radiation damage at particle-matrix interfaces. to evaluate for structural and chemical changes, such as radiation induced segregation and swelling. The use of oxide dispersed constituents increases the likelihood of interfacial defect trapping before reaching a grain boundary, but the same interfaces over the lifetime of radiation evolve with the inclusion of trapped defects, complexes, and helium bubbles which requires the uttermost best spatial and energy resolution to decouple. The progression of these atomic-level interfaces in ODS alloys has thereby been studied in great detail using aberration corrected (S)TEM and energy filtered spectral analysis, EDX and EELS, to fundamentally develop an atomic level perspective into the structural and chemical evolution of these three candidate ODS alloys. The talk will therefore focus on the principle capabilities and use of (S)TEM and EELS to study and develop atomic level perspective into the evolution of interfaces in ODS alloys with the uttermost highest achievable spatial and energy resolution.
3:15 PM - A4.4
Cavity Formation in Multi-Ion-Beam Irradiated ODS Ferritic Steel.
Luke Hsiung 1 , Scott Tumey 1 , Michael Fluss 1 Show Abstract
1 Physical and Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States
One of the major challenges in designing fusion reactors is to develop the high performance structural materials for first wall and breeding-blanket components, which will be exposed to high fluxes of high-energy (14 MeV) neutrons from the deuterium-tritium fusion and helium and hydrogen from (n, α)- and (n, p)-transmutation reactions. Although significant progress has been made recently to understand the processing-microstructure-property relationships of ferritic and martensitic (F/M) steels and oxide dispersion strengthened (ODS) F/M steels, it remains to understand the role of helium and hydrogen transmutation gases on the cavitational swelling of F/M and ODS F/M steels. Since no prototype fusion reactors currently exist, it is difficult to directly evaluate the high-energy neutron damage environment expected to prevail in the first wall of a fusion reactor. One technique commonly used to study the evolution of defect structures and the nucleation and growth of voids utilizes transmission electron microscopy (TEM) examinations of specimens simultaneously bombarded by heavy ions and helium and/or hydrogen ions through so called "dual-beam" and "triple-beam" experiments. The heavy ions create atomic displacements while the gas ions lead to the effects of the transmutation gases, helium (10 appm/dpa) and hydrogen (40 appm/dpa). We have recently conducted HRTEM studies to compare radiation effects on Fe-14Cr alloy and Fe-16Cr ODS steel using (He + Fe) dual-beam and (H + He + Fe) triple-beam techniques. Important results will be presented to address the effects of nanoparticles on the suppression of radiation-induced cavitational swelling. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
3:30 PM - **A4.5
In Situ TEM Studies of Microstructure Evolution under Ion Irradiation for Nuclear Engineering Applications.
Djamel Kaoumi 1 , Jimmy Adamson 1 , Arthur Motta 2 , Brian Wirth 3 , Aaron Kohnert 3 , Mark Kirk 4 Show Abstract
1 Mechanical and Nuclear Engineering, University of South Carolina, Columbia, South Carolina, United States, 2 Mechanical and Nuclear Engineering, Penn State , University Park, Pennsylvania, United States, 3 Nuclear Enggineering, University of Tennessee, Knoxville, Tennessee, United States, 4 Materials Science Division, Argonne National Laboratories, Argonne, Illinois, United States
One of the difficulties of studying processes occurring under irradiation (in a reactor environment) is the lack of kinetics information since usually samples are examined ex situ (i.e. after irradiation) so that only snapshots of the process are available. Given the dynamic nature of the phenomena, direct in situ observation is invaluable for better understanding the mechanisms, kinetics and driving forces of the processes involved. This can be done using in situ ion irradiation in a TEM at the IVEM facility at Argonne National Laboratory. To predict the in reactor behavior of alloys, it is essential to understand the basic mechanisms of radiation damage formation (loop density, defect interactions) and accumulation (loop evolution, precipitation or dissolution of second phases…). In-situ ion-irradiation in a TEM has proven a very good tool for that purpose as it allows for the direct determination of the formation and evolution of irradiation-induced damage and the spatial correlation of the defect structures with the pre-existing microstructure (including lath boundaries, network dislocations and carbides) as a function of dose, dose rate, temperature and ion type Using this technique, different aspects of microstructure evolution under irradiation were studied, such as defect cluster formation and evolution as a function of dose in advanced Ferritic/Martensitic (F/M) steels, the irradiation stability of precipitates in Oxide Dispersion Strengthened (ODS) steels, and irradiation-induced grain-growth. In this paper we will emphasize the work done on model F/M steels (Fe12Cr0.1C and Fe-9Cr-0.1C) which were irradiated with 1 MeV Kr ions at 50K, 180K, 298K, 473K, 573K to doses up to 10 dpa in-situ in a TEM. The microstructure evolution under irradiation was followed and characterized at successive doses in terms of defect formation and evolution, black dot density, and stability of as-fabricated microstructure using weak-beam dark-field imaging and g.b analysis for comparison with computations made using a spatially dependent rate theory model of cluster evolution in both compositional and geometric spaces under conditions of high energy ion irradiation. More specifically, in the model the concentrations of interstitial loops and voids are calculated as a function of time, number of interstitials/vacancies, spatial position, dislocation densities, temperature, dose and dose rate, impurities and so on. This presentation will focus on the experimental in-situ TEM observations.
