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
Session Chairs
Wednesday PM, December 02, 2009
Room 207 (Hynes)
4:30 PM - **V11.1
Radiation Tolerance and Disorder - Can They Be Linked?
Karl Whittle 1
1 Institute of Materials Engineering, ANSTO, Menai, New South Wales, Australia
Show AbstractThe future expansion of nuclear power provides materials challenges that are not easily overcome, for example the safe immobilisation of nuclear waste is an important component in any future expansion of nuclear power. The use of ceramic-based materials, as opposed to borosilicate glasses, is now being investigated widely. The ability of ceramics to be tailored to a specific waste stream is now understood and obtainable quickly and with minimal cost. A second component that limits the expansion of fission-based technologies is the development of materials that are not only tolerant of radiation damage, but are also capable of retaining mechanical strength at high temperatures.One concern for any material however, is the effect of radiation damage, primarily from alpha-decay damage, which in many systems can transform the material from crystalline to amorphous. The effects of such radiation damage on both the structural and chemical properties can range from trivial to critical, for example volume expansion and are the primary focus of much research. As part of a long-term research programme the effects on radiation tolerance of a range of ordered and disordered materials are presented, along with explanations for stability and their use in future nuclear reactors.
5:00 PM - V11.2
Irradiation of Metallic Nano-Foams.
Eduardo Bringa 3 , Joshua Monk 1 , Alfredo Caro 2 , Diana Farkas 1
3 CONICET & Instituto de Ciencias Básicas, Universidad Nacional de Cuyo, Mendoza Argentina, 1 , Virginia Tech, Blacksburg, Virginia, United States, 2 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractMetallic foams can be thought of as candidate materials for application in radiation environments. We present atomistic simulations of the behavior of irradiated nano-scale foams. We find that the large fraction of surfaces allows the formation of dislocations and twin boundaries which are absent in bulk simulations. This additional dislocation and twin density leads to hardening during simulated tension tests. Radiation damage is extended over scales much larger than in bulk, because of recoils being sputtered from foam filaments. For filaments with a size comparable to the cascade size, cumulative cascades decrease the foam surface area indicating high radiation sensitivity.
5:15 PM - V11.3
Nano-scale Materials Design of Pyrochlore for Enhanced Radiation Performance.
Jie Lian 1 , Jiaming Zhang 2 , Antonio Fuentes 3 , Rodney Ewing 2
1 , Rensselaer Polytechnic Institute, Troy, New York, United States, 2 , University of Michigan, Ann Arbor, Michigan, United States, 3 , Universidad de Zaragoza, Zaragoza Spain
Show AbstractNano-scale design strategy is important for developing advanced materials with enhanced performance for nuclear engineering applications. High density of surfaces and grain boundaries of nanostructured materials may behave as sinks for defect annealing and thus mitigate radiation damage; however, radiation tolerance is not inherent for nanostructured materials. Based on systematic energetic beam bombardment studies, we demonstrated that the behavior of pyrochlore compounds, potential host phases for actinide incorporation, is intriguing when approaching nano-scale regime. The interplay of among composition, crystal size, bond nature and degree of disordering defines the structural deviation and defect energetics that may essentially control phase stability of materials upon radiation damage. The integration of high energy ball milling, a highly non-equilibrium materials processing approach, and thermal treatment allows us control the crystal size, crystallinity and structural disordering of nano-scale pyrochlores and offers a new dimension of materials design for extreme radiation environment.
5:30 PM - V11.4
Production of Layered Double Hydroxides for Anion Capture and Storage.
Jonathan Phillips 1 , Luc Vandeperre 1
1 Department of Materials, Imperial College London, London United Kingdom
Show AbstractTechnetium has a long half life of up to 4.2x106 years. It is separated from liquid waste streams with tetraphenylphosphonium bromide, producing a solution containing the pertechnetate anion, TcO4-. Pertechnetate is highly mobile in groundwater and it is therefore highly desirable to capture and immobilise this anion within a solid for interim and ultimately long term storage. Layered Double Hydroxide (LDH) materials are known to possess excellent anion sorption capabilities due to their structure which consists of ordered positively charged sheets intercalated with interchangeable hydrated anions. The composition can be tailored to produce suitable precursors for ceramic phases by varying the divalent and trivalent cations and the anions. LDHs with the general formula Ca1-x (Fe1-y, Aly)x (OH)2 (NO3)x . nH2O were produced by a co-precipitation method from a solution of mixed nitrates. Calcination leads to the formation of Brownmillerite Ca2(Al,Fe)2O5 like compounds for temperatures as low as 400°C, this is close to the lowest temperature at which Tc is known to volatilise (310.6 °C Tc2O7). It was shown that after calcining up to 600°C, the LDH structure is recovered in water allowing rapid ion capture to occur. This suggests that these materials have potential for both capture and as a storage medium for Tc.
5:45 PM - V11.5
Low Temperature Synthesis of Silicon Carbide Inert Matrix Fuel.
Chunghao Shih 1 , Ronald Baney 1 , James Tulenko 2
1 Materials Science and Engineering, University of Florida, Gainesville, Florida, United States, 2 Nuclear and Radiological Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractSilicon carbide (SiC) is one of the prime candidates as the matrix material in inert matrix fuels (IMF) being designed to reduce plutonium inventories and the long half-life actinides through transmutation. However, temperatures above 1700 oC are generally required to achieve a high density SiC pellet even with oxide sintering aids. Lower temperature sintering without the addition of significant amount oxide sintering aids is desirable.To decrease the sintering temperature of SiC, 1 micron SiC powder and a commercially available SiC polymer precursor, allylhydridopolycarbosilane (AHPCS), were used to fabricate the SiC pellet. After several polymer infiltration and pyrolysis cycles, 80% dense SiC pellets were achieved with a processing temperature of 1150 oCMoreover, SiC IMF with full density and 80% density were simulated and compared with traditionally used mixed oxide fuel (MOX) by the neutronics code CASMO, a muti-group two-dimensional nuclear assembly reactivity code. The reactivity versus depletion of the SiC IMF with 80% density was almost identical to the burnup of the SiC IMF with full density SiC.
Symposium Organizers
Gianguido Baldinozzi CEA-CNRS-ECP
Yanwen Zhang Pacific Northwest National Laboratory
Katherine L. Smith Embassy of Australia
Kazuhiro Yasuda Kyushu University
V12: Structural Complexity in Advanced Nuclear Fuels
Session Chairs
Thursday AM, December 03, 2009
Room 207 (Hynes)
9:30 AM - V12.1
Stability and Migration Mechanisms of Volatile Impurities in Uranium Carbide by First-Principles Calculations.
Michel Freyss 1 , Boris Dorado 1 , Marjorie Bertolus 1
1 DEC/SESC/LLCC, CEA, DEN, Cadarache, Saint-Paul lez Durance France
Show AbstractThe scope of this study is to shed light on the behavior of volatile elements in nuclear materials by means of first-principles calculations. We focus here on the uranium monocarbide UC as a first step in the study of the mixed carbide (U, Pu)C, which is a potential fuel for Generation IV nuclear reactors. The study of volatile elements is of interest in order to get some comprehension in the behavior at the atomic scale of the material under irradiation, but also in order to improve the fuel performances. In particular, the goal of our work here is to study the propensity of some volatile elements to be trapped by point defects or to diffuse in the material.The first part of the presentation is devoted to the first-principles modeling of bulk uranium carbide and to the problem that arises from the use of the GGA (Generalized Gradient Approximation) or the GGA+U approximations in the description of the 5f electron correlations.The second part is devoted to the behavior of volatile impurities in UC. Various types of impurities are considered: rare gases and iodine (volatile fission products), helium (created by alpha decays) and oxygen (incorporated by oxidation). The location in the UC lattice and the migration mechanisms of those elements are investigated. The changes they induce in the structural and the electronic properties of the crystal are also discussed, together with role that point defects play in the trapping of those impurities.All calculations are done using the first-principles PAW method (Projector Augmented Waves) based on the DFT and as implemented in the VASP code.
9:45 AM - V12.2
Ab initio Modelling of Incorporation and Migration of Volatile Fission Products in Silicon Carbide.
Marjorie Bertolus 1 , Yannis Major 1 , Boris Dorado 1
1 DEN, DEC/SESC/LLCC, CEA, Saint-Paul-lez-Durance France
Show AbstractDuring in-reactor irradiation actinide fission produces large quantities of volatile fission products, which have a significant influence on the structural and mechanical properties of nuclear fuels. It is therefore essential to get further insight into the behavior of these elements in materials to understand the behavior of fuel systems and improve their performance. The incorporation sites and activation energies determine the mobility in the material, as well as the influence of temperature and defects on this mobility. It is then of major importance to evaluate these parameters. We will present an investigation using Density Functional Theory of the incorporation of krypton, xenon, iodine and cesium several phases of silicon carbide (SiC), which are considered as potential constituents for nuclear fuel of Generation IV future reactors. We have calculated the incorporation energies of the various fission products considered in various sites of stoichiometric or near-stoichiometric cubic and hexagonal SiC crystals. We have also determined several migration pathways between these sites, as well as the activation energies associated. The evaluation of these energies enables us to get further insight into the atomic scale mechanisms involved in the migration and diffusion of Kr, Xe, I and Cs in silicon carbide.
10:00 AM - V12.3
Fully-coupled Engineering and Mesoscale Simulations of Thermal Conductivity in UO2 Fuel using an Implicit Multiscale Approach.
Michael Tonks 1 , Glen Hansen 1 , Derek Gaston 1 , Cody Permann 1 , Paul Millett 1 , Dieter Wolf 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractThough the thermal conductivity of solid UO2 is well characterized, its value is sensitive to microstructure changes. In this study, we propose a two-way coupling of a mesoscale phase field irradiation model to an engineering scale, finite element calculation to capture the microstructure dependence of the conductivity. To achieve this, the engineering scale thermomechanics system is solved in a parallel, fully-coupled, fully-implicit manner using the preconditioned Jacobian-free Newton Krylov (JFNK) method. Within the JFNK function evaluation phase of the calculation, the microstructure-influenced thermal conductivity is calculated by the mesoscale model and passed back to the engineering scale calculation. Initial results illustrate quadratic nonlinear convergence and good parallel scalability.
10:15 AM - V12.4
Effect of Americium and Simulated Fission Products Addition on Oxygen Potential of Uranium-Plutonium Mixed Oxide Fuels.
Kosuke Tanaka 1 , Masahiko Osaka 1 , Ken Kurosaki 2 , Hiroaki Muta 2 , Masayoshi Uno 3 , Shinsuke Yamanaka 2
1 , Japan Atomic Energy Agency, Oarai, Ibaraki Japan, 2 , Osaka University, Suita, Osaka Japan, 3 , Fukui University, Fukui Japan
Show AbstractLow decontaminated mixed oxide (MOX) fuel, which contains several percent of minor actinides (MAs) and fission products (FPs), is a promising candidate for a closed nuclear fuel cycle system based on a fast reactor. Extending burn-up of the fuel has been identified as a practical means of improving the economics of the system. In high burn-up oxide fuels, some FPs dissolve in the fuel matrix and others form oxide or metallic precipitates, which would affect the physical and chemical properties of the fuels. It is, therefore, of crucial to understand the effect of FPs accumulation on the fuel performance. Oxygen potential is an important property for oxide nuclear fuels and is closely related to cladding inner wall corrosion in the high burn-up fuels. In order to investigate the effect of MAs and FPs addition on oxygen potential of MOX fuels, thermogravimetric analyses (TGA) were carried out. MOX fuel containing 3% americium (Am), which represents the MAs, and 1.5 % FPs was defined as a low decontaminated MOX fuels in this study. Based on the composition of the fuel, three simulated compositions were made, representing the burn-ups of 0, 150 and 250 GWd/t. The ORIGEN2 code was used to calculate the compositions. The samples were prepared by a conventional powder metallurgical route in a glove box. Appropriate amounts of UO2, PuO2 containing AmO2 and 26 kinds of simulated FP elements such as Nd, Ce, La, Ba, Sr, Zr, Ru, Mo, Pd were weighed and thoroughly mixed in an agate mortar with a pestle in acetone medium. The mixed powder was then compacted into a columnar pellet by a uni-axial pressing unit followed by the sintering in reducing atmosphere. After crushing the pellet, the specimens were subjected to TGA. Prior to the TGA, sintered pellets were characterized by ceramography, X-ray diffraction (XRD) analysis and electron probe microanalysis (EPMA). The results revealed that the rare earth and other oxides were dissolved in the matrix and small metallic and gray phase oxide precipitates were distributed throughput the pellets. The size and the number of these precipitations increase with increasing the simulated burn-ups. No sign of phase formation without a fluorite-type structure, however, was found by XRD analysis. The oxygen potentials of the all specimens, which were simulated the burn-up level of 0, 150 and 250 GWd/t, were increased with increasing temperature. The oxygen potentials at the each temperature also increased with increasing simulated burn-up levels. These results would be useful for evaluating the performance of the high burn-up low decontaminated MOX fuels in the fast reactors. Present study includes the result of “Study on the physical properties of nuclear fuels with multi phase system: Toward establishment of the closed cycle system with low-decontaminated oxide fuel” entrusted to Osaka University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
10:30 AM - V12: Fuels 2
V12.5 TRANSFERRED TO V16.44
Show AbstractV13: Modelling Complex Materials II
Session Chairs
Thursday PM, December 03, 2009
Room 207 (Hynes)
11:15 AM - **V13.1
Radiation Damage and He Mobility in Plutonium.
