Symposium Organizers
Gianguido Baldinozzi CEA-CNRS-ECP
Yanwen Zhang Pacific Northwest National Laboratory
Katherine L. Smith Embassy of Australia
Kazuhiro Yasuda Kyushu University
V1: Radiation Effects
Session Chairs
Monday PM, November 30, 2009
Room 207 (Hynes)
9:30 AM - **V1.1
Effects of Ionization on Irradiation Damage Evolution and Thermal Recovery in Ceramics.
William Weber 1 , Yanwen Zhang 2 , Ram Devanathan 1
1 Fundamental & Computational Sciences Directorate, Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Environmental Molecular Sciences Laboratory, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractIrradiation with energetic electrons and ions results in the transfer of energy to both atomic nuclei and the electronic structure. Kinetic energy transfer to atomic nuclei results in energetic atomic displacements and the production of atomic-level defects, while ionization energy loss to the electronic structure generates electron-hole pairs and localized electronic excitations. The understanding and modeling of atomic collision cascades and their role in irradiation damage evolution is well advanced. The effects of ionization are less understood. In ceramics, the localized electronic excitations can result in localized charge at defects and interfaces, rupture or change in nature of covalent and ionic bonds, enhanced defect and atomic diffusion, and changes in phase transformation dynamics, which affect the dynamics of atomic processes and the interpretation of the results from ion and electron irradiation experiments. Under irradiation with different ions, the ratio of electronic to nuclear stopping powers varies locally for both the primary ion and the secondary recoils produced. It will be shown that the critical temperature for ion-beam induced amorphization can exhibit a strong dependence on the ratio of electronic to nuclear stopping, which demonstrates that the local rate of in-cascade ionization has a significant effect on the dynamic recovery processes that determine the critical temperatures. Simultaneous electron and ion irradiation are shown to dramatically affect the dynamics of damage accumulation. In post-irradiation studies of ion-irradiated materials, ionization-enhanced recovery and recrystallization due to electron beam irradiation are observed, and the kinetics of the enhanced recovery processes has been determined. In the case of high-energy heavy ions (~0.1 to 2 GeV), such as fission fragments or swift-heavy ions, the intense energy deposition into the electronic structure produces a thermal spike. Computer simulations of thermal spikes in a range of materials demonstrate that the damage produced can range from the production of isolated point defects and defect clusters to the formation of tracks with fine structure.
10:00 AM - V1.2
The Need for Quantum Mechanics in Large-scale Atomistic Simulations of Radiation Damage in Metals.
C. Race 1 , D. Mason 1 , M. Finnis 1 2 , W. Foulkes 1 , A. Horsfield 2 , T. Todorov 3 , A. Sutton 1
1 Department of Physics, Imperial College London, London United Kingdom, 2 Department of Materials, Imperial College London, London United Kingdom, 3 School of Mathematics and Physics, Queen’s University Belfast, Belfast United Kingdom
Show AbstractIt has long been recognised that electronic excitations caused by high velocity particles in metals are central to understanding how these particles are slowed down. Quantum mechanics has played a key role in modelling such processes in idealized free electron gases (jellium models). The imperative now is to develop quantum mechanical treatments of metals with real atomic structures for large-scale atomistic simulation of radiation damage. In this paper we present an example of such large-scale simulation applied to the phenomenon of channelling.When a particle with a high kinetic energy enters a crystalline solid it may travel large distances along channels in the crystal structure. This process is called channelling. It plays a central role in determining the depth of irradiation damage suffered by materials exposed to high energy incident particles in nuclear reactors and in ion implantation. A key question centres on the mechanisms by which such a high energy particle loses its energy as it rattles down a channel in the crystal. It is known that at very high energies the principal mechanism is electronic, that is the channelling particle creates electronic excitations and gradually loses its energy until it has slowed sufficiently to create a cascade of atomic displacements. We present a simulation of this process based on solving the time-dependent Schrodinger equation for the electrons in a crystal as an interstitial particle of high kinetic energy channels through it. Unlike many previous simulations we consider the real atomic structure of the metal, and not a free electron gas. We find good agreement with previous models predicting a stopping force linear in projectile velocity. We also find a new mechanism of electronic excitation arising from the discrete atomic structure of the metal. This mechanism is absent in the earlier free electron models and results in a resonance in the ion charge at low channelling velocity.
V2: Complex Microstructures
Session Chairs
Monday PM, November 30, 2009
Room 207 (Hynes)
11:15 AM - **V2.1
Can We Describe Phase Transition under Irradiation in Insulators within the Random Phase Approximation Framework?
David Simeone 1 2 , Gianguido Baldinozzi 2 1 , Dominique Gosset 1 2 , Laurence Luneville 1 2
1 CEA/DEN/DANS/DMN/SRMA/LA2M-MFE, CEA, Gif sur yvette France, 2 CNRS-ECP/SPMS-MFE UMR 8580, CNRS-ECP, Chatenay Malabry France
Show AbstractThe renewed interest in nuclear energy production and the environmental impact of energy are bringing about a renaissance in materials sciences. The compelling need for valid predictive models and accurate data are needed to forecast the radiation effects and long-term degradation of reactor components and radioactive waste hosts are expected to become increasingly critical over the next decade. The radiation tolerance of insulating ceramics for fusion energy systems and of nuclear fuel for fission systems is also a matter of great concern. The Random Phase Approximation seems to give a valuable framework to understand microstructural transformations induced by radiation damages in metals and alloys. Based on experimental evidences, the aim of this talk is to analyze phase transitions triggered by irradiation damages in two different oxides, pure zirconia and magnesium spinels, within this framework pointing out limitations of this approach.
11:45 AM - V2.2
Phase-Field Simulation of Void and Fission-Gas Bubble Evolution in Irradiated Polycrystalline Materials.
Paul Millett 1 , Anter El-Azab 2 , Michael Tonks 1 , Srujan Rokkam 2 , Dieter Wolf 1
1 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho Falls, ID 83415, Idaho, United States, 2 Scientific Computing, Florida State University, Tallahassee, FL 32310, Florida, United States
Show AbstractThe interactive evolution of both polycrystalline microstructure and irradiation-induced defects such as voids and fission gas-filled bubbles in nuclear fuels and structural alloys is complex and critically important to the long-term performance of fission reactors. Here, the phase-field technique is used to model the evolution of multiple point-defect species (vacancies, self-interstitials, and gas atoms), generated randomly in space and time to represent collision cascade events, thus allowing spatially-resolved simulations of void and gas bubble nucleation and growth both within grain interiors and at grain boundary interfaces (which are shown to be heterogeneous nucleation sites). Illustrative results including the formation of void denuded zones and void peak zones adjacent to grain boundaries, the interlinkage of intergranular gas bubbles leading to fission gas release, and the effects of temperature and stress gradients will be presented. This work was supported by the DOE-BES Computational Materials Science Network (CMSN).
12:00 PM - V2.3
Phase Field Modeling of Void Nucleation and Growth in Irradiated Metals.
Srujan Rokkam 1 , Santosh Dubey 2 , Anter El-Azab 2 , Paul Millett 3 , Dieter Wolf 3
1 Department of Mechanical Engineering, Florida State University, Tallahassee, Florida, United States, 2 Department of Scientific Computing, Florida State University, Tallahassee, Florida, United States, 3 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractIrradiation of materials by energetic particle (e.g., neutrons in nuclear reactors) is accompanied by excessive point defect generation by atomic collision cascades. The diffusion and interaction of these point defects with each other and with pre-existing defects results in microstructure evolution. An important aspect of this evolution is the nucleation and growth of voids, which causes swelling and dimensional instabilities which are detrimental to the structure. Here, we present a phase field model for void nucleation and growth in irradiated metals. The formalism developed herein thus treats both the nucleation and growth processes simultaneously in a spatially resolved fashion. The material is described in terms of free energy functional obtained from the enthalpic and entropic (configurational and vibrational) contributions. Point defect fluxes and defect densities are obtained using a Cahn-Hilliard type description for the vacancy and interstitial concentration fields. The dynamics of void growth are obtained in terms of the evolution of a non-conserved order parameter field, whose evolution is prescribed by a phenomenological Allen-Cahn type equation. Using the case of pure metals as an example, we illustrate model capabilities with regards to void nucleation and growth in the presence of interacting point-defects, and defects interacting with lattice sinks. The effects of vibrational entropy on the defect dynamics and void evolution are investigated. In addition, void nucleation is studied as a function of thermal fluctuations and cascade damage. Furthermore, we use the concept of stochastic point process in space and time to model the generation of point defects due to cascades. Finally, the effect of spatially resolved point defect sinks (such as dislocations) on void nucleation and growth is investigated.This work was supported by DOE-BES Computational Materials Science Network(CMSN)
12:15 PM - V2.4
HRTEM Studies of Nano-Particles in an ODS Steel.
Luke Hsiung 1 , Jeffery Aguiar 1 , Nigel Browning 1 , Michael Fluss 1 , Akihiko Kimura 2
1 Physical and Life Sciences, Lawrence Livermore National Laboratory, Livermore, California, United States, 2 Institute of Advanced Energy, Kyoto University, Kyoto Japan
Show AbstractMany key issues remain unsolved for developing ODS steels for fission and fusion applications including incomplete understanding of the effect of irradiation on low-temperature fracture properties, the role of fusion relevant helium and hydrogen transmutation gases on the deformation and fracture of irradiated material at low and high temperatures, and mechanisms of swelling suppression in ODS steels. In preparation for ion-beam experiments, we are currently performing HRTEM and STEM studies of a 16Cr-5Al-2W-0.3Ti-0.4Y2O3 ODS steel with an emphasis on the crystal and interfacial structures of the nanoscale oxide particles and their coherency with respect to the Fe (Cr) matrix. We will report on some of the studies and will address the critical features which may illuminate the influence of thermodynamics and kinetics on the growth and refinement of the nano-particles. We will also point to those features that may be of interest with respect to the suppression of radiation-induced dimensional changes due possibly to the nano-dispersoids. This work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DEAC5207NA27344.
12:30 PM - **V2.5
Atomic-scale Analysis of Irradiation-induced Structural Change in Magnesium Aluminate Spinel Compound.
Syo Matsumura 1 , Tomokazu Yamamoto 1 , Kazuhiro Yasuda 1
1 Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Fukuoka Japan
Show AbstractThe present talk will give an overview of our recent results on irradiation-induced structural change in magnesium aluminate spinel compound, which is known as a radiation tolerant oxide, especially to volumetric swelling. Magnesium aluminate spinel of MgO-nAl2O3 with n=1.1, was irradiated with swift heavy ions of 200 MeV Xe14+ (Se=24 keV/nm) and 350 MeV Au28+ (Se=34 keV/nm) at a Tandem ion-accelerator. Transmission electron microscopy techniques of high-resolution (HR) imaging, STEM dark-field imaging as well as high angular resolution electron channeling x-ray spectroscopy (HARECXS) were employed in quantitative analysis of irradiation-induced structural change. Dark spotty contrast appears at ion-tracks formed by swift-heavy irradiation in STEM dark-field imaging, indicating lower density inside the ion-tracks. Clear lattice fringes are observed in HR images even inside the ion tracks in both Xe14+ and Au irradiated specimens. However, the fringe pattern inside the tracks is different from that appearing in the matrix, being indicative of formation of a defective NaCl structure. Molecular dynamics (MD) simulations have shown that the spinel structure becomes unstable by accumulation of displaced interstitials and a defective NaCl structure is formed after preferential evacuation of cations from the tetrahedral positions. Quantitative HARECXS analysis showed that cation disordering progresses successively with ion fluence. It was revealed that the disordered regions are extended over about 12 nm in diameter along the ion-tracks, which is much wider than the defective volume detected by HR images. The present study was supported in part by Grant-in-Aid for Scientific Research (A) (#18206068) and for the Junior Scientist from JSPS.