A5: Metallic Systems - Modelling
Tuesday PM, November 29, 2011
Independence W (Sheraton)
4:30 PM - A5.1
Modelling Radiation Effects in ODS Steels.
Roger Smith 1 , Tomas Lazauskas 1 , Steven Kenny 1 Show Abstract
1 , Loughborough University, Loughborough, Leicestershire, United Kingdom
ODS materials are promising candidates for use in both fission and fusion reactors. It has been suggested that such materials can be more radiation resistant and stronger than other steels and can minimise the effect of inert gas bubble accumulation. Here we present the first results arising from a UK-India nuclear collaboration project of the simulation of radiation in these materials. Yttria particles are embedded in a bcc Fe matrix with a size distribution corresponding to those that are experimentally observed in a typical ODS material. The system is modelled using a fixed charge potential for the Y-O interactions and an embedded atom type potential for the Fe-Fe interactions. Collision cascades at various energies are initiated in the Fe matrix and the effect of the embedded nanoparticles on the cascade development is reported.It is shown that under certain circumstance the nanoparticles deflect that moving Fe atoms with a tendency to trap defects at the interface between the nanoparticles and the matrix.
4:45 PM - A5.2
Structure and Properties of the Y2O3/ Fe Interface from First Principles Calculations.
Samrat Choudhury 1 , Christopher Stanek 1 , Blas Uberuaga 1 Show Abstract
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Nanostructured ferritic alloys (NFAs) (with a typical composition of 12–14 wt% Cr, 0.25 wt% Y and 0.5 wt% Ti) are considered excellent candidate materials for structural applications in nuclear reactors as they exhibit exceptionally high creep strength and radiation tolerance due to the presence of highly stable nanometer sized Y-Ti-O precipitates (NPs). It is believed that these properties result from the characteristics of the particle and ferritic matrix interface. Y2O3 has also been shown to form nanoprecipitates in Fe and is a simpler surrogate for the Y-Ti-O precipitates. In this work, we will present the behavior of the interface between the ferritic matrix and Y2O3 using density functional theory. In particular, the role of alloying elements and orientation relationship on the atomic structure of the particle-matrix interface, segregation energies of the alloying elements, electronic structure at the interface, and calculated interfacial energy will be discussed. These results form the basis of a phase-field model that will examine the nucleation and growth of Y2O3 precipitates in Fe.
5:00 PM - A5.3
Evolutions of Oxide Particles and Grain Morphology of 12 Cr ODS Steel.
Jinsung Jang 1 , Tae Kyu Kim 1 , Xiaodong Mao 1 2 , Chang Hee Han 1 , Young Soo Han 1 , Kyu Hwan Oh 2 Show Abstract
1 Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon Korea (the Republic of), 2 Dept. of Materials Science & Engineering, Seoul National University, Seoul Korea (the Republic of)
Oxide dispersion strengthened (ODS) steel is one of good candidate materials for in-core components of Generation IV nuclear systems due to its good high temperature mechanical strenghth as well as the excellent neutron radiation resistance. 12Cr ODS steel samples were prepared by mechacanical alloying(MA) of the elmental metal powders along with 20-30 nm yttria (Y2O3) particels as the strengthening dispersoids. MA powders were reduced by hydrogen mixture gas during degassing process, and then consolidated by hot isostatic pressing(HIP) and hot rolling(HR).Evolutions of yttrium containing oxide particles such as YTaO4 or Y3TaO7 as well as the grain morphology after each step are investigted using SEM/EBSD(Scanning Electron Microscopy /Electron Backscattered Diffraction), TEM(Transmission Electron Microscopy), and the particle size distributions are estimated by SANS(Small Angle Neutron Scattering) and compared with those by other analytical techniqes.
5:15 PM - A5.4
An Atom Probe Study of Radiation-Induced Segregation/Depletion in a Fe-14.25wt%Cr Ferritic Steel.