Roger Smith 2 , Marc Robinson 2 , Steven Kenny 2
2 , Loughborough University, Loughborough, Leicestershire, United Kingdom
Show AbstractThe work will examine the role of interfaces in radiation damage simulations. Two examples will be considered. In the first example grain boundaries in Fe will be investigated. Both cascade interactions and the role of P atoms which can segregate there to cause embrittlement will be considered. In the second example an investigation into the MgO – HfO2 interface, using classical molecular dynamics, and fixed charge potentials is undertaken. This composite system is representative of a dispersion nuclear fuel form concept being investigated for its potential in easing separations and reprocessing. HfO2 is used as a surrogate for UO2, for experimental comparison, while MgO represents a chemically separable host matrix. Computer simulations, using molecular dynamics, have been carried out on three separate models in an attempt understand this behaviour. The first model looks at single ion (Au) bombardment of the interface. Cascades involving a cluster of Argon atoms are also investigated. Finally we create a model of delamination at the interface.
11:45 AM - V13.2
Cascade-Driven Mixing at Metal Oxide-Metal Oxide Interfaces.
Steve Valone 1
1 Materials Science and Technology Division, LANL, Los Alamos, New Mexico, United States
Show AbstractSome advanced nuclear fuel concepts are based on composites of fissile and nonfissile materials. One example might be uranium oxide embedded in an inert matrix material such as magnesium oxide. Of necessity such a composite contains metal oxide-metal oxide interfaces. The concept behind such designs is that the nonfissile phase will capture a significant fraction of the fission products and minor actinides [1]. The success of such strategies will depend in some measure on the stability of the fissile-nonfissile material interfaces. During operation in a nuclear reactor, fission track damage will cross these interfaces. As one element in investigating the stability of these interfaces comes from nano-scale simulations. Within this realm there is the Rare Event Enhanced Domain following Molecular Dynamics (REED-MD) simulations of Jensen et al. [2] that models the deposition of energy along a fission track, as well as the more standard SRIM/TRIM calculations [3]. Here we focus on cascade damage as secondary events from a fission track event to gauge the levels and types of damage that may be simulated through molecular dynamics. Some of that damage will take place at the interface between phases. As a model system, experiments consider a composite in which the nonfissile material is magnesia and the fissile phase is modeled via hafnia as a surrogate. To coordinate with the materials in the experiments, simulation cells are composed of hafnia in the fluorite structure and magnesia in the rocksalt structure. Molecular dynamics simulations of cascade damage across interfaces of these materials shows Hf cations becoming kinetically trapped in the magnesia phase at ambient pressure and a temperature of 500 K. We use potentials for both phases due to Grimes and Catlow [4]. The Hf cations remained trapped for at least the 20-ps duration of the simulations. When the primary-knock-on atom energy is above a few hundred eV in the direction of the interface and is within five lattice spacings, the propensity for trapping is very high. Under these same conditions, an Mg cation will occasionally become trapped in the hafnia. However, at these levels of damage and thermal conditions, the interface remains largely intact.
[1] M. Osaka, S. Miwa, and Y. Tachi, Ceramics International 32, 659 (2006).
[2] B. Jeon and N. Grønbech-Jensen, Comput. Phys. Comm. 180, 231 (2009).
[3] See http://www.srim.org/.
[4] R. W. Grimes and C. R. A. Catlow, Phil. Trans.: Phys. Sci. Engin. 335, 609 (1991).
12:00 PM - V13.3
How to Simulate the Microstructure Induced by a Nuclear Reactor with an Ion Beam Facility : DART.
Laurence Luneville 1 , David Simeone 2
1 CEA/DEN/DANS/DM2S/SERMA-MFE, CEA, gif sur Yvette France, 2 CEA/DEN/DANS/DMN/SRMA-MFE, CEA, Gif sur yvette France
Show AbstractEven if the Binary Collision Approximation does not take into account relaxation processes at the end of the displacement cascade, the amount of displaced atoms calculated within this framework can be used to compare damages induced by different facilities like pressurized water reactors (PWR), fast breeder reactors (FBR), high temperature reactors (HTR) and fusion facilities like the European Spallation Source (ESS) and the International Fusion Material Irradiation Facility (IFMIF) and ion beam facilities on a defined material. In this presentation, a formalism is presented to describe the neutron–atom cross-section and primary recoil spectra taking into account the anisotropy of nuclear reactions extracted from nuclear evaluations. Such a formalism permitted to calculate a displacement per atom rate as well as primary and weighted recoil spectra. Such spectra provide useful information to perform realistic experiments in ion beam facilities.
12:15 PM - V13.4
Energetics of Fission Products in Uranium Dioxide from Electronic-structure Calculations.
Pankaj Nerikar 1 2 , Xiang-Yang Liu 1 , Blas Uberuaga 1 , Chris Stanek 1 , Susan Sinnott 2 , Simon Phillpot 2
1 MST-8 Structure/Property Relations, Los Alamos National Lab, Los Alamos, New Mexico, United States, 2 Materials Science and Engineering Department, University of Florida, Gainesville, Florida, United States
Show AbstractThe stabilities of selected fission products – Xe, Cs, and Sr – are investigated as a function of non-stoichiometry x and temperature in uranium dioxide (UO2). In particular, density functional theory (DFT) is used to calculate the incorporation and solution energies of these fission products at the anion and cation vacancy sites, at the divacancy, and at the bound Schottky defect. Correlation effects are taken into account within the DFT+U formalism. In general, defects with greater charge are found to be more soluble in the fuel matrix and the solubility of fission products increases as the hyperstoichiometry increases. The solubility of fission product oxides is also explored. Cs2O is observed as a second stable phase and SrO is found to be soluble in the UO2 matrix for all stoichiometries. These studies can be extended to include other fission products and microstructural features such as grain boundaries and can lead to a fundamental understanding of nuclear fuel phenomena.This work was supported in part by the DOE-BES Computational Materials Science Network.
12:30 PM - **V13.5
Atomic Scale Simulations of Fission Gas Migration in Nuclear Fuel.
David Parfitt 1 , Clare Bishop 1 , Mark Wenman 1 , Robin Grimes 1
1 Materials, Imperial College London, London United Kingdom
Show AbstractThe evolution and release of fission gases (predominantly krypton and xenon) is one of the limiting factors for the length of time nuclear fuel can spend in a reactor. Such gases exist in environments from isolated atomic species through to large scale gas bubbles. We present here the use of atomic scale models in examining the equilibrium concentrations and migration rates of fission gases when associated with various imperfections within the crystal. Specifically, we examine how atomic scale models contribute to the understanding of clustering and migration rates of isolated atomic species. We examine the enhanced diffusion rates of fission gases that can occur around and along several types of dislocation cores. Additionally, we consider the equilibrium properties and re-solution rates from large scale gas bubbles that exist within the fuel.
V14: Metallic Materials III
Session Chairs
Thursday PM, December 03, 2009
Room 207 (Hynes)
2:30 PM - **V14.1
Deformation Induced Atomic-scale Microstructure Modification in Irradiated Metals.
Yury Osetskiy 1 , Roger Stoller 1 , Dmitry Terentyev 2 , David Bacon 3
1 Materials Science & Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States, 2 Reactor Materials Research Unit, SCK CEN, Mol Belgium, 3 Department of Engineering, University of Liverpool, Liverpool United Kingdom
Show AbstractAccumulation of radiation-induced microstructure leads to specific behavior of irradiated materials under deformation such as hardening, strengthening, loss of ductility, plastic instability, etc. In addition, significant microstructure modification may occur during deformation. Many of these phenomena are controlled by atomic-scale mechanisms of interactions involving moving dislocations and radiation-induced defects. The reactions between dislocations and lattice defects also affect microstructure accumulation at a particular level of plastic deformation during irradiation.We present here a review of results of extensive atomic-scale modeling devoted to the study of the interaction between moving screw and edge dislocations in bcc and fcc metals and irradiation-specific defects such as voids, He-filled bubbles, stacking fault tetrahedra and interstitial dislocation loops. We consider reactions between dislocations and individual defects as well as with groups of different defects. Microstructure changes due to these reactions involve a spectrum of effects, including complete elimination or restoration of defects, their mutual recombination, and change of size and/or structure (shear, Burgers vector change, phase transformation, etc.). We show how the reactions may be classified and discuss them in the light of experimental observations.
3:00 PM - V14.2
Energy Landscape of Small Defect Clusters in bcc Fe.
Mihai-Cosmin Marinica 1 , Francois Willaime 1
1 SRMP, CEA, Gif-sur-Yvette France
Show AbstractThe mobilities of self-interstitial atoms (SIA) and their clusters in metals, especially body-centered cubic (bcc) metals, are one of the main issues in multiscale models for the prediction of the microstructure evolution that these materials undergo under irradiation. In iron, where dumbbells in clusters may have <110>, <111> or <100> orientations, this question is particularly challenging because the number of possible configurations increases rapidly with the number of defects in the cluster. Configurations made of non-parallel dumbbells and with a reduced mobility have been recently identified from high temperature molecular dynamics simulations. Their stability has been confirmed by DFT calculations [1]. These results showed that non-conventional configurations and finite temperature effects must be taken into account. We propose to address these two points more thoroughly using on the one hand an improved version of the activation relaxation technique nouveau [2,3] (ARTn), an eigenvector following method for systematic search of saddle points and transition pathways on a given potential energy surface, and on the other hand lattice dynamics calculations [4]. Using the Ackland-Mendelev EAM potential for iron, we have determined the formation energies of all bonded configurations of clusters containing up to 5 SIAs. For the most stable ones, we have identified their migration mechanism. This shows in particular that some configurations with low saddle point energies have to be considered in the kinetics of the system, although they are not the most stable ones. Lattice dynamics calculations show that at high temperature configurations with <111> dumbbells and/or non-parallel dumbbells are favoured. The low frequency modes at the origin of this stabilisation driven by vibrational entropy are analyzed. DFT calculations performed using the SIESTA code are used to assess the main findings on new low energy configurations and on the low frequency modes.1. D.A Terentyev, T.P.C Klaver, P. Olsson, M.-C. Marinica, F. Willaime, C. Domain, L. Malerba, Phys. Rev. Lett. 100 (2008)145503.2. G.T. Barkema, N. Mousseau, Phys. Rev. Lett. 77 (1995) 4358.3. E. Cances, F. Legoll, M.-C. Marinica, K. Minoukadeh, F. Willaime J. Chem. Phys. 130 (2009) 114711.4. M.-C. Marinica, F. Willaime, Solid State Phen. 129 (2007) 67
3:15 PM - V14:MetalliMat 3
V14.3 TRANSFERRED TO V7.0
Show Abstract3:30 PM - V14.4
Atomic Mechanisms of Delocalized Point Defect Migration in CuNb Interfaces.
Kedarnath Kolluri 1 , Michael Demkowicz 1
1 Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractCu-Nb multilayer nanocomposites exhibit high resistance to radiation damage because Cu-Nb interfaces are strong sinks for radiation-induced defects as well as sites for efficient Frenkel pair recombination. In this presentation, we describe atomistic modeling studies of the mechanisms by which point defects absorbed at Cu-Nb interfaces diffuse. We find that vacancies and interstitials delocalize into jog pairs on Cu-Nb interface misfit dislocations. These defect configurations diffuse by the hopping of individual jogs between misfit dislocation intersections. Defects migrate preferentially along one set of misfit dislocations. The implications of these insights for interface diffusivity and Frenkel pair recombination models are discussed.
3:45 PM - V14.5
Spatially-dependent Cluster Dynamics Modeling of Vacancy and Interstitial Cluster Evolution in Ferritic/Martensitic Fe-Cr Alloys.
Thibault Faney 1 , DongHua Xu 1 , Brian Wirth 1 , Arthur Motta 2 , Djamel Kaoumi 2
1 Nuclear Engineering, UC Berkeley, Berkeley, California, United States, 2 Mechanical and Nuclear Engineering, Penn State University, University Park, Pennsylvania, United States
Show AbstractFerritic/Martensitic Fe-Cr alloys are among the best candidate structural materials for next generation nuclear power plants, yet there is a lack of comprehensive knowledge about the behavior of these alloys under high-dose irradiation. The irradiation behavior and in particular, performance changes are intimately related to microstructural changes, such as the formation of point defect clusters and dislocation loops, as well as radiation enhanced diffusion resulting in solute segregation and precipitation. Computational materials modeling will investigate the mechanisms controlling microstructural evolution in ferritic/martensitic alloys following high dose, high temperature radiation exposure. The aim of this study is to understand and predict primary defects production and defects diffusion, clustering and interaction in a thin foil heavy ion irradiation (50 to 150 nm) of ferritic/martensitic steels. The model involves spatially dependent rate theory, or cluster dynamics, equations that describe the evolution of self-interstitial (mainly interstitial loops) and vacancy clusters (voids) in ferritic-martensitic steels under irradiation. The key parameters that are input to the model (Diffusion coefficients, migration and binding energies, clusters radius, dose rate, initial dislocation density) are determined from a combination of atomistic materials modeling and available experimental data. The Modeling predictions are compared with experimental results obtained at the IVEM facility in Argonne National Laboratory.