V3: Metallic Materials I
Session Chairs
Monday PM, November 30, 2009
Room 207 (Hynes)
3:00 PM - V3.2
Irradiation-Induced Point Defects in Nanocrystalline Molybdenum by Molecular-Dynamics Simulation.
Dilpuneet Aidhy 1 , Paul Millett 2 , Simon Phillpot 3 , Alex Chernatynskiy 3 , Dieter Wolf 2
1 Materials Science and Engineering, Northwestern University, Evanston, Illinois, United States, 2 Nuclear Fuels and Materials, Idaho National Laboratory, Idaho Falls, ID 83415, Idaho, United States, 3 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractEvolution of irradiation-induced point defects in the presence of grain boundaries (GBs) is studied in bcc Molybdenum (Mo) using molecular dynamics (MD) simulation. Point defects created due to the radiation events can annihilate primarily by two mechanisms: mutual recombination of interstitials and vacancies (bulk), and by elimination at the GBs. By calculating the source/sink strength of the GBs, in accord with the rate-theory model, the dominant point-defect annihilation mechanism is predicted. At high temperatures, their high diffusivity leads to mutual annihilation in the bulk. In contrast, at low temperatures, because they less often recombine in the bulk, they annihilate predominantly at the GBs. It is further found that the defect concentration also dictates the annihilation mechanism. At low concentrations annihilation takes place at GBs, while conversely, at high concentrations annihilation takes place in the bulk. Finally, the annihilation mechanism also depends upon the grain size, with GB mechanism prevalent at smaller grain sizes. As the grain size increases, a crossover between the two mechanisms is observed. At ~ 300 K, the critical grain size is tens of nanometer, indicating that the GB mechanism dominates at nanometer grain-size materials. This work was funded by DOE-NERI Awards DE-FC07-07ID14833, and by the DOE-BES Computational Materials Science Network (CMSN).
3:15 PM - V3.3
TEM Characterization of Neutron- and Ion-irradiated Nano-structured Ferritic Alloys.
James Bentley 1 , D. Hoelzer 1
1 Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractMechanically alloyed (MA) nano-structured ferritic alloys (NFA) have the potential to be highly resistant to radiation damage in fission and fusion environments. High concentrations (>1023 m-3) of small (<5 nm) Ti-Y-O nano-clusters (NC) not only result in outstanding mechanical properties, but also are expected to promote point-defect recombination and trap transmutation-produced He in small bubbles. Early results of microstructural characterization of NFA irradiated with neutrons and ions are encouraging: several publications indicate that NC in NFA with 9 and 14%Cr are not detectably changed by irradiation at ~500°C with light ions, heavy ions or neutrons and no bubbles/cavities larger than 2 nm form in MA957 neutron irradiated at 500°C to 9 displacements per atom (dpa) with ~380 appm He. Characterization by conventional transmission electron microscopy (TEM) is supplemented by energy-filtered TEM (EFTEM) methods such as thickness and elemental (Fe-M, O, Ti-L, Cr-L) mapping. Importantly, Fe-M jump-ratio images reliably reveal NC as small as 2 nm diameter for sufficiently thin regions (<50 nm) and are insensitive to surface oxide films or modest surface contamination. Specimens of 12YWT and MA957 have been neutron irradiated to 9 dpa at ~500°C; microstructural characterization is in progress. Unless radical changes in size or concentration are induced, studies of the effects of irradiation on NC are hampered by the highly heterogeneous NC distributions. The dominance of TEM-specimen surfaces as point-defect sinks notwithstanding, we are pursuing the use of in-situ ion irradiation of 14YWT to study the effects of irradiation on NC (and of NC on the development of damage structure). In-situ heavy-ion irradiation at ~25 and ~500°C at the JANNuS facility in France will allow NC to be imaged by EFTEM as a function of ion dose. Experiments are also in progress using an alternative approach that involves characterization, including EFTEM imaging of NC at Oak Ridge National Laboratory (ORNL), of selected regions of 14YWT TEM specimens before and after in-situ ion irradiation at the IVEM-Tandem Facility of Argonne National Laboratory (ANL). The vacuum quality at the specimen during elevated-temperature in-situ irradiation is of great importance because of potential interstitial-impurity (e.g. O, C or N) pick-up or even oxidation, especially since NC imaging by EFTEM is limited to such thin regions. Even without irradiation, in-situ annealing of 14YWT at 500°C for 1 h at ~2 x 10-7 Torr resulted in severe specimen degradation. Research supported by the Division of Materials Sciences and Engineering, and at the ORNL SHaRE User Facility by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy. Special thanks to D. Kaoumi and A.T. Motta (Penn State), and M.A. Kirk (ANL) for exploratory in-situ irradiations at ANL.
3:30 PM - **V3.4
Multiscale Modelling of High Electric Field Effect on Metal Surfaces.
Flyura Djurabekova 1 4 , Helga Timko 2 , Aarne Pohjonen 1 , Leila Costelle 4 , Kai Nordlind 1 4 , Konstantin Matyash 3 , Ralf Schneider 3 , Sergio Calatroni 2 , Walter Wuensch 2
1 , Helsinki institute of Physics, Helsinki Finland, 4 Department of Physics, University of Helsinki, Helsinki Finland, 2 , CERN, Geneve Switzerland, 3 , Max-Planck-Institut fur Plasmaphysik, Greifswald Germany
Show AbstractSparks near metal surfaces cause a considerable damage to metal parts in devices employing high gradient electric fields. The next generation of high-end particle accelerators, needed to unravel the fundamental structure of matter in the universe, willbe linear colliders. The design of future accelerators such as the Compact LInear Collider (CLIC) involves very high gradient electric fields (~ 100 MV/m). Unfortunately, the upper energy limit of the beams is strongly restricted by the significant probability of electrical breakdowns inside of rf-structures, known as sparking. In the same time, fusion reactors, that involve high electric and magnetic field gradients, also experience problemsrelated to sparking phenomena.The trigger of sparking is a matter of long-standing debate, nonetheless, it still remains absolutely unclear. Despite the fact that the surfaces of the inner parts of rf-structures are thoroughly treated before use and operated under ultra-high vacuum conditions, the probability of sparks is still significant. Insight into the triggering of sparking, can help in managing sparking and arcing problems occurring both in the particle accelerators and in fusion reactors. As a successive process to the breakdown triggering, the formation of a near-surface plasma must be considered, where ions can be accelerated towards the surface and cause further surface damage by sputtering.We are developing a three-step multiscale modelling scheme to simulate the onset, plasma buildup and surface damage aspects of sparking.For the onset, we have developed a novel hybrid Electrodynamics-Molecular Dynamics (ED-MD) code on a base of the parcas MD code, which allows simulating the evolution of surfaces under high electric fields. We have tested the model in the regime of dc electric field evaporation (10 GV/m) and clearly observed single atoms being dragged out of the surface. For the plasma buildup, we employ Particle-in-Cell simulations with Monte Carlo collisions (PIC MCC). These show that under conditions relevant to linear colliders, a sheath potential forms in the plasma, which accelerates Cu ions towards the surface with fluxes of the order of 10^25 ions/cm^2/s with an energy distribution peaked around 10 keV. For the surface damage, we use MD simulations taking as input the flux and energy distribution from the PIC simulations.The MD simulations show that the formation of spark craters is due to multiple overlapping heat spikes producedby the ions accelerated in the plasma sheath.
V4: Ceramic Materials and Wasteforms I
Session Chairs
Monday PM, November 30, 2009
Room 207 (Hynes)
4:30 PM - **V4.1
Mechanisms of Radiation Damage and Properties of Nuclear Materials.
Gregory Lumpkin 1 , Katherine Smith 1 , Karl Whittle 1 , Bronwyn Thomas 1 , Nigel Marks 2
1 Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation, Menai, New South Wales, Australia, 2 Nanochemistry Research Institute, Curtin University of Technology, Perth, Western Australia, Australia
Show AbstractA wide range of materials are currently under consideration for use in advanced nuclear fuel cycle applications. The effects of radiation on these materials by exposure to external neutron irradiation and internal alpha and beta decay processes may have significant effects on the physical and chemical properties. This is especially true for materials that are subject to hundreds of displacements per atom during their service life. In this paper, we explore some of the radiation damage mechanisms prevalent in oxide based materials, including mathematical models and other concepts of amorphization (e.g., percolation), the role of defects on picosecond time scales, and longer term effects such as diffusion and recrystallization. As radiation "tolerance" or the ability of a material to maintain crystallinity under intense irradiation is a key issue for many fuel cycle applications, we will briefly review and comment on some of the underlying factors that have been identified as important in driving the short-term damage recovery. These include aspects of the structure (e.g., connectivity, polyhedral distortion), bonding, energetics of defect formation and migration, and melting point and similar criteria. The primary materials of interest here are those under development as special purpose nuclear waste forms, novel materials for separations, inert matrix fuels, and transmutation targets. In this context, we will illustrate the behavior of simple oxides and several more complex oxides such as perovskite, multicomponent fluorite systems, and related derivative structures (e.g., pyrochlore and zirconolite). The damage mechanisms in these materials are briefly compared with those of intermetallic and metallic materials.
5:00 PM - V4.2
Comparison of Microstructural Changes in ZnAl2O4 Spinel Under Ion Irradiation in the Electronic and in the Nuclear Energy Loss Regime.