Rong Hu 1 , George Smith 1 , Emmanuelle Marquis 2 Show Abstract
1 Department of Materials, University of Oxford, Oxford United Kingdom, 2 Department of Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Ferritic chromium steels are important structural materials for future nuclear fission and fusion reactors due to their advantages over traditional austenitic steels, including low swelling rates, better thermal fatigue resistance, and lower thermal expansion coefficients. Radiation-induced segregation or depletion (RIS/RID) of solute atoms at grain boundaries is considered to be a potentially significant phenomenon for structural materials because of its potentially detrimental role in affecting microstructure and furthermore mechanical properties. However, the behaviour of Cr at grain boundaries in ferritic steels is not well understood. Both segregation and depletion of Cr at grain boundary under irradiation have been previously observed and no clear dependency on irradiation condition or alloy type has been presented. To understand the Cr behaviour at grain boundaries in ferritic steels under irradiation, a systematic approach combining EBSD, FIB specimen preparation and atom probe tomography analysis has been applied on a Fe-14.25wt%Cr to investigate the effect of grain boundary orientation, irradiation depth, impurities and other factors to get a better understanding of RIS/RID phenomenon. Both low sigma boundaries and high sigma boundaries have been investigated in detail and systematic differences between the behaviour of different classes of boundaries will be reported.
5:30 PM - A5.5
Calculation of Reaction Constants for Vacancy Migration in α-Iron Using Transition Path Sampling with Lyapunov Bias.
Massimiliano Picciani 1 , Manuel Athenes 1 , Mihai-Cosmin Marinica 1 Show Abstract
1 DEN/DMN/SRMP, CEA Saclay, Gif-sur-Yvette France
Predicting the microstructural evolution of radiation damage in materials requires handling the physics of infrequent-events, in which several time scales are involved. The reactions rates which characterize those events are the main ingredient for simulating the kinetics of materials under irradiation over large time scales and high irradiation doses. Here we propose an efficient, finite temperature method to compute reaction rate constants of thermally activated processes.
The method consists of two steps. Firstly, rare reactive trajectories in phase-space are sampled using a Transition Path Sampling algorithm supplemented with a Lyapunov bias favoring diverging trajectories. This enables the system to visit transition regions separating stable configurations more often, and thus enhances the probability of observing transitions between stable states during relatively short simulations. Secondly, reaction constants are estimated from the unbiased fraction of reactive trajectories , yielded by an appropriate statistical data analysis tool, the Multistate Bennett Acceptance Ratio package.
We apply our method to the calculation of reaction rates for the migration of vacancies and divacancies in an α-Iron crystal, testing different Embedded Atom Model potentials, for temperatures ranging from 250 K to 900 K. Vacancy diffusion rates associated with activation barriers at finite temperature are then evaluated, showing a significant difference from values obtained using the standard harmonic approximation. Finally, the calculated diffusion constants are employed as input parameters in a first passage kinetic Monte Carlo (FPKMC) code in order to model the migration and clustering of defects in resistivity recovery experiments.
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5:45 PM - A5.6
Effects of Li on Zirconium Alloy Corrosion – Li Insertion, and Ion Migration in ZrO2.
Mostafa Youssef 1 , Bilge Yildiz 1 Show Abstract
1 Nuclear Science and Engineering Department, MIT, Cambridge, Massachusetts, United States
It is known that the corrosion resistance of the zirconium alloys is greatly diminished when there is a high concentration of Li in water. Accelerated corrosion of zirconium alloys pose safety and operational challenges as they serve as nuclear fuel cladding. The mechanisms and kinetics that lead to Li incorporation into the zirconium oxide, the passive layer on zirconium metal in water corrosion, and the corresponding acceleration of corrosion are described only macroscopically and empirically. An understanding of the structure and the corrosion characteristics of lithiated zirconium oxide at a fundamental level would allow for the predictive assessment of corrosion kinetics, and strategies against Li trapping. In this study, we investigate the structure of zirconium oxide with Li insertion at interstitial and at substitutional sites, identify the diffusion barriers of Li and of oxygen in the lithiated structure, and the relative stabilities of the monoclinic and tetragonal zirconia phases when Li is inserted, all of which affect the protective characteristics of the zirconium oxide, and thus, the corrosion kinetics. We use first principles methods based on density functional theory calculations in our analysis. Our initial results show that the transition from a protective tetragonal phase to a less protective monoclinic phase is not driven by the presence of interstitial Li atoms. Correlations of Li presence and ion transport barriers are assessed and discussed in connection to corrosion kinetics.