V15: Ceramic Materials and Wasteforms II
Session Chairs
Thursday PM, December 03, 2009
Room 207 (Hynes)
4:30 PM - **V15.1
Track Formation in Amorphous and Crystalline SiO2: A Possible Description by the Inelastic Thermal Spike Model.
Marcel Toulemonde 1 , Christian Dufour 1 , Cristina Rotaru 1 , Jean Paul Stoquert 2 , Christina Trautmann 3
1 , CIMAP-GANIL, Caen-cedex 5 France, 2 , INeSS, Strasbourg France, 3 , GSI , Darmstadt Germany
Show AbstractFor swift heavy ions in the electronic energy loss regime, a large set of data is available on damage formation in SiO2 in the crystalline [1] as well as in the amorphous phase [2,3]. Experimental results include transmission electron microscopic observation of amorphous tracks in quartz [1] whereas in vitreous silica ion tracks seem to consist of a core shell track structure of different density as deduced from small angle X-rays scat-tering [4].Comparing track phenomena with respect of the two different SiO2 phases, there are clear indications that the threshold is significantly lower in vitreous silica than in α-quartz. This applies for track formation (~0.4 and ~1.8 keV/ nm, respectively), for chemical track etching (~4 and 7 keV/nm respectively), and for surface sputtering (5 and 10 keV/nm respectively). Moreover by increasing the specific energy of the incident beam, these thresholds increase (so-called velocity effect).This presentation concentrates on the description of these phenomena with the inelastic thermal spike model [5]. Track formation and sputtering can be quantitatively described, assuming respectively quenching of a molten phase and atom vaporization from the sur-face. The crucial difference between the amorphous and crystalline SiO2 phase consists in the electron-phonon mean free path (λ), the only parameter in this model which is lower for vitreous silica than for crystalline quartz. The convolution of the λ value and the radial energy distribution on the electrons determines the atomic volume in which the energy is deposited. Based on these results, shape changes of metallic nano-cluster embedded in vitreous silica are discussed, including the expected threshold deduced from continuous track formation determined from chemical etching [6].[1] A. Meftah, F. Brisard, J.M. Costantini, E. Dooryhee, M. Hage-Ali, M. Hervieu, J.P.Stoquert, F. Studer, and M. Toulemonde, Phys. Rev. B49(1994)12457[2] A. Benyagoub, S. Klaumünzer, and M. Toulemonde, Nucl. Instr. Meth. B146 (1998) 449[3] T. van Dillen, E. Snoeks, W. Fukarek, C.M. van Kats, K.P. Velikov, A. Van Blaaderen, and A. Polman, Nucl. Instr. Meth. B 175-177 (2001) 350[4] P. Kluth, C.S. Schnohr, O.H. Pakarinen, F. Djurabekova, D.J. Sprouster, R. Giulian, M.C. Ridgway, A.P. Byrne, C. Trautmann, D.J. Cookson, K. Nordlund, and M. Toulemonde, Phys. Rev. Lett. 101 (2008) 175503[5] M. Toulemonde, Ch. Dufour, A. Meftah, and E. Paumier, Nucl. Instr. Meth. B166-167 (2000) 903[6] A. Dallanora, T.L. Marcondes, G.G. Bermudez, P.F.P. Fichtner, C. Trautmann, M. Toulemonde, and R.M. Papaléo J. Appl. Phys. 104 (2008) 024307
5:00 PM - V15.2
Subdivided Grain Formation and Recovery of Dislocation Structure in CeO2 Induced by Swift Heavy Ions.
Kazuhiro Yasuda 1 , Kazufumi Yasunaga 1 , Syo Matsumura 1 , Takeshi Sonoda 2 , Motoyasu Kinoshita 2
1 Dept. of Applied Quantum Phys. and Nucl. Eng, Kyushu University, Fukuoka Japan, 2 Nuclear Technology Research Laboratory, CRIEPI, Komae Japan
Show AbstractHigh burn-up irradiation conditions are known to induce characteristic microstructure change at the peripheral region of UO2 nuclear fuels called as“rim structure”, which is featured by μm-size large bubbles and sub-divided small grains. To gain insights into the formation mechanism of rim structure, we have examined radiation-induced microstructure in CeO2, which is selected as a simulation material with the same fluorite structure of UO2. CeO2 specimens were beforehand irradiated with 240 keV Xe ions or 2.4 MeV Cu ions to induce dislocation networks and implanted Xe ions, and they were subjected to irradiation with swift heavy ions (210 MeV Xe) to simulate radiation of fission fragments. It is found that 210 MeV Xe ions induce subdivided small grains in CeO2 preferentially at the vicinity of grain boundaries and pores, for both specimens preirradiated with either 240 keV Xe or 2.4 MeV Cu ions. A high density of dislocations was remained inside the subdivided grains, suggesting that the subgrains were formed by the recovery and rearrangements of dislocations through a high density of electronic excitation induced by swift heavy ions. The result also reveals that implanted Xe atoms are not always necessary to induce the subdivided grains. This work was financially supported by the Budget for Nuclear Research of MEXT, based on the screening and counseling by the Atomic Energy Commission.
5:15 PM - V15.3
Exciton Model of Materials Damage by Ion Irradiation: Physical Bases.
Antonio Rivera 2 , Miguel Crespillo Almenara 1 , Jose Olivares 3 1 , Fernando Agullo-Lopez 1 4
2 , Instituto de Microelectrónica de Madrid (CNM-CSIC), Tres Cantos. Madrid, Madrid, Spain, 1 , Centro de Microanálisis de Materiales (CMAM), Campus de Cantoblanco. Madrid Spain, 3 , Instituto de Óptica “Daza de Valdés” (CSIC), Madrid, Madrid, Spain, 4 Departamento de Física de Materiales, Universidad Autónoma de Madrid (UAM), Madrid, Madrid, Spain
Show AbstractA non-radiative exciton-decay model has been recently developed to account for swift-ion beam damage to LiNbO3, both in the single-track as well as moderate-fluence regime. The motivation to develop the model was to explain the mechanisms of defect generation by swift ion irradiation before amorphization takes place (sub-threshold conditions) because no many works are devoted to it despite the abundant experimental evidence of defect generation in sub-threshold conditions. The model is mostly phenomenological, although quantitative (provides the defect concentration) and relies on the synergy between localized (possibly self-trapped) excitons and the thermal spike generated by the ion impact. A key parameter of the model is the energy barrier that separates bound and unbound regions of the excited state of the localized exciton. In our present work the physical bases of the model are revised in line with recent experimental information on the dynamics and trapping of electronic carriers generated by femtosecond laser pulses (1,2).The operative processes during irradiation are discussed in reference to the detailed microscopic information gathered for alkali halides irradiated with purely ionizing irradiation3. From the comparative analysis, some general rules for the validity of excitonic models can be formulated and used to predict the irradiation behaviour of other materials (mostly oxides). In particular, the meaning of the energy barrier separating the bound and dissociative regions of the excited state in the f.c.c diagram can be clarified. In addition, one may explain why some oxides (e.g. SiO2), as well as alkali halides, respond to purely ionizing irradiation whereas other oxides, such as LiNbO3, can only be disordered by massive electronic excitation (e.g. swift-ion beams). Finally, we will present results obtained with a recently developed MonteCarlo code based on the exciton model of damage. The results show the evolution of damage with fluence for a variety of ions. Thus, valuable information is obtained on damage morphology as well as damage formation dynamics. 1 P. Herth, D. Schaniel, Th. Woike, T. Granzow, M. Inlau and E. Kratzig, Phys. Rev B 71, 125128 (2005)2 C. Merschjann, D. Berben, M. Imlau and M. Wölecke, Phys. Rev. Lett. 96, 186404 (2006)3 N. Itoh, and A.M. Stoneham, Materials Modification by Electronic Excitation, Cambridge University Press, (2001).
5:30 PM - V15.4
Irradiation Behavior of Nanostructurally-Stabilized Pure Cubic Zirconia.
Yanwen Zhang 1 , Weilin Jiang 1 , Fereydoon Namavar 2 , Jie Lian 3 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States, 2 , University of Nebraska Medical Center, Omaha, Nebraska, United States, 3 Department of Mechanical, Aerospace & Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York, United States
Show AbstractNanostructured materials, due to the ability of altering the physical, electronic, optical properties, make them potential candidates for a variety of technological applications, including advanced nuclear energy systems. As the world increases its reliance on nuclear energy, there is an ever-increasing demand for new radiation tolerate materials that can withstand the extreme radiation environments in reactors, accelerators, and even geologic repositories for nuclear waste. Understanding radiation effects in nanomaterials is an urgent challenge, since it may hold the key to unlock the design of new materials for advanced nuclear energy systems. In this study, irradiation behavior of nanocrystalline cubic zirconia is investigated. Generally, pure zirconia, in nature, is monoclinic at room temperature. This phase is stable up to 1170°C, and transforms into the tetragonal and then into the cubic phase with increasing temperature. Zirconia films are prepared by ion-beam-assisted deposition technique that produces nanostructurally-stabilized pure cubic zirconia at room temperature with average grain size of 7 nm. These films were irradiated with 2 MeV Au ions at 160 K and 400 K to doses up to 35 dpa. The average grain size determined by grazing incident X-ray diffraction increases with dose and saturates at high doses. Under 160 K irradiation, the increase in grain size saturates at about 30 nm, and this saturate value decreases with increasing temperature, indicating that thermal grain growth is not activated. While cubic phase is retained and no new phases are identified, some reduction of O in the irradiated films is demonstrated from the Rutherford backscattering spectroscopy measurements. The ratio of O to Zr decreases from close to 2.0 for the as-deposited films to ~1.65 after irradiated to ~35 dpa. Transmission electron microscopy observations and selected area electron diffraction have also confirmed the grain growth and phase stability. Nanocrystalline materials contain much fewer atoms in each grain with a significant volume fraction of grain boundaries or interfaces that are well within the size of collision cascades during the ion slowing-down process. The underlining mechanism for the radiation behavior in nanocrystalline zirconia will be discussed.
5:45 PM - V15.5
Swift Heavy Ion Damage in Delta-phase Oxides: Sc4Zr3O12 and Lu4Zr3O12.
Ming Tang 1 , Patrick Kluth 2 , Jian Zhang 1 , Blas Uberuaga 1 , Cynthia Reichhardt 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 , The Australian National University, Canberra, Australian Capital Territory, Australia
Show AbstractWhen swift heavy ions (SHI) penetrate a solid, they lose their energy through inelastic inter-actions with the target electrons. The resulting intense electronic excitation can produce a narrow trail of permanent damage along the ion path, a so-called ion track. The purpose of this study is to systematically investigate ion tracks in delta (δ) phase compounds Sc4Zr3O12 and Lu4Zr3O12, as well as irradiation damage evolution in these materials as ion tracks overlap at high ion fluences. Polycrystalline samples of δ-Sc4Zr3O12 and δ-Lu4Zr3O12 were irradiated with 185 MeV Au ions to fluences ranging from 1E11 to 1E13 Au/cm2 at room temperature. The crystal structure of the irradiated samples was investigated using X-ray diffraction (XRD). The microstructures of the irradiated samples were examined using high-resolution transmission electron microscopy (HRTEM) and small angle X-ray scattering (SAXS). XRD measurements revealed an order-to-disorder (O-D) phase transformation from ordered rhombohedral to disordered fluorite in both irradiated compounds at high ion fluence, with the Sc compound transforming at a higher ion fluence than the Lu compound. This result is consistent with our previous study of radiation effects in Sc4Zr3O12 versus Lu4Zr3O12 under ballistic damage conditions (conditions in which nuclear energy loss dominates over electronic loss). We also observed a “core-shell” cylinder structure of single ion tracks in cross-section (disordered core/ordered shell). We will discuss details of this core-shell structure as well as track radii, based on HRTEM observations and SAXS measurements, in this presentation.
V16: Poster Session: Advanced Materials for Nuclear Energy
Session Chairs
Gianguido Baldinozzi
Kath Smith
Kazuhiro Yasuda
Yanwen Zhang
Friday AM, December 04, 2009
Exhibit Hall D (Hynes)
9:00 PM - V16.1
Mechanical and Electronic Properties of CeO2, ThO2, and (Ce, Th)O2 Alloys.
Cem Sevik 1 , Tahir Cagin 1
1 Artie McFerrin Department of Chemical Engineering & Material Science and Engineering, Texas A&M University, College Station, Texas, United States
Show AbstractA systematic first principle study is conducted to calculate bulk modulus, elastic constants, phonon dispersion curves and electronic structures of CeO2, ThO2 and their ordered binary alloys CexTh8-xO16 with x = 1, 2, 4, 6, and 7 using LDA, GGA, LDA+U and GGA+U approaches. In order to get accurate results for these type of systems including f-electrons (Ce(4f) and Th(5f)) we optimized the U parameter for use in LDA+U and GGA+U approaches. The computed structural, mechanical, and electronical properties of CeO2 and ThO2 are observed to display strong correlation with experimental data. In particular the best agreement with experiment is obtained within the LDA+U in which on site Coulomb interaction parameter (Ueff) for Ce and Th are taken as 6.0 eV and 5.0 eV. To check the stability of alloy forms, phonon dispersion curves of CexTh8-xO16 with x = 2, 4, and 6 are computed. In all concentrations, mechanical stability conditions are satisfied for alloys. Furthermore, we observed no negative phonon branches in the phonon spctrum of alloys. Our calculations indicated a strong effect of concentration, x, on the electronic structure of CexTh8-xO16.