Alexis Quentin 1 , Isabelle Monnet 1 , Dominique Gosset 2 , David Simeone 2 , Christina Trautmann 3 , Laurence Herve 4 , Serge Bouffard 1
1 , CIMAP-CIRIL, Caen France, 2 , CEA/DEN/DMN/SRMA/LA2M, Gif-sur-Yvette France, 3 , GSI, Darmstadt Germany, 4 , CRISMAT, Caen France
Show Abstract ZnAl2O4 is a typical ternary compound spinel that belongs to the space group Fd-3m where the anion sublattice is arranged in a cubic close-packed network and the cations are distributed in one-eighth of the tetrahedral sites and in half of the octahedral sites. This structure is known for exhibiting cation exchange versus temperature, and the space group remains Fd-3m over a broad temperature range. Under irradiation, ZnAl2O4 undergoes additional structural change. Whatever the nature and the energy of the incident particles, a crystal-crystal transition occurs at room temperature for different fluence values [1,2,3]. In the nuclear energy loss regime, the irradiation with 4 MeV Au ions transformed part of the initial phase into a random phase characterized by cations randomly occupying octahedral and tetrahedral sites. With increasing fluence, the volumic fraction of this beam-induced random phase follows an S-like shape and finally saturates at 80% [2]. In the electronic energy loss regime, high energy ions can also induce cation inversion in spinels [4], and in addition amorphisation by defect accumulation [5]. For 91-MeV Xe ions, the amorphous phase appears above a critical fluence of 4×1012 cm-2 and grows with increasing fluence. Using X-ray diffraction in combination with Rietveld analysis and transmission electron microscopy, the inversion parameter, the amorphous fraction, and the size of diffracting domains were analyzed for polycrystalline samples irradiated with different swift heavy ions such as 83-MeV Kr, o91-MeV, Xe, 740-MeV Zn , and, 2-GeV Au. provided at GANIL and GSI.[1] D. Simeone, C. Dodane-Thiriet, D. Gosset, P. Daniel, M. Beauvy, Journal of Nuclear Materials 300 (2002) 151[2] G. Baldinozzi, D. Simeone, D. Gosset, M. Dolle, L. Thomé, L. Mazérolles , Nuclear Instruments and Methods in Physics Research B 250 (2006) 119[3] G. Baldinozzi, D. Simeone, D. Gosset, S. Surblé, L. Mazérolles, L. Thomé, Nuclear Instruments and Methods in Physics Research B, 266 (2008) 2848[4] K. Yasuda T. Yamamoto, M. Shimada, S. Matsumura, Y. Chimi, N. Ishikawa , Nuclear Instruments and Methods in Physics Research B, 250 (2006) 238[5] A. Quentin, I. Monnet, D. Gosset, B. Lefrançois, S. Bouffard, Nuclear Instruments and Methods in Physics Research B, 267 (2009) 980
5:15 PM - V4.3
Transmission Electron Microscopy Observations of Alpha-Al2O3 Irradiated at High Temperature with 10 MeV Au Ions.
Jonghan Won 1 , Igor Usov 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractAlpha-alumina (α-Al2O3) is a widely-used industrial ceramic that is also being considered for application in certain radiation environments. There has been significant prior work regarding radiation damage behavior of alumina under neutron, electron, and ion irradiation conditions. However, the behavior of α-Al2O3 under high-energy, high-mass, high-temperature ion irradiation conditions, has not been studied to date. We report here on such a study in which high-temperature ion irradiation damage evolution in α-Al2O3, due to 10 MeV Au ions, was analyzed using cross-sectional transmission electron microscopy (TEM).The pristine α-Al2O3 samples used for this study included both polycrystalline alumina (commercially-available sintered alumina from Coors Tek) and single-crystalline sapphire (c-cut, (0001) sapphire from Union Carbide). We irradiated these samples with 10 MeV Au3+ ions at elevated substrate temperatures (up to 1273 K). Irradiations were performed to an ion fluence of 5x1015 Au/cm2. The ballistic damage profile for these irradiation conditions (estimated using the Monte Carlo code SRIM) indicates that the peak displacement dose is approximately 12 displacements per atom (dpa) at fluence 5x1015 Au/cm2, and this peak occurs approximately 2 μm beneath the sample surface. Grazing incidence X-ray diffraction (GIXRD) measurements indicate that there is no phase transition following ion irradiation. However, cross-sectional TEM observations of polycrystalline alumina samples revealed irradiation-induced dislocation loops at a depth of ~2.4 μm from the surface, and a high density of voids closer to the free surface. The size of these voids was found to range from 1 to 10 nm in diameter. These voids seem to differ from those observed in previous neutron and ion irradiation experiments. These voids seem to be randomly oriented and increase in size closer to the free surface and to pre-existing pores.
5:30 PM - **V4.4
Molecular Dynamics Simulation of Radiation Damage Accumulation in Pyrochlores.
Ram Devanathan 1 , William Weber 1
1 Chemical & Materials Sciences Division, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractWe have used molecular dynamics simulations to examine fundamental mechanisms of radiation damage accumulation and phase transformation in pyrochlores. In the present study, high energy recoils were simulated in Gd2Ti2O7 and Gd2Zr2O7 using rigid ion potentials. In addition, the accumulation of cation and anion sublattice defects was also studied. Our results show that the high mobility of defects, such as oxygen vacancies, plays an important role in defect annihilation. Moreover, recoil energy is dissipated by replacement collision sequences in pyrochlore, which reduces damage accumulation. In gadolinium zirconate pyrochlore, there are mechanisms for the accommodation of radiation damage, which result in minimal volume expansion and energy increase. As a result, there are considerable differences in the evolution of mechanical properties in Gd2Ti2O7 and Gd2Zr2O7. These results provide valuable insights into experimental observations of radiation damage in pyrochlores and will be discussed in light of experimental findings.
Symposium Organizers
Gianguido Baldinozzi CEA-CNRS-ECP
Yanwen Zhang Pacific Northwest National Laboratory
Katherine L. Smith Embassy of Australia
Kazuhiro Yasuda Kyushu University
V5: Modelling Complex Materials I
Session Chairs
David Simeone
William Weber
Tuesday AM, December 01, 2009
Room 207 (Hynes)
9:30 AM - **V5.1
Irradiation Studies in Non-metallic Materials : Impact of ab-initio Calculations.
Yves Limoge 1 , Layla Martin-Samos 2 , Guido Roma 1
1 DMN/SRMP, C.E.A. FRANCE, Gif sur Yvette France, 2 Democritos and Sissa, CNR-IFNM , Trieste Italy
Show AbstractThe study of points defects in solids, and their contribution to matter transport and related properties, has been renewed in the last ten years or so by the use of the so called ab-initio methods. Being based on the determination of the properties of the defects using a fully quantum treatment of the bonding, and involving a very small number of adjustable parameters, if not zero, these techniques have proved to be able to allow the determination of the base parameters of defects with a great precision. In non-metallic materials these capabilities however are for a part impeded by several problems, either of a computational nature, or in a deeper way linked to the approximations used for solving the Schrödinger equation. Among the first source of errors is the so called ~Simages interaction problem~T, linked to the periodic boundary conditions frequently used as a model of an infinite body. The last kind of difficulties is due to the well kown problems suffered by most of the ab-initio methods in handling the band gap of non-metallic systems. As is well known the DFT approach of the electronic structure is tailored for ground state studies, so it underestimates generally the width of the forbidden band. These errors have many detrimental consequencies for the defects having deep levels in the gap, which indeed can as a consequence be also badly described. This drawback is particularly severe for the charged states of these defects, the formation energy of which is prone to be given in error by several eV. In this work we will show on a few simple systems how the two kinds of errors can be solved in a more satisfactorily manner using state of the art electronic structure tools, in particular the GW method. We will also discuss the materials conditions leading to such a catastrofic failure of the stand
10:00 AM - V5.2
First-principles Modeling of Nuclear Fuel Materials with High Efficiency and Accuracy.
Fei Zhou 1 , Vidvuds Ozolins 1
1 Materials Science & Engineering, UCLA, Los Angeles, California, United States
Show AbstractActinide compounds present serious challenge to moderndensity-functional theory (DFT) based electronic-structure techniques due to strong electron correlations and orbital ordering phenomena of the localized f-electrons. While advanced quantum-mechanical methods such as self-interaction-corrected local-density approximation (SIC-LDA), hybrid functional and dynamical mean-field theory (DMFT) have been used to investigate actinide compounds, these methods are usually computationally expensive and limited to small system. The LDA+U method, which combines the efficiency of LDA with an explicit treatment of correlation for the f-electrons, has received much interest for studying actinide compounds. High computational efficiency means that relatively large and complicated systems can be modeled with this method.We have identified a critical problem with the currently available versions of LDA+U: although the method was invented to remove self-interaction of localized electrons, significant orbital-dependent self-interactions remain. These aspherical self-interaction errors are up to 0.4 eV per electron, leading to erroneous orbital ground states in many cases. An alternative scheme that improves upon the original LDA+U is proposed as a remedy. We show that our method reproduces the expected degeneracy of $f^1$ and $f^2$ states in free ions and the correct ground states in the PrO2 and UO2 solids. In particular, the Gamma 5 orbital state of UO2 is confirmed as the ground state. As a first application, the crystal-field excitation energies to the Gamma 3, 4 and 1 states of UO2, between 0.1 to 0.2 eV, are reproduced with good accuracy compared to experiment.This work was supported by the U.S. Department of Energy, Nuclear Energy Research Initiative Consortium (NERI-C).
10:15 AM - V5.3
First-principles Calculations of Lattice Defects in γ-Uranium.
Daniel Aberg 1 , Paul Erhart 1 , Babak Sadigh 1
1 Condensed Matter and Materials Division, Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractUranium has a polymorphic phase diagram and undergoes several structural transitions as a function of temperature. The low- and high-temperature phases are related to the α-U and γ-U structures, respectively. To predict the microstructural evolution under irradiation and thermal gradients using KMC or PFM simulations, point defect information is needed. As these quantities must be obtained as a function of temperature we have performed ab-initio MD simulations of γ-U at elevated temperatures to study the stabilization of γ-U, point defect formation energies and diffusion. For the ideal material, we show that at temperatures near the α-γ transition the cubic phase is stabilized by anharmonic vibrations and that the short-range order is α-like whereas the long-range order is γ-like. We also obtain vacancy and interstitial formation energies on the order of 1 eV and predict that the barriers for their migration are very small. Prepared by LLNL under Contract DE-AC52-07NA27344.
10:30 AM - V5.4
Activation Energies for Xe Transport in UO2±x From Density Functional Theory Calculations.
David Andersson 1 , Pankaj Nerikar 1 , Blas Uberuaga 1 , Christopher Stanek 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractFrom a thermodynamic perspective most fission gases have low solubility in the fuel matrix and as a result there is a significant driving force for segregation of gas atoms to heterogeneities such as grain boundaries and subsequently for nucleation of gas bubbles. Under this assumption one of the controlling steps for evolution of fission gas micro-structures is diffusion of individual gas atoms through the fuel matrix to existing bubbles or grain boundaries (sinks). This process is largely governed by the activation energy for bulk diffusion of gas atoms, the driving force for segregation to existing sinks (bubbles or grain boundaries) and their saturation limit. Here we have studied the bulk diffusion mechanisms of Xe, which is one of the most important fission gases, by calculating the corresponding activation energies as function of the UO2±x stoichiometry using density functional theory (DFT) methods. In the present study we assume Xe diffusion to occur via uranium vacancies that bind to the stable Xe trap sites. Estimating the activation energy involves determining Xe migration barriers as well as thermodynamics of Xe trap sites in UO2±x and their interactions with Uranium vacancies that enable Xe in trap sites to move. We present results for all these components of the activation energy and discuss the importance of appropriately treating charge-compensation for defects in UO2±x in order to best reproduce experimental data. Since diffusion of Xe atoms is closely connected to diffusion of Uranium vacancies, we have also analyzed the stoichiometry dependent activation energy for diffusion of uranium ions via vacancy mechanisms. Due to the complex nature of the point defects and the clusters of point defects that interact with Xe atoms, the DFT based modeling is inevitably associated with uncertainties that in some cases may be rather significant. In order to mitigate this issue, we have tried to identify systematic errors, after which we categorize the most probably diffusion mechanisms. In order to achieve self-consistency and assess the accuracy of our conclusions, this exercise is complimented by re-analysis of key experimental data within the framework of proposed diffusion models.