9:00 PM - V16.10
Quantum Chemical Molecular Dynamics Study of Chemical Reaction Process on Ni-base Alloy Surface in Gas-cooled Fast Reactor.
Ken Suzuki 1 , Yoichi Takeda 1 , Hideo Miura 1
1 Fracture and Reliability Research Institute, Tohoku University, Sendai Japan
Show AbstractNi-base heat-resistant alloys are used for the structural components in gas cooled reactors and exposed to high-temperature gas environment. The degradation phenomena of the heat-resistant alloy such as corrosion and creep are important issues concerning safety of plant operation and further development of the alloys applied in Very High Temperature Reactor (VHTR) and Gas-cooled Fast Reactor (GFR). In the GFR reactors using helium gas as coolant, although helium is an inert gas, impurities in the helium coolant actually react with the component materials such as the heat transfer tubes of the intermediate heat exchanger. It is well known that the chemical effects of impurities significantly shorten the lifetime of the components. In this study, in order to obtain a better knowledge of surface chemistry of Ni-base heat-resistant alloys, we applied the tight-binding quantum chemical molecular dynamics simulations for the chemical reaction process of impurities on Ni-base alloy in high-temperature helium environments. As a first approximation, Ni [111] surface was used as a periodic slab model and H2, H2O, CO and CH4 molecules were used as impurities. Tight-binding quantum chemical molecular dynamics simulations enabled us to present a clear view of the chemical reaction process of Ni surface such as the initial stage of oxide formation on atomic level. Impurity molecules decomposed on Ni surface and OH groups and oxygen atoms gradually moved closer to the surface and form Ni-O bonds. Hydrogen atoms liberated after the decomposition reaction migrated into the Ni lattice. Hydrogen atoms being in the lattice of Ni possessed the highly negative charge which indicated the surface oxidized by this negative charge hydrogen atom. We found that slow-formation of Ni-O bonds with increasing the amount of hydrogen atoms in the environments, implying that hydrogen atoms act as an inhibitor of the formation of Ni oxides. The effect of hydrogen on the formation oxides was confirmed by surface observation of Ni-base alloy using ESCA (Electron Spectroscopy for Chemical Analysis). It was found that characteristics of the oxide formed on the surface are significantly affected by the chemical composition of impurities.
9:00 PM - V16.11
Damage Formation in Rutile TiO2 by Swift Heavy Ion Irradiation: Analysis with a Non-radiative Exciton Decay Model.
Antonio Rivera 2 , Miguel Crespillo Almenara 1 , Jose Olivares 3 1 , Jens Jensen 5 , Ruy Sanz 4 , Fernando Agullo-Lopez 1 6
2 , Instituto de Microelectrónica de Madrid, (CNM-CSIC , Tres Cantos. Madrid, Madrid, Spain, 1 , Centro de Microanálisis de Materiales (CMAM), Campus de Cantoblanco. Madrid Spain, 3 , Instituto de Óptica, CSIC , Madrid, Madrid, Spain, 5 Thin Film Physics Division, Dept. Physics, Chemistry and Biology, Linköping University, Linköping Sweden, 4 , Instituto de Ciencia de Materiales de Madrid (ICMM-CSIC) , Madrid, Madrid, Spain, 6 Departamento de Física de Materiales, Universidad Autónoma de Madrid (UAM) , Madrid Spain
Show AbstractIn the last few years, there has been a great interest to investigate the damage caused by MeV ion irradiation of titanium dioxide (TiO2) rutile single crystals.Previous experiments have revealed that interesting micro- and nano-patterns can be produced on rutile following irradiation through self-assembled masks and subsequent wet etching [1]. Recently, a non-radiative exciton decay model [2, 3] has been satisfactorily applied to describe the damage and amorphization induced by swift heavy ions irradiation of congruent LiNbO3. The model considers the formation and localization of coupled electron-hole pairs (excitons) and their non-radiative decay under the high-temperature conditions (thermal spike) prevailing after every ion impact. In fact, it is assumed that a certain energy barrier ε has to be overcome to go from the bound to the dissociative regions of the excited exciton state in a configurational coordinate diagram. It is believed that the model may have general validity and could thus also be applied to other oxides. In this work we have used the model to describe the damage induced in rutile TiO2 single crystals after irradiation with Br ions at various energies (13 MeV and 25 MeV) with electronic stopping powers at the surface of 7 and 11 keV/nm respectively. In order to investigate the damage and amorphization levels, the irradiated samples have been characterized by Rutherford backscattering spectroscopy in channeling geometry (RBS/C) and optical reflectivity. At low ion fluences the generated latent tracks are not overlapping and the crystal behaves as an effective medium with an average refractive index lower than that of the virgin crystal. Above a certain fluence the latent tracks completely overlap yielding a homogeneous amorphous top layer on the crystalline sample giving fully random scattering and thus, the corresponding interference profile. The thickness of the amorphous layer monotonically increases with fluence. The experimental data can be reasonably described within the exciton model using a barrier energy ε = 0.4 eV and an amorphization threshold of 6 keV /nm, confirming the generality of the theoretical scheme. [1] M. Supinski, K. Hjort, R. Sanz and J. Jensen, Nucl. Instrum. Meth. Phys. Res. 266, 3113 (2008)[2] F. Agulló-López, A. Méndez, G. García, J. Olivares and J. M. Cabrera, Phys. Rev. B74, 174109 (2006)[3] A. Rivera, A. Mendez, G. García, J. Olivares, J.M. Cabrera, F. Agulló-López, J. Lumin.128, 703 (2008)
9:00 PM - V16.12
Fuel Rod Cladding Material Behavior Under Severe Nuclear Accident Conditions.
Mirco Grosse 1 , Juri Stuckert 1 , Martin Steinbrueck 1 , Leo Sepold 1
1 Institute for Material Research, Forschungszentrum Karlsruhe, Karlsruhe Germany
Show AbstractIn the framework of the QUENCH program the severe nuclear accidents behaviour of fuel rod bundles was simulated experimentally at the Forschungszentrum Karlsruhe (Germany). Up to now 15 large scale tests and a lot of laboratory tests were performed to investigate separate effects . At temperatures of 1000°C and above various physico-chemical processes occur in steam or/and air containing atmospheres:-oxidation and nitridization of cladding tubes-hydrogen uptake and release in steam containing atmospheres-diffusion and evaporation of alloying elements in Zr-Sn alloys-interacting between different material, for instance the formation of eutectics with low melting pointThe paper gives an overview about the former large-scale QUENCH tests, a review of their results and an outlook of future tests which shall be performed until 2014. The presentation is focused on the steam oxidation behaviour of cladding materials. Oxidation in pure oxygen or steam and nitridization in pure nitrogen follow a parabolic kinetics except for temperatures at which the so called breakaway effect occurs. This effect is caused by a tetragonal to monoclinic phase transition in the oxide which results in the formation of a large amount of cracks. Due to these cracks the protective behaviour of the oxide layer is reduced. Oxidation in air follows a nearly linear kinetics. N2 acts in a catalytic manner.The hydrogen absorption by the remaining metal was determined by means of quantitative analysis of neutron radiographs. A maximum in the H concentration during steam oxidation is reached after short times at temperatures at which a compact oxide grows. Later the hydrogen concentration decreases with the power of -¼. At temperatures at which the breakaway effect occurs, cracks in the oxide act as “hydrogen pumps”, i.e. the H concentration in the remaining metal is much higher than expected for a compact oxide layer.Most of the advanced claddings consist of homogeneous zirconium alloys. However, also duplex material is applied. The AREVA D4/Dx material consists of a Zry-4 bulk and an oxidation protection layer with reduced Sn content. The diffusion of Sn from the bulk into the layer at temperatures between 1000°C and 1400°C was investigated by means of X-ray fluorescence analysis. A faster diffusion connected with lower activation energy was found compared to values given in the literature for pure Zr.In the sub-program QUENCH-ACM the severe accident behaviour of advance cladding materials and of the classical alloy Zry-4 are compared. Tests with similar temperature scenarios were performed with the classical Zry-4, the Russian VVER cladding alloy E110, the M5™ alloy produced by AREVA and the Westinghouse alloy ZIRLO™. Differences in the oxidation kinetics will be discussed.
9:00 PM - V16.13
Chloride Ion Binding to the Surface of Calcium-Silicate-Hydrate.
Mostafa Youssef 1 , Roland Pellenq 2 , Bilge Yildiz 1
1 Department of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 Department of Civil and Environmental Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States
Show AbstractThe ingress of chloride ions from geological repository environment to a metallic surface confining radioactive nuclear wastes leads to the corrosion on these surfaces. However, the use of cement-based materials as an outer barrier can immobilize chloride ions through a binding process. This binding is suggested to result from a combination of chemistry and electrostatics. Moreover it is not amenable to a simple electrical double layer (EDL) interpretation. The ultimate goal of this work is to improve our understanding of the surface reactions on cement-based materials and the long-term prediction of the performance of a cement-based geological repository from an atomistic point of view. Calcium-Silicate-Hydrate (C-S-H) is the most abundant and main binding solid phase in cements tailored specifically for nuclear wastes confinement. We develop an atomic-scale mechanistic picture for the binding process through molecular dynamics simulations and structure optimization techniques. A recently developed molecular model for C-S-H is used as the starting point. Structure optimization using a set of parameterized equations to describe the interatomic forces is being performed to calculate the surface energy of C-S-H and adsorption energy of water molecules on it. Currently available data inferred experimentally by Nuclear Magnetic Resonance indicate that significant fraction of the chloride bound to cement solid phases occurs in ion clusters containing chloride ion and an accompanying cation. We are investigating different clusters of ion(s)/water molecules to examine energetically most favorable configurations. The results of structure optimization simulations are essential to interpret the more realistic molecular dynamics simulations. Molecular Dynamics is being performed using the same set of interatomic potentials. Recorded atomic trajectories and velocities will be analyzed to determine the binding sites, calculate the orientation order parameter for water molecules in the first hydration shell and compute spatial-dependent diffusion coefficients, mass and charge density profiles. This wealth of information will provide insight into the binding mechanism, binding capacity and the effect of the accompanying cation (monovalent versus divalent).
9:00 PM - V16.14
Cathodic Arc Thin Film Synthesis and Characterization of Mn+1AXn Phases: Interconversion from Ti2AlC to Ti3AlC2.
Robert Aughterson 1 , Marcela Bilek 2 , Johanna Rosen 2 , Gregory Lumpkin 1 , Darren Attard 1 , Daniel Riley 3 4
1 Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Lucas Heights, New South Wales, Australia, 2 School of Physics A28, The University of Sydney, Sydney, New South Wales, Australia, 3 School of Engineering, The University of Newcastle, Callaghan, New South Wales, Australia, 4 Department of Mechanical and Manufacturing Engineering, The University of Melbourne, Parkville, Victoria, Australia
Show AbstractMAX phase materials display many of the desirable properties of both ceramic and metallic materials. The Mn+1AXn phases consist of transition metals (M), group 13 or 14 element (A) and either carbon or nitrogen (X), n is 1, 2 or 3. These materials have previously shown good thermal stability, high oxidation resistance, ductility, machinability and high electrical conductivity. Such phases are of interest within nuclear materials science due to their desirable properties, and have potential application in both Gen4 and fusion technologies. The major challenge in preparing these materials is creating a pure ternary phase, there are often impurity phases based on stable binary phases, such as TiC, which are undesirable. This work explores the pathways involved in the dissolution of one phase into another as a reliable method for fabrication. In particular the aim is to study the inter-diffusion of Ti2AlC and TiC to determine conditions for Ti3AlC2 formation. Using cathodic arc plasma deposition several thin films consisting of one layer of TiC and one layer of Ti2AlC were prepared. These films were annealed with in-situ X-ray diffraction patterns collected to determine and track the phases present. Further ex-situ characterisation was performed using high resolution imaging transmission electron microscopy (TEM) to study the nanoscale properties, such as grain size and composition. Results show that uniform layers of TiC and Ti2AlC each of ~60nm were fabricated with some minor inclusions. A phase change from Ti2AlC to Ti3AlC2 was detected at 1200°C via in-situ XRD.
9:00 PM - V16.15
Thermal Stability of Microstructure in Grain Boundary Character Distribution-Optimized and Cold-Worked Austenitic Stainless Steel Developed for Nuclear Reactor Application.
Shinichiro Yamashita 1 , Yasuhide Yano 1 , Ryuusuke Tanikawa 2 , Norihito Sakaguchi 2 , Seiichi Watanabe 2 , Masanori Miyagi 3 , Shinya Sato 3 , Hiroyuki Kokawa 3
1 Oarai Research and Development Center, Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Ibaraki Japan, 2 Center for Advanced Research of Energy Conversion Materials, Hokkaido University, N-13, W-8, Kita-ku, Sapporo Japan, 3 Department of Materials Processing, Graduate School of Engineering , Tohoku University, 6-6-02, Aramaki-aza-Aoba, Aoba-ku, Sendai Japan
Show AbstractRecent studies on grain boundary structure have shown that an optimal distribution of a high frequency of coincidence site lattice (CSL) boundaries and consequent discontinuity of random boundary network in the material is one of very effective methods to improve the intergranular corrosion resistance [1]. This advantageous property, one of important ones for structural material of nuclear reactor, can be obtained through simple thermomechanical treatment process without any change of original chemical composition.In this study, grain boundary character distribution (GBCD)-optimized Type 316 corresponding austenitic stainless steel has been developed as a nuclear material for next generation energy systems. Some of steel sheets were cold-rolled additionally for making the GBCD-optimized and cold-worked (GBCD+CW) specimens. These specimens, including as GBCD-optimized, were thermally-aged at 973 K for 1 and 100 h and were examined by transmission electron microscopy (TEM) to evaluate thermal stability of the microstructure, mainly focusing on the precipitation behavior during thermal ageing and the influence of additional cold-working on thermal stability of the GBCD-optimized microstructure.A base material used is Ti-modified Type 316 austenitic stainless steel originally developed for fast reactor core application and has a typical composition of Fe-16Cr-14Ni-0.05C-2.5Mo-0.7Si-0.025P-0.004B-0.1Ti-0.1Nb in wt% [2]. The base material was cold-rolled to 3% reduction in the specimen thickness. This pre-strained specimen was annealed at 1400-1420 K for 3 h and quenched in cold water. Besides that, the cold-worked specimens were made by additional cold-rolling of the GBCD-optimized one. The GBCD was assessed by orientation imaging microscopy (OIM) and also the microstructure by TEM. The detail of experimental optimization of thermomechanical treatment conditions can be found elsewhere [3].The OIM results showed that the average grain sizes and the frequencies of CSL boundaries in the typical specimens were 40-47 μm and more than 70%. On the other hand, the TEM results revealed that the GBCD+CW specimens contained dislocation cells and networks as well as deformation twin structures whereas the as GBCD-optimized ones have few dislocations. After ageing at 973 K for 100 h, the precipitates were formed not only on random grain boundaries but also on dislocations in all of the examined specimens. These precipitates were identified as M23C6 and M6C on grain boundaries and MC on dislocations from the diffraction pattern analyses, respectively. Furthermore, it was indicated that there was no significant microsturactural change, such as recrystallization and dislocation recovery, due to precipitation on random boundaries and dislocations during thermal ageing.Reference[1] M. Shimada et al., Acta. Materialia. 50 (2002), 2331.[2] I. Shibahara et al., J.Nucl.Mater. 204 (1993) 131.[3] S. Yamashita et al., to be published.
9:00 PM - V16.16
Band Gap and Optical Properties of Icosahedral B12As2: A New Material for Nuclear Applications.
Silvia Bakalova 1 , Y. Gong 1 , C. Cobet 2 , N. Esser 2 , Y. Zhang 3 , J. Edgar 3 , Y. Zhang 4 , M. Dudley 4 , M. Kuball 1
1 HH Wills Physics Laboratory, University of Bristol, Bristol United Kingdom, 2 , ISAS - Institute for Analytical Sciences, Berlin Germany, 3 Department of Chemical Engineering, Kansas State University, Manhattan, Kansas, United States, 4 Department of Materials Science and Engineering, SUNY, Stony Brook, New York, United States
Show AbstractIcosahedral boron arsenide, B12As2, a boron-rich semiconductor, is a perspective material for nuclear applications, including neutron detectors and betavoltaic devices due to its extraordinary radiation tolerance. However, little is known about its fundamental properties. In this work, we report the experimental determination of its optical and dielectric properties derived using spectroscopic ellipsometry, including experimental details on band gaps, direct and indirect.B12As2 films deposited on 4H-SiC (0001) substrates by chemical vapour deposition were analyzed by ellipsometry using a Woollam ellipsometer for the photon energy range from 1.2 to 5.4 eV and the BESSY synchrotron source ellipsometer for the vacuum UV spectral range up to 10 eV. Parameterized optical models were built to extract the dielectric function, index of refraction and optical absorption of the B12As2 films. We determine an indirect optical bandgap of 3.9 eV which is higher than the previously reported of 3.47 eV [1] and significantly higher than 2.3-2.8 eV reported in theoretical estimates [2]. Optical transitions involving phonon absorption and emission were resolved and the threshold energy of the high energy process accompanied by phonon emission was estimated to 4.08 eV. Furthermore direct optical transition was found at 5.8 eV. Spectral positions of critical points in the high energy part of the dielectric function were also determined, corresponding to interband transitions at higher symmetry points in the Brillouin zone. Comparisons with previous ab initio calculations of the electronic structure of B12As2 are made.References:[1] G.A. Slack, T.M. McNelly, E.A. Taft, J.Phys.Chem.Solids, Vol 44 (10) pp. 1009-1013 (1983)[2] I. Morrison, D.M. Bylander, L. Kleinman, Phys Rev B, Vol 45 (4) pp. 1533-1537 (1992); D. R. Armstrong, J. Bolland, P. G. Perkins, Theoret. Chim. Acta (Berl.) 64, pp. 501-514 (1984); Dong Li, W.Y. Ching, Phys Rev B, Vol 52 (24) pp. 17073-17083 (1995)
9:00 PM - V16.17
Thermochemical Modeling of Transuranic Oxide Fuels.
Theodore Besmann 1 , Stewart Voit 1 , John Vitek 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe current U.S. emphasis on transuranic fuels as vehicles for consuming long-lived radionuclides has resulted in a requirement for operating to high burnups with complex, multi-actinide oxide fuels. In order to sufficiently understand behavior to assure safe and efficient reactor operation it is necessary to understand the phase equilibria and thermochemistry of these transuranic fuel systems, including potential fuels and targets for advanced burner reactors, particulate fuels for advanced gas-cooled reactors and “deep burn,” and targets for plutonia consumption. This paper reports on efforts to use literature data with current defect structure solution models such as the compound energy formalism to represent complex, actinide fluorite structure oxides.This research was sponsored by the U.S. Department of Energy through the Office of Nuclear Energy–Fuel Cycle Research and Development Program at Oak Ridge National Laboratory under contract DE-AC05-00OR22725 with UT Battelle, LLC.
9:00 PM - V16.18
Reprocessing Silicon Carbide Inert Matrix Fuel by Molten Salt Hot Corrosion Enhanced by Moisture.
Ting Cheng 1 , Baney Ronald 1 , James Tulenko 1
1 Material Science and Engineering, University of Florida, Gainesville, Florida, United States
Show AbstractSilicon carbide is one of the prime candidates as the matrix material in inert matrix fuels (IMF) being designed to reduce plutonium inventories and the long half-life actinides through transmutation. Since complete transmutation is not practical in a single in-core run, it is necessary to reprocess the inert matrix fuels. The current reprocessing techniques of many inert matrix materials involve dissolution of spent fuels in acidic aqueous solution. However, SiC cannot be dissolved by that process, which requires new reprocessing techniques. An efficient process has been developed for separating transuranic (actinide) species from the bulk silicon carbide (SiC) matrix. At high temperature (above 850 degree C) SiC can be corroded in molten potassium carbonate (K2CO3) in the lab atmosphere and form water soluble potassium silicate. It has been found that the SiC corrosion rate could be significantly enhanced approximately 20-50 times by employing water moisture in the corroding atmosphere. Separation of Ceria a surrogate for the plutonium fissile fuel was achieved by dissolving the SiC corrosion product in boiling water. Ceria (CeO2) was not corroded in these molten salt environments with and without water moisture and was successfully separated. Reaction-bonded SiC rods and Hot-pressed SiC sheets were tested in this research. The reaction mechanism of the hot corrosion moisture-applied process will be discussed.
9:00 PM - V16.19
Heteroepitaxial B12As2 on SiC Substrates.
Yi Zhang 1 , J. Edgar 1 , Yinyan Gong 2 , Martin Kuball 2 , Yu Zhang 3 , Michael Dudley 3 , E. Kenik 4 , H. Meyer III 4
1 Chemical Engineering, Kansas State University, Manhattan, Kansas, United States, 2 H.H.Wills Physics Laboratory, University of Bristol, Bristol United Kingdom, 3 Materials Science and Engineering, Stony Brook University, Stony Brook, New York, United States, 4 Materials Science and Technology, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractIcosahedral boron arsenide, B12As2, is a wide band gap semiconductor with extraordinary self-healing ability from radiation damage, which makes it useful for betavoltaic cells, a device which can directly convert nuclear energy into electrical power. In this study, epitaxial thin films of B12As2 were deposited on on-axis c-plane 6H-SiC and 4H-SiC substrates by chemical vapor deposition from diborane (B2H6) and arsine (AsH3). Raman spectroscopy showed that the films were B12As2 with good crystalline quality. X-ray diffraction patterns demonstrated an c-axis oriented crystalline B12As2 films deposited at temperatures ranging from 1200 to 1400 °C. The As/B ratio present in the films, revealed by x-ray photoelectron and Auger spectroscopy, was 6:1, consistent with B12As2 for different As/B reactant ratios, and throughout the layer thickness. The orientational relationship between B12As2 and the SiC substrate was (0001)<10-10>||(0001)<11-20>.Synchrotron x-ray Laue pattern showed that the B12As2 film deposited was a combination of twinned and untwinned B12As2. The average growth rate of B12As2 on SiC substrates depended on the deposition temperature, increasing from 0.6 μm/h at 1200 °C to 1.2 μm/h at 1300 °C and decreasing to 0.8 μm/h at 1400 °C. The maximum growth rate observed at 1300 °C might be due to significant decomposition of the reactants at higher temperatures. The films were p-type with the carrier concentration on the order of 1012 and 1014 cm-3 at 300K and 370K, respectively. The activation energy of the p-dopants extracted from temperature-dependent Hall-effect measurement was 565 meV. 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.
9:00 PM - V16.20
Effect of Y2O3 and Ti Additive Elements on Mechanical Alloying of NiAl Intermetallics.
Yong Deog Kim 1 , Brian Wirth 1
1 Department of Nuclear Engineering, University of California Berkeley, Berkeley, California, United States
Show AbstractThe intermetallic compound, NiAl, is a promising material for high temperature structural applications such as in aviation jet engines or gas turbines, provided that its high temperature creep strength can be improved. In this presentation, we describe the results of a study to determine whether it is possible to incorporate a high number density of very thermally stable Y-Ti-O nanoclusters, akin to those recently observed to improve creep strength and radiation resistance in nanostructured ferritic alloys. We have used mechanical alloying of NiAl with Y2O3 and Ti powders to produce oxide dispersion strengthened (ODS) NiAl alloys. The mixed powders were consolidated by Spark Plasma Sintering and hot extrusion. Here we present results to assess the effect of Y2O3 and Ti alloying elements on the microstructure and strength of intermetallics with a composition of NiAl-1wt.%Y2O3 and NiAl-1wt.%Y2O3-1wt.%Ti, as a function of milling, consolidation and thermal annealing conditions.
9:00 PM - V16.21
Role of Charge Transfer in Defect Formation for Low-Energy Recoils in SiC.
Fei Gao 1 , Haiyan Xiao 1 , William Weber 1
1 , Pacific Northwest National Lab, Richland, Washington, United States
Show AbstractUnderstanding the basics of ion-solid interaction has led to significant developments in state-of-the-art molecular dynamics (MD) simulations with empirical potentials, and these simulations have dramatically advanced the understanding of defect and defect processes in a number of materials, ranging from metals to semiconductors to insulators. However, these classical MD approaches do not provide information on the basic questions regarding the response in the electronic structure from ion-solid interactions, which represents a computational challenge beyond classical MD simulations. Large-scale ab initio MD simulation methods have been developed for the study of ion-solid interactions in 3C-SiC, and these methods are used to explore the effects of charge transfer and charge-density redistribution on the dynamics and charge-state of defect formation. The displacement threshold events along four main crystallographic directions for both C and Si recoils have been investigated. The results reveal that significant charge-transfer occurs between atoms, and the resulting defects can enhance charge transfer to surrounding atoms. Furthermore, it is shown that the variation of charge on atomic recoils can alter the energy barrier for stable defect formation, and the corresponding dynamic evolution is a charge-transfer-assisted process. This phenomenon should have significant impact on defect generation and evolution in a broad range of covalent and ionic materials. The simulations also provide important insights into the formation of charged vacancy defects. The C vacancy is a positively-charged defect that exhibits a significant Jahn-Teller distortion, whereas the Si vacancy is a negatively-charged defect. The displacement threshold energies along principle directions are determined for both C and Si recoils, and the weighted average Ed values for the four main crystallographic directions are about 25.5 eV for C and 46.2 eV for Si. These values are smaller than the weighted values for the same directions determined using classical MD simulations with a modified Tersoff potential, which may be due to altered energy barriers for stable Frenkel pair formation due to charge transfer.
9:00 PM - V16.22
Evaluation of Adhesive Strength of Corrosion-resistance Layer on Structure Materials at Elevated Temperatures.