11:15 AM - **V5.5
Chemical Evolution through Radioactive Decay: A Case Study in Sr-90.
Nigel Marks 1 , Ashley Lawler 1 , Damien Carter 1 , Chao Jiang 2 , Chris Stanek 2 , Kurt Sickafus 2 , Blas Uberuaga 2
1 Nanochemistry Research Institute, Curtin University of Technology, Perth, Western Australia, Australia, 2 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractWasteform science has traditionally addressed questions such as radiation tolerance, uptake of fission products into host phases and aqueous durability. Much less attention has been paid to transmutation-driven chemical evolution arising from the radioisotopes themselves. For alpha-decay it is arguably reasonable to neglect variations in the chemistry over time, since the recoil of the daughter nucleus is extremely energetic (tens of keV) and amorphization-recrystallization processes are paramount. With beta-decay, however, chemical effects provide the dominant driving force for change within the wasteform.
In this work we quantify chemical evolution within the solid-state using a combination of molecular dynamics (MD) and density functional theory (DFT). Considering Sr-90 as a prototypical beta emitter, we study the physical and chemical processes occurring when Sr-90 decays first to Y-90 (t½=29 years) and then onwards to stable Zr-90 (t½=64 hours). By combining calculations of beta-decay energetics with MD simulations of threshold displacement energies, we show that Sr-90 recoil within strontium-titanate (SrTiO3) does not induce defects. Consequently, the chemical makeup of the system evolves over time with its crystal structure intact. To study the effect of this unusual behavior, we turn to DFT calculations to elucidate the changing structural and energetic stability with increasing Zr fraction. Two illustrative materials are considered: Sr(Zr)TiO3, which becomes markedly less stable over time, and Sr(Zr)H2, which undergoes an ionic/metallic transition in which the heat of formation remains largely unchanged.
11:45 AM - V5.6
Molecular Dynamics Simulations of Ordered Li4SiO4.
Samuel Murphy 1 , David Parfitt 1 , Robin Grimes 1
1 Department of Materials, Imperial College London, London United Kingdom
Show AbstractA number of lithium containing ceramic materials are currently under consideration for use as a tritium breeding material in the European Helium Cooled Pebble Bed (HCPB) breeder blanket concept. One of the leading candidate materials is lithium orthosilicate (Li4SiO4) due to its high lithium density and good chemical compatibility with other blanket materials. Transmutation of the Li+ cations into 3T+ and He will lead to an increase in the concentration of lithium vacancies which in turn will affect the diffusion coefficient for Li+ migration. Here we use molecular dynamics to investigate the self diffusion of lithium in the perfect Li4SiO4 crystal and as a function of the vacancy concentration. The mechanisms underpinning Li+ transport are also discussed.
12:00 PM - V5.7
Multiscale Modeling of Helium-Vacancy Cluster Nucleation under Irradiation: A Kinetic Monte-Carlo Approach.
Tomoaki Suzudo 1 , Masatake Yamaguchi 1 , Hideo Kaburaki 1 , Ken-ichi Ebihara 1
1 , Japan Atomic Energy Agency, Tokai-mura Japan
Show AbstractStructural materials used in the future advanced reactors, such as fusion reactors and fast breeder reactors, are exposed to high energy neutrons, and helium atoms are produced in the materials through the transmutation reactions. The helium production causes a significant difference in irradiation effects of these materials from those used in current light-water-cooled reactors, because the accumulation of helium atoms and the nucleation of helium bubbles lead to, what we call, helium embrittlement whose detailed mechanism in not known. Because locating experimentally helium atoms is difficult, computational modeling is expected to play an important role in the identification of this mechanism.We describe the application of an object kinetic Monte-Carlo modeling that ab initio calculations provide critical parameters, such as the migration and formation energies of point defects and the dissociation energies of helium and vacancy from helium-vacancy clusters. We simulate radiation by the production of frenkel pairs and helium atoms and track the fate of point defects such as SIAs, vacancies, and helium atoms. The method is useful for locating helium atoms. The materials studied in the model are pure face-centered-cubic iron and pure body-centered-cubic iron; they are used as surrogate materials for austenitic and ferritic/martensitic steels, respectively. We put a special emphasis on the modeling of helium-vacancy-cluster growth in these materials and critically analyze the results in light of the similar studies and experimental results.
12:15 PM - V5.8
Stochastic Mean-field Approach for Simulations of Radiation Effects in Complex Materials.
Vasily Bulatov 1
1 , LLNL, Livermore, California, United States
Show AbstractRate Theory (RT) has been used for simulations of irradiated materials for over 40 years. RT is a mean-field method in which material microstructure is represented by volume-averaged populations of various defect species evolving under irradiation. By neglecting correlations and fluctuations, RT achieves high computational efficiency allowing simulations of damage accumulation on the reactor time-scales. However the method becomes unwieldy if and when complex defect populations need to be considered, e.g. vacancy clusters containing helium, oxygen, hydrogen, carbon and other impurities. The number of differential equations required to resolve the evolving complex defect population can become too large to fit into computer memory and/or practically solved: situations like this are sometimes referred to as combinatorial explosion. In bio-chemistry, a practical solution to this unpleasant problem has been proposed by D. T. Gillespie in 1977 [1]. Here we extend Gillespie’s ideas to modeling complex materials under irradiation by re-casting the standard Rate Theory in the form of integer-valued populations of defect clusters in a finite material volume. The discrete populations are then evolved stochastically, using an appropriate dynamic Monte Carlo algorithm. Taking a relatively simple material model as an example, we show that the stochastic method predicts the same evolution of average defect concentrations as in the standard deterministic Rate Theory. At the same time, unlike the standard method, the discrete stochastic approach captures finite-volume variations in defect populations and, most importantly, the method’s computational complexity does not depend on the complexity of the defect cluster population but is scaled by the size of simulation volume. This highly desirable property makes it principally possible to extend the Rate Theory method to simulations of defect populations of arbitrary complexity. [1] D. T. Gillespie (1977). "Exact Stochastic Simulation of Coupled Chemical Reactions". J. Phys. Chem. 81 (25): 2340–2361.
12:30 PM - **V5.9
Radioparagenesis: The Evolution of Crystalline Waste Form Structure via Transmutation.
Chao Jiang 1 , Christopher Stanek 1 , Nigel Marks 2 , Kurt Sickafus 1 , Blas Uberuaga 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Nanochemistry Research Institute, Curtin University of Technology, Perth, Western Australia, Australia
Show AbstractAs the world enters a nuclear renaissance, the challenges associated with nuclear power become even more pressing. Foremost amongst these is the disposal of nuclear waste. While several strategies have been investigated in the past, the possibility of a closed nuclear fuel cycle with the separation of individual fission products allows for the use of customized waste forms for each of those fission products. That is, for each fission product in the waste stream, a different crystalline host may be considered. While many past studies have focused on the radiation tolerance or leachability of candidate materials, very few have examined the structural stability of the material as the radioactive species transmutes into a new element.In this talk, we introduce the concept of radioparagenesis, or the formation of novel crystalline structures via the radioactive decay of one of the constituent species. Using density functional theory, we study the possible crystal structures that an initial model waste form may evolve towards when the radioisotope decays. We find that most materials form novel structures that were not anticipated, structures that, while metastable, are mechanically and dynamically stable. The formation of such radioparagenetic phases suggests a backward design approach to waste forms in which the material is chosen such that it becomes more stable as the transmutation occurs.
V6: Metallic Materials II
Session Chairs
Tuesday PM, December 01, 2009
Room 207 (Hynes)
2:30 PM - V6.1
BCC Fe-Cr Surfaces under Stress: A First Principles Study.
Anna Nikiforova 1 , Bilge Yildiz 1
1 Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge , Massachusetts, United States
Show AbstractAdvanced materials development for nuclear energy requires a fundamental understanding of materials behavior in extreme environments. Stress corrosion cracking (SCC), a sudden failure of normally ductile metals, is one of the main causes of degradation of materials subjected to a tensile stress in a corrosive environment. The objective of this work is to determine the atomistic relations of the chemo-mechanical behavior of interfaces, in particular the bonding characteristics, to the initiation of SCC in Fe-Cr. In the present work, investigation of the initiation of SCC of bcc Fe-Cr alloy started at the electronic level using ab initio codes to capture the chemical and micromechanical characteristics. Density Functional Theory (DFT) as implemented in VASP is being used for study the electronic state of the Fe-Cr surfaces as well as chemical reaction of oxygen and the alloy surface to determine the changes in surface reactivity with tensile strain. This approach can allow us to tie the stress-driven changes in electronic structure and reactivity to the SCC initiation mechanism. The stress-strain relation in <001> direction of Fe and Cr was calculated using ab initio simulations and was benchmarked with the results available in the literature [1-5] in order to validate the model and the code performance. The ideal stress-strain relation is of interest because it can reveal fundamental insights on the connections between the bonding and symmetry of the crystal. The changes in Fermi energy and density of electronic states (DOS) during deformation were calculated. The stress-driven changes in DOS and Fermi surface can be linked to bonding characteristics of Fe and Cr. We found that both the shape of Fermi surface and the DOS of Fe and Cr changed significantly with tensile strain. The hypothesis for the effect of strain on the reactivity of bcc Fe-Cr surface to oxygen and the initiation mechanism of SCC will also be discussed in the presentation.Bibliography[1]S. V. Okatov, A. R. Kuznetsov, Yu. N. Gornostyrev, V. N. Urtsev, M. I. Katsnelson, "Effect of magnetic state on the - transition in iron: First-principles calculations of the Bain transformation path", Phys. Rev. B, 79, 094111, pp. 1-4 (2009)[2]M. Friak, M. Sob, V. Vitek, "Ab initio calculation of phase boundaries in iron along the bcc-fcc transformation path and magnetism of iron overlayers", Phys. Rev. B, 63, 052405, pp. 1-4 (2001)[3]M. Friák, M. Šob, V. Vitek, "Ab initio calculation of tensile strength in iron", Philosophical Magazine, 83, 31-34, pp. 3529-3537 (2003)[4]D. M. Clatterbuck, D. C. Chrzan, J. W. Morris Jr, "The inherent tensile strength of iron", Philosophical Magazine Letters, 82, 3, pp. 141-147 (2002)[5]D.M. Clatterbuck, D.C. Chrzan, J.W. Morris Jr, "The ideal strength of iron in tension and shear", Acta Materialia, 51, p. 2271–2283 (2003)
2:45 PM - V6.2
Radiation Response of Nanostructured Ferritic Alloys.