Manabu Satou 1
1 Quantum Science and Energy Engineering, Tohoku University, Sendai Japan
Show Abstract Adhesive strength of the layers that protect from corrosive environments for structure materials in nuclear systems including fusion and fission reactors is of interest not only from fundamental point of view but also engineering point of view, because the protective function of the layer is effective if the layers maintained on the materials with suitable strength. In this paper, the adhesive strength of the layers was evaluated by means of a laser shock method, which uses a pulsed laser to generate shock wave that creates tensile stress inside the specimen and can exfoliate the coating layer. The coating layers examined were magnetite on steels and yttria on vanadium alloys. The magnetite layers were prepared by oxidation at elevated temperature in oxidizing atmosphere. Oxidation for 200 hours in air created 10 μm-thick magnetite on carbon steel. Typical strength of the layer was evaluated to be about 50MPa at ambient temperature. The yttria layers were prepared by reduced pressure plasma spray or hot-dipping method followed by heat-treatment for crystallization. Typical strength of the layer was evaluated to be about 400 MPa for the plasma-sprayed yttria coating. The adhesive strength was varied from around one-tenth of yield stress to the ultimate tensile strength of the base materials depending on the combination and method preparing the layers. It was indicated that the adhesive strength of the layer was an essential parameter for the evaluation of the protective layers.
9:00 PM - V16.23
Modeling of Diffusion and Radiation Induced Segregation in Fe-Cr-Ni Alloys Using Ab-initio Based Multi-scale Approach.
Samrat Choudhury 1 , Benjamin Swoboda 2 , Leland Barnard 3 , Anton Van der Ven 4 , Todd Allen 1 2 3 , Dane Morgan 1 3
1 Materials Science and Engineering, University of Wisconsin, Madison, Wisconsin, United States, 2 Department of Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, Wisconsin, United States, 3 Materials Science Program, University of Wisconsin, Madison, Wisconsin, United States, 4 Materials Science and Engineering, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractFe-Cr-Ni alloys are likely to form the basis of many materials used in next generation nuclear reactors. It is well known that under an irradiation environment a large concentration of point defects form in these alloys. Radiation induced segregation (RIS) is defined as the process through which the local composition of an alloy is altered due to the transport of point defects to sinks, e.g., grain boundaries. Although RIS had been observed experimentally more than three decades ago, a clear understanding of the effect of temperature, dose and composition on RIS is still missing, mostly due to a lack of understanding of the complex diffusion mechanisms of point defects in multi-component alloys. In this work, we study the diffusion mechanisms of vacancy and interstitial dumbbells underlying RIS as a function of composition. Both bcc and fcc structures are considered. In dilute Fe-based bcc and Ni-based fcc alloys, the phenomenological and diffusion coefficients are calculated for interstitial and vacancy mediated diffusion using a combined ab- intio calculations and statistical mechanics approach. It is found that in bcc Fe-Cr alloys diffusion through vacancy and interstitial mechanism will have opposite effects on RIS. Similar trends for fcc alloys are found, but larger species dependent diffusion constants suggest a greater RIS tendency. For concentrated alloys we use a combination of ab initio energetics, cluster expansion formalism and kinetic Monte Carlo approach to calculate phenomenological and diffusion coefficients as a function of composition.
9:00 PM - V16.24
Grain Boundary Characteristics Evaluation by Atomistic Investigation Methods.
Yoshiyuki Kaji 1 , Tomohito Tsuru 1 , Youji Shibutani 2
1 , Japan Atomic Energy Agency, Ibaraki Japan, 2 , Osaka University, Osaka Japan
Show AbstractThe grain boundary (GB) has been recognized for one of the major defect structures in determining the material strength. It is increasingly important to understand the individual characteristics of various types of GBs due to the recent advances in material miniaturization technique.In the present study three types of GBs of coincidence site lattice (SCL), small angle (SA), and random types are considered as the representative example of GBs. The GB energies and atomic configurations of SCL are first evaluated by first-principle DFT and the embedded atom method (EAM) calculations. SA and random GBs are subsequently constructed by the same EAM, and the fundamental characteristics are investigated by the discrete dislocation mechanics models and the Voronoi polyhedral computational geometric method. As the result, it is found that the local structures are well accorded with the previously reported high resolution-transmission electron microscope (HR-TEM) observations, and that stress distributions of CSL and SA GBs are localized around the GB core. The Random GB shows extremely heterogeneous core structures including a lot of pentagon-shaped Voronoi polyhedral resulting from the amorphous-like structure.
9:00 PM - V16.25
An Ab Initio Investigation of Interstitial Motion and Diffusion in Fe-Ni-Cr.
Leland Barnard 2 , Samrat Choudhury 1 , Dane Morgan 1 2
2 Materials Science Program, University of Wisconsin-Madison, Madison, Wisconsin, United States, 1 Materials Science and Engineering, University of Wisconsin-Madison, Madison, Wisconsin, United States
Show AbstractThe Fe-Ni-Cr alloy system forms an important class of structural materials in the nuclear power industry, both in existing reactors and for future reactor designs. Point defect diffusion under irradiation leads to microstructural changes (e.g., dislocation loops and void formation) in this system that can impact long-term performance. It is therefore essential to understand the mechanisms and consequences of point defect transport in these alloys. Here we focus on interstitial-mediated diffusion and how the mechanisms and rates of migration are influenced by the multiple species in the alloy. In this study, we utilize ab initio-based calculations to map out the energy landscape for interstitial migration in dilute Fe-Cr and Fe-Ni alloys, revealing features such as migration pathways with multiple energy minima, which would be difficult to predict through other means. We also explore the feasibility of ab initio molecular dynamics as a tool for determining species- and concentration-dependent diffusivities in Fe-Ni-Cr alloys through the interstitial mechanism.
9:00 PM - V16.26
Characterization of Changes in Properties and Microstructure of Glassy Polymeric Carbon Following Ag Ion Irradiation.
Malek Abunaemeh 1 , Mohamed Seif 2 , Abdalla Elsamadicy 3 , Ibidapo Ojo 1 , Kudus Ojbara 1 , Claudiu Muntele 1 , Daryush Ila 1
1 Center for irradiation of Materials, Alabama A&M University, Normal, Alabama, United States, 2 Mechanical Engineering Department, Alabama A&M University, Normal, Alabama, United States, 3 Physics, University of Alabama in Huntsville, Huntsville, Alabama, United States
Show AbstractThe TRISO fuel has been used in some of the Generation IV nuclear reactor designs. It consists of a fuel kernel of UOx coated in several layers of materials with different functions. Pyrolytic carbon (PyC) is one of these layers. In this study we investigate the possibility of using Glassy Polymeric Carbon (GPC) as an alternative to PyC. GPC is used for artificial heart valves, heat-exchangers, and other high-tech products developed for the space and medical industries. This lightweight material can maintain dimensional and chemical stability in adverse environment and very high temperatures (up to 3000οC). In this work, we are comparing the changes in physical and microstructure properties of GPC after exposure to irradiation fluences 5 MeV of Ag and 5 of MeV Au equivalent to a 1displacment per atom (DPA) at 1000, 1500 and 2000οC. For surface analysis we are using scanning electron microscopy, nano-indentation, X-ray photoelectron spectroscopy and Raman spectroscopy. The GPC material is manufactured and tested at the Center for Irradiation Materials (CIM) at Alabama A&M University.
9:00 PM - V16.27
Atomic Collision and Ionization Effects in Oxides,
Yanwen Zhang 1 , In-Tae Bae 2 , William Weber 1
1 PO Box 999, Pacific Northwest National Laboratory, Richland, WA 99352, Washington, United States, 2 Small Scale Systems Integration and Packaging Center, P.O. Box 6000 , State University of New York at Binghamton, Binghamton, 13902, New York, United States
Show AbstractRare-earth silicates with the apatite structure, titanate-based perovskites and pyrochlore compounds are potential matrices for immobilization of actinides and some long-lived fission products. These materials must endure high radiation doses associated with alpha-decay of the actinides and beta-decay of the fission products. Accumulation of radiation damage in the host phases may ultimately compromise the physical and chemical durability. Thus, it is important to understand and predict the behaviour of these materials in a radiation environment. Irradiation with ions and electrons provides accelerated study of radiation damage in nuclear materials, such as those proposed for immobilization of actinides and long-lived fission products. The effects of ion irradiation in SrTiO3, Sm2Ti2O7 and Sr2Nd8(SiO4)6O2, as representative materials, are studied using 1 MeV Au+ ions. The irradiation-induced disorder, due to atomic collisions processes, increases nonlinearly with irradiation dose and is well described by a disorder accumulation model that includes contributions from amorphous domains, point defects and defect clusters. Ioni-zation from 200 keV electrons induces recrystallization at the amorphous/crystalline (a/c) interface in SrTiO3 and Sr2Nd8(SiO4)6O2 that exhibits several distinct stages associated with residual defect annihilation near the interface, epitaxial regrowth at the interface, and a surface-stabilized amorphous state. Understanding ionization effects and the coupled effects of electronic and atomic dynamics on material behavior is a challenging area for scientific research.
9:00 PM - V16.28
GIXRD and TEM Analysis of the Structural and Microstructural Modifications Induced in MgZn2 Laves Phase by Low Energy Ion Irradiation.
Gianguido Baldinozzi 1 2 , David Simeone 2 1 , Dominique Gosset 2 1 , Laurence Luneville 2 1 , Marie-Genevieve Barthes-Labrousse 3 , Patricia Donnadieu 4 , Marc de Boissieu 4 , Stefan Bruehne 5 , Nathalie Moncoffre 6
1 MFE, ECP SPMS Lab, CNRS, Chatenay-Malabry France, 2 MFE, DEN/DMN/SRMA/LA2M, CEA, Gif-sur-Yvette France, 3 ICMMO, CNRS, Orsay France, 4 SIMAP, CNRS, St Martin d'Hères France, 5 Physikalisches Institut, University Frankfurt, Frankfurt Germany, 6 IPNL, CNRS IN2P3 UCBL, Lyon France
Show AbstractAlloys can endure severe external forces in a number of situations: wear, fatigue, ball milling and energetic particle irradiation. In many of these situations, alloys can be driven far from their equilibrium phases. It is common knowledge that irradiation can drastically affect the structural stability and the physical properties of the metallic compounds extensively used in the nuclear industry. These compounds are mostly simple alloys and little is known about the behavior of more complex ones. In particular, complex metallic alloys are systems with an intrinsic high level of chemical and topological disorder responsible for the variety of structural configurations. These systems seem to offer a challenging playground to study opposite effects: the incident particles increase the concentrations of point defects and may destroy the long range order in the alloy while, on the other hand, the annealing temperature restores the long range order via diffusion processes. It is then possible to explore the phase diagram and, in particular and to track new thermodynamic states with new/unusual properties. The Laves phases, which form the largest group of the known intermetallics, are chosen as model systems. These phases are much more complex than important commercial alloys and promising candidates for the design of new metallic materials with superior properties. The hexagonal Laves phase MgZn2 seems an interesting system to study these effects. The structural and microstructural modifications occurring in MgZn2 samples irradiated by low energy ions are studied by GIXRD and TEM. The formation of nanocrystalline structures is observed in the damaged region.
9:00 PM - V16.3
Evolution of Xe and He in Irradiation-Induced Point Defects in UO2 by Molecular-Dynamics Simulation.
Dilpuneet Aidhy 1 , Alex Thompson 1 , Chris Wolverton 1
1 Materials Science and engineering, Northwestern University, Evanston, Illinois, United States
Show AbstractWe illustrate the evolution of Xe and He atoms in the presence of the irradiation-induced point defects in UO2. During irradiation, large numbers of uranium and oxygen point defects created during the initial (ballistic) phase annihilate during the kinetic phase, leaving behind only few surviving ones. Using molecular-dynamics (MD) simulation with a classical potential, it was shown previously1 that the surviving oxygen interstitials form cuboctahedral clusters (COT), and oxygen and uranium vacancies form Schottky defects arranged in <111> directions. Furthermore, the concentration of oxygen defects depends upon the concentration of surviving uranium defects; when there are very few oxygen defects, new ones can even be spontaneously created.Here, we extend the previous work by incorporating Xe and He atoms to elucidate their effect on the evolution and concentration of point defects. We find that a Xe atom energetically prefers a uranium site (plus a corresponding uranium interstitial) compared to Xe in an interstitial site. Consequently, in the MD simulation, Xe atoms will displace uranium atoms from their lattice sites, thereby creating uranium interstitials. Subsequently, new oxygen point defects are created as a result of the created uranium defects, and these oxygen/uranium point defects then form COT and Schottky clusters. We also perform similar studies of the effect of He atoms in UO¬2. In addition, we also perform density functional theory (DFT) calculations using GGA+U formulation of the most relevant defect processes uncovered by the MD simulations. This work is funded by Department of Energy under contract DE-F607-071D1Y8931. Aidhy DS, Millett PC, Desai T, Wolf D, Phillpot S, Physical Rev. B (submitted).
9:00 PM - V16.30
Radiation Response of Nanocrystalline Rutile (TiO2).
Jiaming Zhang 1 , Jie Lian 2 , Feredoon Namavar 3 , Jianwei Wang 1 , Rodney Ewing 1
1 Departments of Geological Sciences and Materials Science & Engineering, University of Michigan, Ann Arbor, Michigan, United States, 2 Department of Mechanical, Aerospace & Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York, United States, 3 , University of Nebraska Medical Center, Omaha, Nebraska, United States
Show AbstractThe radiation response of nanocrystalline materials is of importance because of its potential application to design advanced nuclear materials with mitigation of radiation damage. We report, for the first time, the results of high energy ion irradiation of nanocrystalline rutile (TiO2) in dense thin-film synthesized on Si substrate by ion beam assisted deposition (IBAD). Cross-sectional transmission electron microscopy (TEM) reveals a crystal-to-amorphous transformation in the nanocrytalline rutile at room temperature under 1 MeV Kr2+ bombardment at a fluence of 1.25 × 1015 ions/cm2 , similar to the behavior of the bulk counterpart. The rutile phase remains stable upon irradiation at elevated temperature (575 K). The amorphous phase in the interfacial area between rutile and Si substrate, produced during the IBAD process, recrystallized as nanocrystals with rutile structure, which cannot be induced by only thermal annealing at 575 K. The complex interplay among irradiation-induced defect formation and crystallization, defects annealing activated by thermal energy and irradiation is discussed.