Michael Miller 1 , David Hoelzer 1 , Kaye Russell 1
1 MSTD, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractIn order to meet future energy demands, advanced materials will be required that maintain their mechanical properties under extreme doses of radiation at elevated temperatures. In order to meet this requirement to the end of life of a component, a microstructure that is highly resistant to radiation damage is essential. One class of material that is under consideration for these extreme environments is the nanostructured ferritic alloys (NFA) - formerly referred to as oxide dispersion strengthened (ODS) steels. Nanostructured ferritic alloys, such as 12YWT, 14YWT and MA957 alloys are produced by mechanically alloying pre-alloyed metals and yttria powders. This fabrication method forces all the elements in the powders into solid solution and produces a high concentration of vacancies. Atom probe tomography has shown that there is a high number density of titanium-, oxygen- and yttrium-enriched nanoclusters in these nanostructured ferritic alloys. The nanoclusters and the grain size are remarkably stable during high temperature isothermal aging at temperatures up to 1400 °C and during long term creep at elevated temperatures (850 °C). Consequently, these unique materials are candidates for use under extreme conditions in future generations of advanced reactors. However, atomic displacement cascades produced during neutron or ion irradiations can induce mechanisms that can potentially destabilize or destroy these nanoclusters, change the vacancy and interstitial atom distribution, and thereby change the properties. The increase in vacancy concentration may enhance diffusion, which may result in a coarsening of the nanoclusters or a change in the number density. Therefore, the solute distribution associated with, and the stability of the nanoclusters under high dose irradiation conditions, have been investigated by atom probe tomography.The radiation response of a 12YWT alloy was characterized after neutron irradiation to doses of up to 9 dpa and temperatures between 300 and 600 °C and a MA957 alloy was characterized after neutron irradiation to doses of up to 3 dpa at 600 °C. For comparison, the MA957 was also characterized after isothermal creep under an applied tensile stress of 100 MPa for 38,500 h at 800 °C. Details of the changes in the size, number density, and compositions of the nanoclusters will be presented. For all conditions studied, high number densities of ultrafine scale titanium-, oxygen-, and yttrium-enriched nanoclusters were observed.This research was sponsored by the U.S. Department of Energy, Division of Materials Sciences and Engineering; research at the Oak Ridge National Laboratory SHaRE User Facility was sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.
3:00 PM - V6.3
A Molecular Dynamics Study on Hydrogen Embrittlement of a Grain Boundary in α-iron.
Tomoko Kadoyoshi 1 , Hideo Kaburaki 1 , Mitsuhiro Itakura 1 , Masatake Yamaguchi 1
1 Center for Computational Science and e-Systems, Japan Atomic Energy Agency, Tokai, Ibaraki, Japan
Show AbstractThe hydrogen embrittlement has been known for over a century when metals are under corrosive, welding, and irradiation conditions, however, its mechanism has not yet been precisely identified. It is empirically established that steels are susceptible to hydrogen embrittlement as the tensile strength exceeds approximately 1 GPa. In particular, it is shown in recent experiment of high strength steels that a clear crossover in fracture mode from quasi-cleavage to intergranular fracture is observed as a function of charged hydrogen bulk concentration. Here, we concentrate on studying the grain boundary embrittlement of α-iron in the high hydrogen concentration region. Molecular dynamics method is mainly employed to study decohesion properties of a Σ3 grain boundary in the presence of hydrogen. A newly developed empirical potential of Fe-H is used based on the empirical Fe potential. Firstly, cohesive strength or surface energy of a grain boundary is estimated as a function of separation distance to check the validity of the empirical potential by comparing the molecular dynamics results with the first principles results. Moreover, cohesive strength of a grain boundary is observed as a function of temperature under the tensile stress condition. Secondly, cohesive strength of a Σ3 grain boundary is measured as a function of the number of hydrogen atoms. The results of the first principles calculation show that strong segregation of hydrogen occurs at interstitial sites in the grain boundary region and at the opening surface. A detailed comparison of segregation energy and cohesive strength over various configurations of hydrogen atoms in the grain boundary is made using the molecular dynamics and first principles results. All these facts indicate that the grain boundary embrittlement does occur due to the decohesion mechanism only in the presence of hydrogen. Finally, we present the molecular dynamics results of crack advancement in the grain boundary region as a function of hydrogen concentration under the mode I loading condition. The atomistic results are compared with the mesoscopic cohesive zone continuum model based on the first principles results. This study was carried out as a part of research activities of "Fundamental Studies on Technologies for Steel Materials with Enhanced Strength and Functions" by Consortium of JRCM (The Japan Research and Development Center of Metals). Financial support from NEDO (New Energy and Industrial Technology Development Organization) is gratefully acknowledged.
3:15 PM - V6.4
Microstructure-Property Evolution of Steels at High Damage Levels for Advanced Nuclear Reactors.
Khalid Hattar 1 , Luke Brewer 1 , Ping Lu 1 , Janelle Branson 1 , Barney Doyle 1
1 , Sandia National Laboatories, Albuquerque, New Mexico, United States
Show AbstractThe evolution of cladding steel microstructures and the resulting mechanical properties at high damage levels, greater than 100 displacements per atom (dpa), is a key area of study for advanced reactor technologies. However, the high damage levels of interest are created over several years in fast neutron reactors. Ion beam irradiation has been and is currently used as one way to accelerate the irradiation damage in metals to simulate neutron damage. Ion beam damage is spatially localized and requires microscale techniques for study.The approach presented here investigates radiation effects on the microstructure and properties of steels by combining high energy, heavy ion irradiation with in situ scanning electron microscopy (SEM) imaging, micropillar compression tests, and ex situ transmission electron microscopy (TEM) analysis. The Sandia tandem accelerator is used to provide the ions. An SEM attached to an endstation of the accelerator permits time sequence imaging of swelling and other microstructural evolution during implantation. In order to determine the effect of radiation damage and resulting microstructural changes on mechanical properties within the irradiated volume, micropillar compression testing is done. Finally, analytical TEM investigations are performed to evaluate the change in bubble density and size as well as to evaluate the precipitation of intermetallics within the steel microstructure. The results of this study will be compared to models, simulations, previous ion irradiation studies, and fast neutron exposures at lower damage levels.This work is supported by the Division of Materials Science and Engineering, Office of Basic Energy Sciences, U.S. Department of Energy. Sandia is a multi-program laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United States Department of Energy’s National Nuclear Security Administration under Contract No. DE-AC04-94AL85000.
3:30 PM - V6.5
Evaluation of the Fracture Toughness of the SA508 Gr.4N Alloy Steels Based on the Master Curve Approach.
Ki Hyoung Lee 1 , Min Chul Kim 2 , Bong Sang Lee 2 , Dang Moon Wee 1
1 Material Science & Engineering, KAIST, Dae-jeon Korea (the Republic of), 2 Nuclear Materials Research Division, KAERI, Dae-jeon Korea (the Republic of)
Show AbstractDemands for materials with higher strength and toughness are rising to increase power generation capacity and operation life of nuclear power plants. SA508 Gr.4N Ni-Mo-Cr low alloy steel, which has higher Ni and Cr contents compared to SA508 Gr.3 alloy steel, is considered as a candidate due to the excellent strength and toughness from its tempered martensitic microstructure. In this study, an evaluation of the fracture toughness behavior was performed on the SA508 Gr.4N steel model alloys based on the master curve approach in the transition temperature region. Model alloys were fabricated by changing the contents of alloying elements such as Ni, Mo and Cr based on chemical composition range of SA508 Gr.4N alloy steels in the ASME specification. Fracture toughness was evaluated from 3-point bend tests with pre-cracked Charpy V-notch(PCVN) specimens according to ASTM E1921-08. The Weibull plots of the fracture toughness values presented that the master curve approach based on Weibull statistics was suitable to evaluate the fracture toughness of the reference model alloy. However, the data sets showed that the fracture toughness value increased faster with temperature than that of the bainitic SA508 Gr.3 steels though the overall behavior followed the tendency of standard master curve. Moreover, T0 values determined from single-temperature data sets were lowered as test temperature increased and were much different from T0 value, -135.3°C, determined from a multi-temperature procedure. In order to compensate the steeper temperature dependency of the fracture toughness, adjustment of exponential parameter in master curve equation related to shape of curve were attempted by fitting data sets. As a result, the modified master curve equation described the fracture toughness behavior properly through the overall transition temperature region. For the other model alloys with different chemical composition, the shape of modified master curve could be better suited to the distribution of test data. Therefore, it was considered that the modified master curve described the overall temperature dependency of the fracture toughness in the tempered martensitic SA508 Gr.4N alloy steels appropriately.
3:45 PM - V6.6
FeCr Swelling under Helium Irradiation.
Magdalena Caro 1 , Alexander Stukowski 2 , Paul Erhart 1 , Babak Sadigh 1 , Alfredo Caro 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States, 2 , Technical University Darmstadt, Darmstadt Germany
Show AbstractFe-Cr alloys with 9-12% Cr content are the base matrix of advanced ferritic/martensitic (FM) steels envisaged as fuel cladding and structural components of Gen-IV reactors, and in future fusion power plant first wall and blanket structures. These steels show good mechanical properties and good resistance to swelling. However, Helium can accelerate the nucleation of cavities in FeCr based steels and a detailed understanding of the thermodynamic aspects of Cr and He segregation is required to develop the capability of designing swelling resistant microstructures. We have developed a formulation of an empirical interatomic potential that incorporates the complexities of the thermodynamics of the FeCr system, adding He as a third element in the alloy, using results for Fe-He and Cr-He interactions developed by K. Nordlund's group. We use a novel numerical approach based on variance-constrained transmutation ensemble implemented in a massively parallel hybrid Molecular Dynamics/Metropolis Monte Carlo code to study precipitation of He and Cr in grain boundaries and segregation at surfaces as a function of Cr composition. We present preliminary results on FeCr swelling under Helium irradiation, as well as on Helium pressure inside the bubbles. Our work represents a first step in the development of modeling capabilities to describe Cr and He segregation kinetic effects induced by radiation.Work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.
V7: Complexity in Advanced Fuels
Session Chairs
Tuesday PM, December 01, 2009
Room 207 (Hynes)
4:30 PM - V7.0
On the Origin of Large Interstitial Clusters in Displacement Cascades in Iron.
Andrew Calder 1 , David Bacon 1 , Alexander Barashev 1 , Yuri Osetsky 2
1 Department of Engineering, University of Liverpool, Liverpool United Kingdom, 2 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractDisplacement cascades with wide ranges of primary knock-on atom (PKA) energy and mass in iron have been simulated by MD. New visualization techniques show how the shock-front dynamics and internal structure of a cascade develop over time. They reveal that the nature of the final damage is determined early on before the onset of the thermal spike phase of the cascade process. A zone (termed ‘spaghetti’) in which atoms are moved to new lattice sites is created by a supersonic shock front expanding from the primary recoil event. A large cluster of self-interstitial atoms can form on the periphery of the spaghetti if a hypersonic recoil creates damage with a supersonic shock ahead of the main supersonic front. When the two fronts meet, the main one injects atoms into the low-density core of the other and they become interstitial atoms during the rapid recovery of the surrounding crystal. The hypersonic recoil that gives rise to an interstitial cluster occurs in less than 0.1 ps after the primary recoil event. The equivalent number of vacancies forms at times one to two orders of magnitude longer in the spaghetti core as the crystal cools. By using the spaghetti zone to define cascade volume, the energy density of a cascade is shown to be almost independent of the PKA mass. This throws into doubt the conventional energy-density interpretation of an increased defect yield with increasing PKA mass in ion irradiation of metals.
4:45 PM - V7.1
Chemical Sputtering of Metals.
Kai Nordlund 1 , Carolina Bjorkas 1 , Katharina Vortler 1 , Mooses Mehine 1 , Niklas Juslin 1
1 Department of Physics, University of Helsinki, Helsinki Finland
Show AbstractNumerous experiments have shown that carbon-based materials can sputter chemically by low-energy H isotope bombardment in fusion reactors, at energies where physical sputtering is impossible. At high temperatures this can be understood to be due to thermally activated desorption. At low temperatures the sputtering can be understood interms of of the swift chemical sputtering mechanism, in which an incoming D penetrates into an energetically unfavourable state between two carbon atoms, leading to bond breaking [Salonen et al, Phys. Rev. B 63 (2001) 195415; Nordlund, Physica Scripta T124 (2006) 53; Krstic et al, New J. Phys. 9 (2007) 209].Metals have in general not been observed to show as pronounced chemical sputtering by low-energy H ions as carbon-based materials, and Nordlund et al. argued that the swift chemical sputtering mechanism would not be significant in metals since they have much more neighbours than covalently bonded materials like C and Si [Nordlund etal, Pure and Applied Chemistry 78 (2006) 1203]. However, recent experiments by Doerner et al. show that for low (~ 50 eV) D energies, most of the outcoming Be is in fact in the form of BeD molecules, a clear signature of chemical effects.Using molecular dynamics simulations of the D bombardment of Be, Be2C and WC, we have now explored whether metals and metal carbides can sputter chemically. Our results show pronounced chemical erosion of BeD molecules, in agreement with experiments. Detailed analysis of the atom trajectories showed that the erosion can indeed be explained by the swift chemical sputtering mechanism, contrary to the earlier prediction by Nordlund.
5:00 PM - V7.2
The Formation of an Effective Space Charge in UO2.
Christopher Stanek 1 , Pankaj Nerikar 1 , Blas Uberuaga 1 , Anders Andersson 1 , Simon Phillpot 2 , Susan Sinnott 2
1 Material Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Dept. of Materials Science, University of Florida, Gainesville, Florida, United States
Show AbstractAtomistic simulations of defect processes that explicitly consider microstructural features such as surfaces and grain boundaries often reveal non-intuitive phenomena that are crucial for fully understanding material behavior as well as to construct an informed microstructural model. In this study, we employed pair potential and density functional theory atomistic simulations to consider the structure of several UO2 grain boundary types (e.g. Σ5 tilt, Σ5 twist and amorphous). Our simulations predict that for the symmetric Σ5 tilt boundary there is a lower energy distorted structure than the typically considered symmetric structure. Small distortions on the oxygen sublattice lead to an asymmetry of the boundary structure. Although the distortions from normal lattice positions are quite small, they are sufficient to induce an electric field in the ideal (0K) structure of the grain boundary. The implications of this field are analogous to conventional space charges in ceramics. For example, grain boundary segregation is spatially much more pronounced for charged defects (e.g. fission products) if the field exists than if segregation was dominated by strain effect alone. Discussed in this presentation will be: the formation of an electric field in UO2 due to asymmetric grain boundary structure, asymmetric boundary formation in other fluorite compounds (e.g. CeO2, ZrO2 and CaF2), implications for meso and continuum scale models and for nuclear fuel performance.
5:15 PM - V7.3
Can Alpha-damage Studies Help to Understand In-pile Behaviour of UO2 Fuels?
Thierry Wiss 1 , Vincenzo Rondinella 1 , Dragos Staicu 1 , Rudy Konings 1
1 , European Commission - Joint Research Centre - Institute for Transuranium Elements, Karlsruhe Germany
Show AbstractThe most commonly used nuclear fuel, UO2, is subjected to radiation damage not only during in-pile irradiation, but also during cooling and storage. Although the magnitude, rate and conditions of the damage accumulation are different for reactor irradiation and for (long time, low temperature) storage conditions, the damage pattern is very similar to some extent. Radiation effects in irradiated fuels have been characterized by different techniques including transmission electron microscopy, X-ray diffractometry, in combination with thermal annealing methods. In order to simulate alpha-damage accumulation in UO2, 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. Comparative analysis of irradiated fuel and alpha-doped materials allowed assessing the superimposition of alpha-decay effects onto in-pile radiation damage after fuel discharge. It is shown that the 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. Our results confirm that UO2 shows a remarkable ability to maintain its original fluorite structure even under severe irradiation conditions.
5:30 PM - V7.4
Hybrid Monte Carlo Simulation of Nuclear Fission Gas Bubbles Transportation in Nuclear Fuel.
Liangzhe Zhang 1 , Timothy Bartel 2 , Mark Lusk 1
1 Physics, Colorado School of Mines, Golden, Colorado, United States, 2 , Sandia National Laboratories, ABQ, New Mexico, United States
Show AbstractNuclear fission product noble gas atoms are known to nucleate and grow into bubbles that subsequently influence the macroscopic thermomechanical state. The classic example is swelling and and its tie to fracture. At the meso-scale, though, fission product bubbles influence grain boundary evolution, localized plastic deformation, and stress distributions at grain junctions. Meso-scale simulation tools can facilitate the characterization of such processes. The present investigation considers meso-scale microstructural evolution in response to fission product bubble formation within a fully coupled thermomechanical environment. Non-uniform distributions of grain boundary energy, elastic energy and dislocation energy generate driving forces for polycrystalline grain boundary motion and bubble movement. Bubble nucleation, growth, migration, coalescence are accounted for in the presence of temperature and stress gradients.To facilitate the numerical simulation of nuclear fuels, a hybrid Monte Carlo approach is used. This method directly couples an explicit time integration material point method (MPM) for mechanical stresses with a time calibrated Monte Carlo (cMC) model for grain boundary kinetics and bubble transport. A novel Monte Carlo plasticity (MCP) algorithm accounts for dislocation motion. As opposed to the conventional MC model, the cMC model endows MC simulation with physical time and length scales so that time accurate transport can be simulated. Where possible, the requisite database is informed using data generated with Density Functional Theory calculations. The resulting program is highly parallelized.The hybrid Monte Carlo paradigm is applied to study the high temperature evolution of UO2 microstructures in the presence of an evolving distribution of fission product gas bubbles. The results are compared with experimental observations.
5:45 PM - V7.5
The Study of the Noble Gas Bubbles Trapped in the UO2 Matrix.
Andrei Jelea 1 2 3 , Fabienne Ribeiro 1 , Roland Pellenq 2
1 DPAM/SEMCA/LEC, Institut de Radioprotection et Surete Nucleaire (IRSN), Saint Paul lez Durance France, 2 CINaM, CNRS, Marseille 13288 France, 3 , Institute of Physical Chemistry "IG Murgulescu", Bucharest Romania
Show AbstractThe aim of the present study is to improve the understanding at an atomic level of the behavior of Xe and Kr trapped in a UO2 matrix.In the first stage the variation of the elastic properties of UO2 (bulk modulus, elastic constants, Young modulus) versus porosity is studied through atomistic simulations with semiempirical potentials. For this purpose the energy minimization is employed. In order to describe the interactions between the atoms three potentials available in the literature [1] are chosen: Basak, Morelon and Arima. A good agreement was found between the elastic properties calculated in the present atomistic simulations and those coming from the homogenization calculations [2].The effect of the temperature on the stability of the voids (diameters ranging from 0.8 nm to 2.0 nm) is then studied through molecular dynamics simulations in the NVT and NPT statistical ensembles. Only the Basak form of potential is used to treat the interactions between the atoms. For the pressure P=0 atm and temperatures lower than 1200K the voids are stable but for T>2000K they crumble. The solid-liquid phase transition as calculated with this method occurs between 3400K and 3500K (the experimental value is T=3150K). Since the system has to cross a potential barrier associated with the creation of the solid-liquid interface, one may expect the NPT molecular dynamics simulation to give a higher temperature for this transition for the perfect UO2. The presence of voids induces a decreasing of the solid-liquid transition temperature.In the second stage Xe bubbles are created by filling the voids with Xe at constant temperature. This is achieved through Grand Canonical Monte Carlo simulations. Then, molecular dynamics calculations in NVT ensemble help to give an estimation of the stress induced in the UO2 matrix by the Xe contained in the bubbles. In these simulations the Xe-Xe interactions are described by a Buckingham potential as parametrized by Brearley and MacInnes [3]. For the Xe-UO2 interactions two kinds of potential are used: one proposed by Geng and al. [4] and another one which we have computed. Some preliminary results of this study will be presented. References[1] K. Govers, PhD thesis, Université libre de Bruxelles (2008).[2] J.M. Gatt, Y. Monerie, D. Laux, D. Baron, J. Nucl. Mater., 336 (2005) 144.[3] I.R. Brearley, D.A. MacInnes, J. Nucl. Mater., 95 (1980) 239.[4] H.Y. Geng, Y. Chen, Y. Kaneta, M. Kinoshita, J. Alloys Comp., 457 (2008) 465.
Symposium Organizers
Gianguido Baldinozzi CEA-CNRS-ECP
Yanwen Zhang Pacific Northwest National Laboratory
Katherine L. Smith Embassy of Australia
Kazuhiro Yasuda Kyushu University
V8: Carbides
Session Chairs
Wednesday AM, December 02, 2009
Room 207 (Hynes)
9:45 AM - V8.2
Role of Grain Size and Grain Boundaries in Irradiation Defect Production in Nanocrystalline SiC.
Narasimhan Swaminathan 1 , Dane Morgan 1 2 , Izabela Szlufarska 1 2
1 Material Science and Engineering, University of Wisconsin-Madison, Madison, Wisconsin, United States, 2 Material Science Program, University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractCubic silicon carbide (3C-SiC), known for its excellent mechanical properties and low neutron cross section, is being considered as a prospective structural material for nuclear fission and fusion reactors. Additional improvements in SiC mechanical properties may be possible through use of nanocrystalline (nc) SiC. It is also believed that grain boundaries (gb) can act as sinks for point defects created during primary radiation damage, suggesting that shrinking grain size may enhance the material’s radiation resistance. To better understand the connection between radiation damage and nc structure, point defect production in nanocrystalline (nc) SiC with varying grain sizes (5nm, 7nm, 10nm and 12nm) is studied for a Si primary knock on atom with an energy of 4KeV using molecular dynamic simulations. The defect concentrations in the grains are compared with that produced in a single crystal to assess the role of grain size and grain boundaries during the cascade. Comparisons between single and nc SiC include the total defect production, production rates of each defect type, and the size and spatial distribution of the cascade and its damage. This work will thus provide a qualitative understanding on the role of grain size and grain boundaries during low energy primary damage cascades in 3C-SiC.
10:00 AM - V8.3
Lattice Disordering in Ion-Irradiated Nano- and Single-Crystal SiC.
Weilin Jiang 1 , Haiyan Wang 2 , Ickchan Kim 2 , In-Tae Bae 3 , Yanwen Zhang 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , Texas A&M University, College Station, Texas, United States, 3 , State University of New York at Binghamton, Binghamton, New York, United States
Show AbstractDue to its outstanding physical and chemical properties, silicon carbide (SiC) has been considered as a prominent candidate for a variety of applications, including advanced electronic devices and future nuclear energy systems. Extensive experimental and theoretical research efforts have been devoted to the study of irradiation effects in SiC single crystals over the past decades. However, similar studies of nanocrystalline SiC are non-existent until very recently. Because of a large fraction of grain boundaries or interfaces that could serve as strong sinks for mobile point defects produced during irradiation, it is generally believed that nanostructured materials are more resistant to lattice damage. This study employs energetic ion beams for irradiation of single-crystal and nano-crystal 3C-SiC under the identical irradiation conditions at room temperature and above. In addition, 6H-SiC single crystals also have been irradiated to study any polymorph-dependent effects, especially above room temperature. The nanocrystalline 3C-SiC specimens with an average crystallite size on the order of a few nanometers were prepared using pulsed laser deposition. The primary methods for material characterization include ion channeling, x-ray diffraction, and transmission electron microscopy. For irradiation at room temperature, similar disordering behavior has been observed in single-crystal and nano-crystal 3C-SiC; full amorphization occurs at a comparable dose in both materials. This behavior is attributed to the high dose rate and sluggish migration of point defects in SiC at room temperature. Further experiments at 400 K just below the critical temperature for amorphization are currently undertaken and the results will be also presented and discussed.