9:00 PM - V16.31
Multiscale Modeling of Fission Products Diffusion in SiC for TRISO Fuels.
Andrew Heim 1 , Sarah Khalil 2 , David Shrader 2 , Tyler Gerczak 2 , Todd Allen 1 2 3 , Dane Morgan 1 2 , Izabela Szlufarska 1 2
1 Materials Science and Engineering, University of Wisconsin - Madison, Madison, Wisconsin, United States, 2 Materials Science Program, University of Wisconsin - Madison, Madison, Wisconsin, United States, 3 Engineering Physics, University of Wisconsin - Madison, Madison, Wisconsin, United States
Show AbstractThe next generation of high-temperature gas-cooled fission reactors uses a TRIstructural ISOtropic (TRISO) particle fuel form, the silicon carbide (SiC) layer of which serves as the main barrier to the release of metallic fission products. During operation significant release of radioactive isotopes of silver (Ag) and cesium (Cs) from TRISO particles has been identified, causing safety and maintenance concerns. The specific mechanism and rates of transport for these fission products through SiC are not known. We are using a multiscale approach, combining ab initio, interatomic potential, and kinetic Monte Carlo methods, to study fission product transport in SiC bulk and grain boundaries. Realistic grain boundary structures are constructed with interatomic potentials, and then both bulk and grain boundary defect and hopping energetics are determined ab initio. The resulting energies are used in simple analytic and kinetic Monte Carlo models to predict fission product diffusion. We demonstrate that in bulk SiC, although Ag has relatively low hopping barriers, it is trapped in Si sites, leading to very slow diffusion. We also demonstrate, however, that dramatic changes in the SiC defect energetics in grain boundaries can enhance the diffusion by many orders of magnitude, potentially yielding fission product transport consistent with that seen in integral release measurements. The modeling is being done in close collaboration with diffusion couple experiments, and the connection between the results will be discussed.
9:00 PM - V16.32
Fabrication and Thermophysical Characterization of the Erbium-Doped Uranium Dioxide.
Hiromichi Gima 1 , Ken Kurosaki 1 , Masahito Katayama 2 , Hiroaki Muta 1 , Masayoshi Uno 3 , Sinsuke Yamanaka 1 , Takeishi Kuroishi 2 , Masatoshi Yamasaki 2
1 Division of Sustainable Energy and Environmental Engineering, Osaka University, Osaka Japan, 2 , Nuclear Fuel Industries, Ltd., Osaka Japan, 3 Research Institute of Nuclear Engineering, Fukui University, Fukui Japan
Show AbstractIn order to reduce the number of spent fuels of nuclear power plants, it is effective to develop super high burnup fuels with higher enrichment of 235U than current criteria of 5 wt.%. However, the Japanese nuclear facilities are restricted within the upper limit of the fuel enrichment to be 5 wt.% from the viewpoint of criticality safety. In such a situation, by adding small amounts of erbia (Er2O3) to all uranium dioxide (UO2) powders whose enrichment of more than 5 wt.%, the super high burnup will be achieved with maintaining the criticality safety, because Er would be a slow burnable poison suitable for use in a light water reactors. In addition, the critical safety of the Er-doped UO2 fuels will be equivalent to the UO2 fuels with the enrichment of 5 wt.% or less. We call this concept “Erbia Credit”. When utilizing the Er-doped UO2 fuel, it is very important to understand the thermal and mechanical properties, in order to evaluate the fuel performance. Here we show the results of fabrication and thermophysical characterization of the Er-doped UO2 pellets, where the erbium content is up to 10 at.%. We fabricated the Er-doped UO2 pellets by a conventional solid state reaction method and measured their thermal and mechanical properties. The effect of erbium addition on the microstructure of the pellets was evaluated. The empirical equations describing the thermal conductivity and Young’s modulus of the Er-doped UO2 pellets as a function of the erbium content and/or temperature were proposed.
9:00 PM - V16.33
Fabrication and Characterization of Zirconium-Lanthanide Mixed Hydrides.
Yuki Kitano 1 , Ken Kurosaki 1 , Hiroaki Muta 1 , Masayoshi Uno 2 , Kenji Konashi 3 , Shinsuke Yamanaka 1
1 Graduate School of Engineering, Osaka University, Osaka Japan, 2 Research Institute of Nuclear Engineering, Fukui University, Fukui Japan, 3 Institute for Materials Research, Tohoku University, Ibaraki Japan
Show AbstractWe have proposed a new concept of using the hydrides of zirconium-gadolinium alloys (Zr-Gd hydrides) as a burnable poison in fast reactors (FRs). If the Zr-Gd hydrides are introduced in FRs, the initial reactivity can be decreased, leading to improvement of the safety and economic efficiency of the reactors. Similarly, zirconium-erbium hydrides and zirconium-dysprosium hydrides would have possibilities for utilizing as a burnable poison in FRs. Although, the thermal and mechanical characteristics of the hydrides are very important from a practical perspective, these properties of the hydrides have been scarcely reported. Therefore, the purpose of the present study is to fabricate the hydrides of zirconium-lanthanide alloys (Zr-Ln hydrides) and to evaluate the properties such as Young’s modulus and thermal conductivity. We will show the fundamental physical properties data of the Zr hydrides, Gd hydrides, Er hydrides, Dy hydrides as well as the Zr-Ln hydrides.Present study includes the result of “Development Study of Fast Reactor Core with Hydride Neutron Absorber” entrusted to Tohoku University by the Ministry of Education, Culture, Sports, Science and Technology of Japan (MEXT).
9:00 PM - V16.34
Thermal Conductivities of Cs-M-O (M = Mo or U) Ternary Compounds.
Kazuyuki Tokushima 1 , Ken Kurosaki 1 , Kosuke Tanaka 2 , Hiromichi Gima 1 , Hiroaki Muta 1 , Masayoshi Uno 3 , Shinsuke Yamanaka 1
1 Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University, Osaka Japan, 2 , Japan Atomic Energy Agency, Ibaraki Japan, 3 Research Institute of Nuclear Engineering, Fukui University, Fukui Japan
Show AbstractIt is important to understand the behavior of the fission products (FPs) for evaluation of the fuel performance. For example, in the high-burnup oxide fuels, some FPs dissolve in the fuel matrix and others form oxide or metallic inclusions, which would affect the physical and chemical properties of the fuels. Here we investigated the thermal conductivities of Cs-M-O (M = Mo or U) ternary compounds, because such compounds have been observed in the pellet-cladding gap region of high-burnup oxide fuels. We prepared high-density pellets of Cs2MoO4 and Cs2UO4 by hot pressing or spark plasma sintering, and the thermal conductivity (κ) was evaluated from the thermal diffusivity measured by the laser flash method. The κ values of Cs2MoO4 and Cs2UO4 were quite low compared with UO2 (e.g. 0.6 Wm-1K-1 at 300 K for Cs2MoO4).
9:00 PM - V16.35
Design Optimization of Electrode Materials for Schottky and Photoconductive CdZnTe Radiation Detector.
Xiaoyan Liang 1 , Jiahua Min 1 , Changjun Wang 1 , Jun Chen 1 , Chenying Zhou 1 , Linjun Wang 1 , Jijun Zhang 1
1 School of Materials Science and Engineering, Shanghai University, Shanghai China
Show AbstractCadmium Zinc Telluride has aroused great interests due to its advantages of high resistivity, atomic order, large energy band gap, and is adjustable according to Zinc concentration. Detector made from CdZnTe performs lower leakage current, higher detecting efficiency, and is not sensitive to humidity. Device is small and states good energy resolution to X-rays, γ-rays at room temperature, and shows no polarization, with detectable energy in the range from 10 keV to 6 MeV. Effective choice of electrode material is essential for preparation of high-performance CZT devices. Rectifier effect and Ohmic effect, i.e. forming Schottky contacts and Ohmic contact, are the two important effects which may occur in the contact of metal and CZT. Therefore, the correct contact preparation is the key technology to fabricate Schottky and photoconductive devices. In this paper, electrode materials design of Schottky and photoconductive CdZnTe devices is studied by I-V electric analyses, barrier heights calculation and spectrum response test. The results showed that the Ohmic linear coefficient of Al-n/CZT-AuCl3 devices was closest to 1, with minimum leakage current, when comparing with other three different electrode materials designs ( In-n/CZT-AlCl3, Au-p/CZT-Au, Au / Cr-p / CZT-Au / Cr). Thereby, electrode materials design of Al-n/CZT-AuCl3 with better Ohmic coefficient, lower leakage current is propitious to achieve the effective photoconductivity devices. Besides compared with five electrode designs for Schottky devices, the barrier height of In-p/CZT-Au device was 0.948eV, with the reverse leakage current only ~ 8nA at 360V. According to electrical performances analyses, rectifier effect appeared to be distinct for electrode materials designs on P-CZT Schottky devices. The energy resolutions of Al-n/CZT-AlCl3 device and In-p/CZT-Au device were 25.2% and 30%, and the ratios of signal to noise were 1.75 and 2.08, respectively. It indicated that In-p/CZT-Au minished noise affection effectively, though the resistance of In-p/CZT-Au was low reversely. The reason could be that the space charge region caused by Schottky barrier produces a strong electric field, which could accelerate electronic-hole flow to the respective electrodes, resulting in the decline of composite probability. Thereby, compared with the photoconductive detector, Schottky detectors can be achieved to enhance the capacity of charge collection without increasing leakage current and bringing in more noise.
9:00 PM - V16.36
Radiation Response of Nano-sized Monoclinic ZrO2.
Fengyuan Lu 1 , Jiaming Zhang 2 , Alexandra Navrotsky 3 , Fereydoon Namavar 4 , Rodney Ewing 2 , Jie Lian 1
1 Mechanical, Aerospace & Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, New York, United States, 2 Departments of Geological Sciences and Materials Science & Engineering, University of Michigan, Ann Arbor, Michigan, United States, 3 NEAT ORU and Peter A. Rock Thermochemistry Laboratory, University of California at Davis, Davis, California, United States, 4 University of Nebraska Medical Center, University of Nebraska, Omaha, Nebraska, United States
Show AbstractNanocrystalline zirconia recently attracts extensive research interests due to their unique mechanical, thermal and electrical properties as compared to bulk zirconia counterparts, and it is of particular importance to control the phase stability of different polymorphs (amorphous, cubic, tetragonal and monoclinic phases) at different size regimes. The radiation response of monoclinic nano-sized ZrO2 has been studied by energetic beam bombardment under in-situ transmission electron microscopy (TEM) observation, and the effects of temperature and size on the phase transformation processes have been investigated. A phase transformation from monoclinic-tetragonal polymorphs occurs in monoclinic ZrO2 with sizes of 17 nm and 50 nm upon 1 MeV Kr2+ irradiation at room temperature, similar to that observed in bulk ZrO2. The critical dose inducing monoclinic-to-tetragonal phase transformation increases at higher irradiation temperature, suggesting that defect annealing (e.g., oxygen vacancies) at elevated temperature inhibits the monoclinic to tetragonal phase transformation process. Larger-sized monoclinic zirconia (50 nm) is more resistant to the radiation-induced tetragonal phase formation, consistent with the fact that the monoclinic phase is thermodynamically more stable at larger size regime. The correlation among the tendency of phase transformation, crystal size and structure, defect production and dynamic annealing, and the thermodynamic properties of nanostructured zirconia will be discussed.
9:00 PM - V16.38
Evaluation of Nano-mechanical Properties of Irradiated Materials by Kinetic Indentation Method.
Jong J. Lee 1 , Yong Choi 1 , Kee N. Choo 2 , Do S. Kim 2 , Bong G. Kim 2 , Young W. Lee 2 , Young H. Kang 2
1 Advanced Materials Engineering, Sunmoon University, Asan Korea (the Republic of), 2 HANARO, KAERI, Daejeon Korea (the Republic of)
Show AbstractSmall specimen with simple geometry has advantage for irradiation tests because the tests have limits for experiments and should reduce nuclear wastes after post irradiation tests. In this study, kinetic indentation method is applied to evaluate various nano-mechanical properties of irradiated materials for the establishment of a reliable test method using disc shape of irradiated specimen with 3 mm in diameter. The method is theoretically based on both the proportion of elastic and plastic deformation and the values obtained by using a nano-indenter. The parameters for the evaluation are obtained from the diagram indentation loading, indentation depth and time. The relationship of parameters in the diagram was obtained to evaluate hardness, wear and fatigue behaviors. Emphasis is on the evaluation of nano-mechanical properties of irradiated 1007 aluminum for thermal insulator of an experimental reactor, TRISO layers of coated nuclear fuel particle for high temperature gas cooled reactor and copper-silver alloys for high performance conducting wires for fusion reactor.