10:15 AM - V8.4
Elaboration of Zirconium Carbide-Based Materials with Controlled Porosity from the Formulation of Slurries.
Gaetan Martinet 1 2 , Sylvie Foucaud 1 , Alexandre Maitre 1 , Fabienne Audubert 2
1 Laboratoire SPCTS, Faculté des Sciences et Techniques, Limoges France, 2 CEA Commissariat à l'Energie Atomique, DEN/DEC/SPUA/LTEC, Saint Paul Lez Durance France
Show AbstractIn the context of the development of the new nuclear system, the concept selected by the CEA is the Gas-Cooled Fast Reactor system cooling by helium gas. Several nuclear fuel forms have been considered for their potential to operate at very high temperatures and to ensure an excellent retention of fission products. Uranium-Plutonium carbides (U,Pu)C are one of the candidate fuels for Generation IV nuclear plant systems. Within the framework of fission product management, the selected nuclear fuel must have a controlled opened and closed porosity (e.g. pore size, distribution, morphology and volume fraction). Consequently, this study consists in implementing a new ceramic process to elaborate porous nuclear fuels. This work has been focused in the shaping and the sintering of ZrC bodies with controlled porosity. From an experimental point of view, ZrC has been retained as a simulating material because it shows similar physical-chemistry properties then (U,Pu)C. The ceramic process using the slurry precursors because it should allow a better homogeneity and consolidation of the green body microstructure. The wet process suggests the formulation of a suspension of ZrC. The stability conditions of dispersion depend on the choice of the suspension components: solvent, surfactant, porous forming agent, particle size and morphology of the powder of ZrC. The experimental conditions of formulation, casting and consolidation of the green bodies have an impact on the sintering material.The first experimental results consist in showing the role played by the granulometry of the starting ZrC powder in the elaboration of a stable slurry. Indeed, due to its high density (6.7) and its wide dispersion of grain size, it was necessary to reduce the intermediate size of the particles around 4 μm by sieving process. Otherwise, several couples solvent-surfactant were considered. In order to minimize the oxide formation during sintering, non-aqueous additives have been required. In addition, it is necessary to have a viscous solvent to obtain a better stability of ZrC particles in the slurry. The couple Ethylene Glycol–Polyethylene Glycol (400 g.mol-1) has been retained. Among the various fractions of organic additives, the composition which presents the higher stability was obtained for a composition of 15 vol% in ZrC and 3 vol% in surfactant. The suspension was first elaborated using sonotrode which breaks the softest agglomerates, and allows a better distribution of the surfactant in the slurry. It will be noted that a stability of suspension of 3 hours is sufficient considering the shaping stage. The use of casting process with porous mould in plaster was retained for its simplicity of implementation. The compaction rates after casting of the green bodies tend to 46 %. To consolidate the green pieces, two heat treatments were used: the first allows the elimination under nitrogen of the organic residues and the second the densification of material under argon.
V9: Designing Materials for Nuclear Energy I
Session Chairs
Wednesday PM, December 02, 2009
Room 207 (Hynes)
11:15 AM - **V9.1
Materials Research Needs to Advance Nuclear Energy.
Rodney Ewing 1 , Mark Peters 2
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States, 2 Applied Science and Technology, Argonne National Laboratory, Argonne , Illinois, United States
Show AbstractDuring the past several years, there have been a number of workshops, reviews and research programs for the development of new materials for advanced nuclear energy systems. The Office of Science of the U.S. Department of Energy sponsored a workshop in 2006 that resulted in a report, Basic Research Needs for Advanced Nuclear Energy Systems, that outlined a number of high priority research directions: i.) nanoscale design of materials in extreme environments; ii.) physics and chemistry of actinide-bearing materials; iii.) microstructures and properties under extreme conditions; iv.) chemical selectivity at nano- and meso-scales; v.) radiation effects and radiolysis; vi.) thermodynamics and kinetics of nuclear processes; and vii.) predictive multiscale modeling under extreme conditions. Scientific “Grand Challenges” included: i.) physics and chemistry of actinide-bearing materials; ii.) first principles, multi-scale modeling of complex materials under extreme conditions; iii.) the design of molecular systems for chemical selectivity during processing. More recently, the Advanced Fuel Cycle Initiative (AFCI) has matched processing technologies and geological disposal to the design of materials as advanced nuclear fuels and nuclear waste forms. In the latter case, the “extreme” environment is the need to model and extrapolate the behavior of nuclear materials over hundreds of thousands of years. This presentation will provide examples of several of the research topics in each of these areas, as well as discuss cross-cutting research themes that will support the development of the next generation of nuclear materials. An important aspect of the required research programs is the need to develop research facilities for the synthesis, characterization and testing of nuclear materials.
11:45 AM - V9.2
Nanoscale Crossover in Dependence of Radiation Damage Accumulation on Grain Size.
Yi Yang 1 , Hanchen Huang 1 2 , Steven Zinkle 3
1 Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York, United States, 2 Mechanical Engineering, University of Connecticut, Storrs, Connecticut, United States, 3 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractNanostructured materials often behave against conventional wisdom. The reversion of Hall-Petch relationship at nanoscale is a well-known case, giving rise to the crossover in the dependence of mechanical strength on grain size. The authors report a new crossover in the dependence of radiation produced vacancy point defect accumulation on grain size. The crossover is a nanoscale phenomenon, occurring near a critical grain size on the order of 30 nm; and within a temperature window, 15%-25% of the melting temperature. Based on a combination of atomistic simulations and theoretical formulations, the authors also reveal that the nanoscale crossover is a result of competition between two atomic-level mechanisms: grain boundary absorption and bulk recombination of point defects, each of which has a different characteristic length and time scale. A simple metal copper is studied as a model face-centered cubic material and electron radiation is the source of non-cascade defect production, both choices aiming at simplicity for identifying physical mechanisms. The new crossover will likely be a generic nanoscale phenomenon in various materials processing using energetic beams including electrons, ions, and neutrons.
12:00 PM - V9.3
Synthesis and Characterization of Modified Machinable Tantalum Oxide Aerogels for Inertial Confinement Fusion Targets.
Hongbo Ren 1
1 , Research Center of Laser Fusion, Chinese Academy of Engineering Physics, Mianyang China
Show AbstractThe synthesis and characterization of the low-density monolithic tantalum oxide aerogel for Inertial Confinement Fusion (ICF) targets were investigated. The monolithic aerogels were prepared through the sol–gel polymerization of tantalum pentachloride in ethanol using ammonium hydroxide and epichlorohydrin as gelation initiators. A certain functional polymer was used to enhance the mechanic properties of brittle aerogels. The dried tantalum oxide aerogel was characterized by field emission-scanning electron microscopy (FESEM), high-resolution transmission electron microscopy (HRTEM), energy dispersive spectrometry (EDS) and nitrogen adsorption/desorption analyses. The aerogel network was determined to be composed of primary particles with diameter of 1.5 nm. The tantalum oxide aerogel possesses high surface area (835 m2/g) and pore diameters in the micro- and meso-porous range.
12:15 PM - V9.4
Densification of Inert Matrix Fuels Using the Naturally-occurring Material as a Sintering Additive.
Shuhei Miwa 1 , Masahiko Osaka 1
1 , Japan Atomic Energy Agency, Higashi-ibaraki-gun, Ibaraki, Japan
Show AbstractInert matrix fuels (IMFs) with a high content of minor actinides (MAs) are currently considered as one promising option for the rapid incineration of MAs in a future fast reactor cycle system. On a related R&D of IMFs, we proposed a new concept for densification of IMFs with Molybdenum (Mo) and magnesium oxide (MgO) by using the waste of asbestos as a sintering additive. This concept should contribute especially to the effective utilization of resources and protection of public safety. In this concept, magnesium silicates, which are formed by the decomposition of asbestos in low temperature heat-treatment, are used as a sintering additive for the achievement of high-performance IMFs having no defects, a high density, and a homogeneous dispersion of MAs oxides host phase. In this study, effects of magnesium silicates additives on densification of component materials of IMFs, i.e. host phase and inert matrix, were experimentally investigated for the purpose of establishing a sophisticated fabrication procedure based on the powder metallurgy. CeO2-x was chosen as a representative of MAs oxides for the host phase. Sintering tests of component materials of IMFs containing various magnesium silicate, i.e. forsterite (Mg2SiO4) and enstatite (MgSiO3), were carried out at 1473 – 1873 K. The densification behaviors were characterized by the density, microstructure and hardness. The densities of MgO sintered at 1873 K increased with only 1wt.% additives of silica (SiO2), Mg2SiO4 and MgSiO3, and showed above 95 %TD. The densities were independent on the amount of additive up to 10 wt.%. The densities of CeO2-x sintered at 1673 K were also increased with only 1 wt.% additives of SiO2, Mg2SiO4 and MgSiO3, and showed about 95%T.D. The densities were decreased with increasing the amount of additives above 1 wt.%. It should be noted that the sign of the liquid phase formation was observed for CeO2-x at 1673 K with the additives of 5wt.% SiO2 and 5 wt.% MgSiO3. This result indicated that the eutectic reaction of Ce-Si-O system would occur below 1673 K. From this result, there is a high possibility that the liquid phase would be also formed in Am-Si-O and Pu-Si-O system at low temperature from the similarity of thermochemical properties of CeO2-x with those of PuO2-x and AmO2-x. On the other hand, the sintered densities of Mo sintered at 1873 K showed little change with the additives. The present results have shown that magnesium silicates additives are effective for the densification of MgO and CeO2-x. Therefore, it is believed that these additives should make MgO based IMFs dense by a simple way with a relatively low sintering temperature. In addition, there is also possibility that the densification of Mo based IMFs would be enhanced by the formation of liquid phase of host phase with a relatively low sintering temperature.
12:30 PM - **V9.5
Nanostructured Ceramic Materials: From Powder Synthesis to Preparation of Dense or Porous Ceramic Architectures.
Christian Guizard 1
1 LSFC UMR 3080, CNRS/SAINT-GOBAIN, Cavaillon France
Show AbstractDevelopments in advanced nuclear energy systems can benefit from the latest advances in materials design down to the nanometer scale, in particular for ceramic materials. The general advantages of engineered ceramics such as alumina, silicon nitride, silicon carbide and zirconia, in comparison with steel are light weight, chemical and thermal stabilities at elevated temperature and excellent wear resistance. Some of these ceramics are already under investigation in view of nuclear applications both as structural materials or inert matrices. The intrinsic properties of ceramics are due to the strong chemical bonds involved in their inner structure, although it also leads to unreliable mechanical properties responsible for brittle failure. Interestingly, significant improvements are expected from nanostructured ceramics because bulk ceramic materials with grain sizes less than 100 nm exhibit novel mechanical and physical properties as compared with their microcrystalline counterparts. This presentation will consider three current research areas in our laboratory related to the preparation of nanostructured ceramics. The first one deals with the synthesis in supercritical fluids of high specific surface area crystalline powders which are a prerequisite of the fabrication of nanostructured ceramics. Secondly, sintering methods able to preserve a nanosized grained structure during thermal consolidation will be discussed, in particular the SPS (Spark Plasma Sintering), which recently has been used for superfast densification of ceramic powders by simultaneous application of pulsed high dc current densities and load. Finally recent developments in the field of new ceramic architectures obtained by freeze-casting will be presented. The technique consists of freezing a liquid suspension (aqueous or not), followed by sublimation of the solidified phase from the solid to the gas state under reduced pressure, and subsequent sintering to consolidate and densify the walls. A hierarchic porous structure can be obtained, with unidirectional channels in the case of unidirectional freezing, where macropores are a replica of the solvent crystals. Such ceramics can be engineered to combine several advantages inherent from their architecture: they are lightweight, can have open or closed porosity making them useful as insulators or filters, can withstand high temperatures and exhibit high specific strength, in particular in compression.
V10: Structural Complexity of Nuclear Fuels
Session Chairs
Wednesday PM, December 02, 2009
Room 207 (Hynes)
2:30 PM - V10.1
Properties of Vacancy Defects Induced in UO2 by Irradiation and Probed by Using Positron Annihilation Spectroscopy.
Marie-France Barthe 1 , Stephanie Leclerc 1 , Laszlo Liszkay 1 , Moineau Virginie 1 , Hicham Labrim 1 , Pierre Desgardin 1 , Catherine Corbel 2 , Gaelle Carlot 3 , Philippe Garcia 3
1 CEMHTI, CNRS, Université d'Orléans, Orléans France, 2 LSI, UMR 7642 CEA - CNRS - Ecole Polytechnique, Palaiseau France, 3 LLCC, DEN/DEC/SESC, CEA Cadarache, Saint Paul lez Durance France
Show AbstractThe 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 if it is used as a nuclear waste storage matrix. Uranium dioxide is a major component of nuclear fission fuel and is used as a model material to study the behavior of nuclear fuel. The behavior of UO2 under irradiation has been extensively studied by using different techniques such as Channeling Rutherford Backscattering, RX diffraction and so on, but only very few studies have been focused on the direct observation of point defects and the determination of their properties. In this work, we have used positron annihilation spectroscopy (PAS) to determine annihilation characteristics in UO2. Both 22Na based positron lifetime spectroscopy (PALS) and coincidence Doppler annihilation-ray broadening spectrometry (CDBS) have been used to characterize the vacancy defects induced by irradiation in sintered UO2 disks that have been polished and annealed at high temperature (1700°C/24h/ArH2) and alpha uranium polycrystalline samples. The UO2 disks have been irradiated with electrons (1 MeV and 2.5 MeV) at LSI (Palaiseau) and alpha particles (45 MeV) by using the cyclotron at CEMHTI (Orléans) with fluences in the range from 1x1016 cm-2 to 1x1019 cm-2. After irradiation, SPBDB and PALS measurements show the formation of vacancy defects after a 2.5 MeV electrons or 45 MeV alpha irradiation with a lifetime of τ= 307±3 ps whereas no defects are detected for an irradiation with 1MeV electrons. The nature of these defects and their positron annihilation characteristics will be discussed in comparison with the results obtained in alpha uranium.
2:45 PM - V10.2
Thermochemical Modeling of High Burnup, Transuranic Gas-Cooled Reactor Fuel.
Theodore Besmann 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractParticulate nuclear fuel in thye TRISO configuration is being considered for utilizing and eliminating excess plutonium and related transuranics in a modular helium reactor. This concept will thus require extremely high fuel burnups to be efficient, and therefore challenge the fuel with regard to maintaining integrity in-reactor. It particular, issues such as kernel migration where carbon in the buffer layer and inner pyrolytic carbon layer transports from high to low temperature volume in the particle, become important to assess.. Gettering agents have been found to mitigate this problem and the addition of SiC or ZrC for that purpose is analyzed. The thermochemical analysis predicts oxygen potential behavior in the fuel to burnups of 50% FIMA with and without the presence of oxygen gettering SiC and ZrC and relates that to effects on potential particle integrity.This work was funded under the U.S. Department of Energy - NE Deep Burn program with Oak Ridge National Laboratory under contract DE-AC05-00OR22725 with UT Battelle, LLC.
3:00 PM - V10.3
TEM Characterization of As-Fabricated Dispersion Fuels.
Jian Gan 1 , Dennis Keiser 2 , Brandon Miller 3 , Jan-Fong Jue 2 , Daniel Wachs 2 , Todd Allen 3
1 Basic Fuel Properties and Modeling, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Fuel Performance and Design, Idaho National Laboratory, Idaho Falls, Idaho, United States, 3 Engineering Physics, University of Wisconsin, Madison, Wisconsin, United States
Show AbstractThe United States nuclear fuels program on Reduced Enrichment Research and Test Reactors (RERTR) is to develop low enrichment fuels (< 20%U235) to replace the highly enriched fuels used in the research and test reactors for nuclear nonproliferation. Dispersion type plate fuels are popular fuels used in many research and test reactors worldwide. A typical dispersion fuel plate is about 1.5 mm thick consisting of three layers with the outer layers of aluminum cladding and the middle layer of aluminum alloy dispersed with U-xMo (x=7-10 in wt%) fuel particles. There is an interaction layer formed at the interface of fuel particle and Al alloy matrix as a result of fuel fabrication. The radiation stability of this layer could strongly affect the fuel performance in the reactor. This work reports the microstructural characterization using TEM on two batches of dispersion fuels (U-7Mo dispersed in Al-2Si matrix) through different fabrication process. The RERTR-9A dispersion fuel was fabricated using roll-bonding followed by a Hot Isostatic Pressing (HIP) step. The RERTR-9B dispersion fuel was fabricated using just roll-bonding. A 1.0 mm diameter small fuel punching was used for TEM preparation. Preliminary TEM results show that significant fraction of original γ-(U, Mo) (bcc) is transformed to α-U (Orthorhombic) and γ’-(U2Mo) in the HIP processed sample, resulting in development of interaction layer deep into the fuel particles. This was not observed in the sample that was just roll-bonded. In both cases, most part of interaction layers consists nanocrystalline (U, Mo)(Al, Si)3 phase. Precipitates with a composition of approximately 77Al-13Fe-10Si are found in the Al alloy matrix. The implication of these observed microstructure on the fuel irradiation performance will be discussed.
3:15 PM - V10.4
SEM Characterization of U-Mo Dispersion Fuels Irradiated in the Advanced Test Reactor.
Dennis Keiser 1 , Jan-Fong Jue 1 , Adam Robinson 1 , Pavel Medvedev 1
1 , Idaho National Laboratory, Scoville, Idaho, United States
Show AbstractThe Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low enriched U-Mo fuels for use in reactors that currently employ fuels containing highly enriched uranium. As part of this development, U-Mo fuel plates are being irradiated in the Advanced Test Reactor and then characterized to determine the microstructural development during irradiation. This paper will describe recent results of scanning electron microscopy characterization that has been performed on irradiated fuel plates. Past work has focused on the behavior of dispersion fuels that contain U-7Mo particles, but this talk will report the first SEM characterization results for irradiated dispersion fuels that contain U-10Mo particles. The fuel plates with U-10Mo particles exhibit different microstructural evolution during irradiation compared to what has been observed with irradiated U-7Mo dispersion fuels. In particular, more interaction between the fuel and matrix is observed. There appears to be a link between the starting microstructure of the fuel plate after fabrication, which can depend on the composition of the fuel particles, and the performance of the fuel plates during irradiation.
3:30 PM - V10.5
Neutron Diffraction Study of the Structural Changes Occurring During the Low Temperature Oxidation of UO2.
Gianguido Baldinozzi 1 2 , Lionel Desgranges 3 , Gurvan Rousseau 3 1 2
1 MFE, SPMS Lab, CNRS Ecole Centrale Paris, Chatenay-Malabry France, 2 MFE, DMN/SRMA/LA2M, CEA Saclay, DEN, Gif-sur-Yvette France, 3 DEC/SESC/LLCC, CEA Cadarache, DEN, St. Paul-lez-Durance France
Show AbstractThe oxidation of uranium dioxide has been studied for more than 50 years. It was first studied for fuel fabrication purposes and then later on for safety purposes to design a dry storage facility for spent nuclear fuel that could last several hundred years. Therefore, understanding the changes occurring during the oxidation process is essential, and a sound prediction of the behavior of uranium oxides requires the accurate description of the elementary mechanisms on an atomic scale. Only the models based on elementary mechanisms should provide a reliable extrapolation of laboratory results over timeframes spanning several centuries. The oxidation mechanism of uranium oxides requires accurately understanding the structural parameters of all the phases observed during the process. Uranium dioxide crystal structure undergoes several modifications during the low temperature oxidation which transforms UO2 into U3O8. The symmetries and the structural parameters of UO2, β-U4O9, β-U3O7 and U3O8 were determined by refining neutron diffraction patterns on pure single-phase samples. Neutron diffraction patterns, collected during the in situ oxidation of powder samples at 483 K were also analyzed performing Rietveld refinements. The lattice parameters and relative ratios of the four pure phases were measured during the progression of the isothermal oxidation. The transformation of UO2 into U3O8 involves a complex modification of the oxygen sublattice and the onset of complex superstructures for U4O9 and U3O7, associated with regular stacks of complex defects known as cuboctahedra which consist of 13 oxygen interstitial atoms. The structural modifications and kinetics of the oxidation process are discussed. The results obtained in this study provide a comprehensive structural description of the transformation of UO2 into U3O8 at temperatures below 700 K and a sound structural basis for the use of two different oxygen diffusion coefficients in U4O9 and U3O7.
3:45 PM - V10.6
Characterization of Mixed Oxide Fuels.
Tarik Saleh 1 , Daniel Schwartz 1 , Franz Freibert 1 , Fredrick Hampel 1 , Jeremy Mitchell 1 , Stephen Willson 2
1 Nuclear Materials Science, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Actinide and Fuel Cycle Technology Group, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractCurrently, Los Alamos National Laboratory is engaged in producing mixed actinide (i.e., U, Np, Pu, and Am) oxides as a participant in an international collaboration to study candidates for nuclear fuels. Critical to understanding and predicting the performance of these fuels is the correlation of composition and processing technique with initial morphology, crystallographic structure and thermal and physical properties. In this presentation, we will communicate the results of characterization of fuels, ranging in actinide composition from U0.8Pu0.2 to U0.75Np0.02Pu0.2Am0.03, from recently fabricated fuel candidates. Results from ceramography, X-ray diffraction, dilatometry, resonant ultrasound spectroscopy, immersion density and mechanical testing will be discussed.
V11: Designing Materials for Nuclear Energy II