9:00 PM - V16.39
Raman Scattering Spectra and IR Microscope Studies on Te Precipitates in CdZnTe Crystals.
Changjun Wang 1 , Jiahua Min 1 , Xiaoyan Liang 1 , Jun Chen 1 , Jijun Zhang 1 , Chenying Zhou 1 , Linjun Wang 1
1 , Shanghai University, Shanghai China
Show AbstractSamples with different Te-rich concentration were grown by Low Pressure Vertical Bridg-man Method. IR microscope, Raman Scattering Spectra and IR transmittance Spectra were used to study Te precipitates in Te-rich CdZnTe crystals. The IR microscope picture showed that in Te-rich CZT samples, Te precipitates, which existed in the form of square, triangle and hexagonal. And the Te precipitates varied in size, the small ones were within 10 to 30 nm, and the sizes of Te inclusions were mush larger than 1μm. They could influence the point defect structure and the free carrier concentrations. In Raman spectra, we observed five peaks at 124 cm-1, 140 cm-1, 162 cm-1, 173 cm-1 respectively. The dominant peaks at 140and 160cm-1 were identified as Transverse optic(TO) mode and longitudinal optic(LO) phonons respectively in CZT. While the emission peak at 124cm-1 was attributed to the phonon with A 1 symmetry of the Te precipitates in CdTe. The origin of the peak at 150 cm-1 is not fully understood but could be the phonon with symmetry E seen in the Te single crystals at 147 cm-1. Similar two-mode (LO, TO) behaviors in CZT, which arise from CdTe and ZnTe-like vibrations were observed by Harada and Narita. The vacancies and interstitials caused by Te precipitates lead to lattice damages, which should be responsible for the change of the peak shift and relative intensity of A 1 and TO 1 in Raman spectra. The IR transmittance in the range of 1000-4000cm-1 was influenced by Te precipitates.
9:00 PM - V16.4
Vacancy Assisted Migration of Xenon at UO2 Grain Boundaries.
Emily Moore 1 2 , L. Rene Corrales 1 , Ram Devanathan 2
1 Department of Material Science and Engineering, The University of Arizona, Tucson, Arizona, United States, 2 Chemicals and Materials Sciences Division, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractGlobal concerns about potential climactic changes from energy use have resulted in increased interest in nuclear energy. In fission reactors, uranium dioxide is the most commonly used fuel. The combined effects of elevated temperature, radiation and fission gas production during reactor operation can result in degradation of thermal and mechanical properties of nuclear fuel. Point defects are created during operation and play a considerable role in the structural evolution and performance of the fuel. Xenon is one of the highest fractional released fission gases in uranium dioxide and previous work has shown that the mobility of xenon is dependent on the presence of defects. More specifically, it has been shown that xenon prefers to occupy a neutral Schottky defect, and will not migrate in crystalline uranium dioxide even at elevated temperatures. This work has used molecular dynamics simulations to investigate the effects of defects on fission gas mobility. This work focuses on the mobility of xenon at and near grain boundaries with respect to Schottky defect concentration. Polycrystalline UO2 with a large concentration of trivacancies (one uranium vacancy and two adjacent oxygen vacancies) was simulated at elevated temperatures to probe the mobility of the fission gas around the trivancancies. Vacancy assisted diffusion and displacement energies of Xe at grain boundaries at elevated temperatures will be discussed.
9:00 PM - V16.40
Effects of Gradient Heat Treatment on the Resistivity of CdZnTe Crystal.
Jun Chen 1 , Jiahua Min 1 , Changjun Wang 1 , Xiaoyan Liang 1 , Chenying Zhou 1 , Jijun Zhang 1 , Linjun Wang 1
1 Depatrment of Electronic Information Materials, Shanghai University, Shanghai China
Show AbstractCdZnTe(CZT) has gained a lot of attention because it has prospect potential for nuclear radiation detection applications. It has some unique physical properties, such as high-average atomic number, large enough band-gap, high resistivity, and good electron transport properties lead CZT to work at room temperature. However, it is well know that various impurities introduces different defect levels, which affect the electrical properties of CZT, especially crystal resistivity. Some of the impurity could get on the Te inclusions. In this record, we use gradient heat treatment to transfer Te inclusions and improve crystal resistivity. CZT crystal was grown by the Vertical Bridgman Method(VBM) under Te-rich condition. Effects of gradient heat treatment in vacuum on CZT properties such as infrared(IR)transmission, the density of Te inclusions, the distribution of various impurities and crystal resistivity were discussed by infrared(IR), infrared microscope, Hall test and time-of-flight secondary ion mass spectrometry (Tof-SIMS). The results indicated that CZT crystal grown under Te-rich condition contains large-size Te inclusions, with IR transmission less than 55%. The micron-size Te inclusions were mostly removed and IR transmission was increased to over 60% after the gradient heat treatment. According to the images of infrared microscope, Te inclusions were concentrated at the end of the CZT crystal. In addition, it could be observed that Te inclusions effectively getter the impurities of Na, Ag, In and Bi by Tof-SIMS. And it gave the direct evidence of Te inclusion moving towards the end in CZT. The impurity gettering in Te inclusions transferred to the end during gradient heat treatment. Finally, the resistivity of CZT crystal is approach to 2.5x10 10 Ωcm. Therefore, gradient heat treatment provides a method to decrease Te inclusions which can improve the resistivity of CZT crystal.
9:00 PM - V16.41
Electron and Hole Traps in Lithium Tetraborate (Li2B4O7) Crystals Being Developed for Neutron Dosimetry.
John McClory 1 , James Petrosky 1 , Mathew Swinney 1 , Shan Yang 2 , Adam Brant 2 , Larry Halliburton 2
1 Engineering Physics Department, Air Force Institute of Technology, Wright-Patterson AFB, Ohio, United States, 2 Physics Department, West Virginia University, Morgantown, West Virginia, United States
Show AbstractLithium tetraborate (Li2B4O7 or LTB) crystals are being developed for use as radiation detectors. When doped with Cu, these crystals serve as highly sensitive tissue-equivalent thermoluminescence dosimeters. They are of even more interest for neutron detection because of the large lithium (6Li) and boron (10B) cross sections for neutron capture. The relevant nuclear reactions are 6Li(n,α)3H and 10B(n,α)7Li. Isotopically enriched LTB crystals are expected to improve the neutron detection response. In our present study, electron paramagnetic resonance (EPR) and electron-nuclear double resonance (ENDOR) experimental techniques have been used to identify the basic electron and hole traps in LTB crystals. A crystal was irradiated at 77 K with x-rays, and EPR and ENDOR data were then taken near 20 K without warming the sample. The resulting EPR spectra show two dominant centers, one electronlike and one holelike. The electron center has S = 1/2 and exhibits hyperfine splittings due to the interaction of the unpaired spin with one boron nucleus (well resolved hyperfine lines from 10B and 11B nuclei are observed). This defect is formed during the x-ray irradiation when a pre-existing oxygen vacancy traps an electron (this electron is primarily localized on one boron ion neighboring the oxygen vacancy). The hole center also has S = 1/2 and exhibits a seven-line hyperfine pattern due to nearly equal interactions of the unpaired spin with two I = 3/2 nuclei. The ENDOR spectra from this hole center show that the two nuclei responsible for the seven-line pattern are both boron. Our holelike center is formed during the low-temperature x-ray irradiation when a hole is trapped at an oxygen ion in the regular lattice (the unpaired spin has nearly equal overlap of its wave function onto the two neighboring boron ions). Stabilization of the hole on the oxygen ion occurs either by self-trapping induced by a local lattice relaxation or as a result of the electrostatic attraction of an adjacent lithium vacancy. Thermoluminescence was observed when the LTB crystals were warmed to room temperature after being initially irradiated at 77 K with x rays.
9:00 PM - V16.42
Optical Waveguides Obtained by Swift-ion Irradiation on Silica.
Javier Manzano 1 2 , Miguel Crespillo Almenara 1 , Adriano Zabot 1 , Jose Olivares 4 1 , Fernando Agullo-Lopez 1 3 , Alejandro Morono 2 , Eric Hodgson 2
1 , Centro de Microanálisis de Materiales (CMAM), Campus de Cantoblanco. Madrid Spain, 2 , Euratom/CIEMAT Fusion Association, Madrid Spain, 4 , Instituto de Óptica “Daza de Valdés” (CSIC), Madrid Spain, 3 Departamento de Física de Materiales, Universidad Autónoma de Madrid (UAM), Madrid Spain
Show AbstractOptical waveguides were produced in silica by light ion implantation as early as 1968. The waveguiding was caused by the increase in refractive index associated to the lattice compaction brought about by nuclear collision damage. They were non-tunneling waveguides with low light attenuation1. More recently, the processes induced by higher energy irradiation with heavier ions (swift ions), where electronic stopping power is dominant, has been investigated. It appears that electronic excitation processes also cause a similar compaction effect at quite low fluences (10e12-10e13 cm-2) and even viscous flow [2]. In order to explore the possibilities of swift-ion irradiation to fabricate optical waveguides, silica plates have been irradiated with F 5 MeV, Cl 20 MeV and Br 20 MeV and fluences in the range of 10e12-10e14 cm-2. Useful waveguides have been achieved whose refractive index profiles have been determined from the dark mode method and show a clear correlation with the electronic stopping power curves. The results have been complemented with information on the point defects generated by electronic excitation and measured by complementary techniques (optical absorption and EPR). They have been discussed in terms of a recent exciton-decay model recently developed in our laboratory. Apart from the technological interest, the waveguiding effect offers an useful tool to analyze the damage processes triggered by electronic excitation. 1. P.D. Townsend, P.J. Chandler and L. Zhang, Optical Effects of Ion Implantation, Academic Press, 1994.2. A. Benyagoub, S. Klaumünzer, M. Toulemonde, Radiation induced compactation and plastic flowof vitreous silica. Nucl. Instrum. Method B, 146 (1998) 449-454.
9:00 PM - V16.43
Study of Crystallization Kinetics of Amorphous Layers and Tracks Generated in Lithium Niobate by Swift-ion Beam Irradiation.
Miguel Crespillo Almenara 1 , Antonio Rivera 2 , Jose Olivares 3 1 , Fernando Agullo-Lopez 1 4
1 , Centro de Microanálisis de Materiales (CMAM), Campus de Cantoblanco. Madrid Spain, 2 , Instituto de Microelectrónica de Madrid (CNM-CSIC), Madrid Spain, 3 , Instituto de Óptica “Daza de Valdés” (CSIC), Madrid Spain, 4 Departamento de Física de Materiales, Universidad Autónoma de Madrid (UAM), Madrid Spain
Show AbstractIon irradiation with high energy heavy ions is a well known method to produce latent amorphous tracks in LiNbO3 and other crystals, by means of the electronic damage. Most previous work was performed with swift heavy ions in the range of 1-10 MeV/amu [1, 2]. Recently the interest of this electronic damage has been expanded to use the low energy region of 0.1-1 MeV/amu (even benefiting somehow of the velocity effect) showing the possibility of producing from (micrometer length) isolated latent tracks to homogeneous amorphous layers by track overlapping. This has been used to develope a new route for the fabrication of optical waveguides [3, 4] with fluences several orders of magnitude lower than the standard method based on nuclear damage with ligh ion implantation. Here we present a study of the recrystallization properties of the amorphous tracks submitted to annealing treatments. This study is interesting both from the fundamental point of view and with the aim of optimizing and tayloring the optical waveguides. LiNbO3 X- and Z-cut was irradiated at RT with Br 45 MeV ions to produce latent tracks and with F 20 Mev ions to generate buried amorphous layers. Annealing was performed at several temperatures in the range 250-350 celsius degrees and for various times. The damage evolution was characterised by RBS-c measurements and optical measurements (reflectance, transmitance and guide modes). The activation energy of the recrystallization process has been obtained and is discussed in terms of local reordering and epitaxial regrowth. [1] M. Toulemonde, W. Assman, C. Dufour, A. Meftah, F. Studer, and C. Trautmann, in: Ion Beam Science: Solved and Unsolved Problems, edited by P. Sigmund, (The Royal Danish Academy of Sciences and Letters, Copenhagen, 2006), p. 263.[2] B. Canut, S. M. M. Ramos, R. Brenier, P. Thevenard, J. L. Loubet, and M. Toulemonde, Nucl. Instrum. Meth. Phys. Res. B 107, 194 (1996).[3] J. Olivares, G. García, A. García-Navarro, F. Agulló-López, O. Caballero and A. García-Cabañes, Appl. Phys. Lett. 86, 183501 (2005)[4] J. Olivares, A. García-Navarro, G. García, A. Méndez and F. Agulló-López, Appl. Phys. Lett. 89, 071923 (2006)
9:00 PM - V16.44
Characterization of UO2 Fracture During an Annealing Test.
Mathieu Marcet 1 , David Simeone 1
1 , CEA, Saint Paul Lez Durance France
Show AbstractAfter irradiation in normal operating conditions, used UO2 fuel exhibits different microstructure as a function of the radial position in the fuel pellet. These microstructures reveal the thermal gradient existing in the fuel pellet during operation, which is due to homogeneous production of heat by fission in the pellet. More specifically the High Burn-up Structure appears at the outer rim of the pellet, its coldest part, when the local burn-up reaches a 60-70 GWd/tM threshold. The corresponding fission gases behaviour during a temperature transient is a key issue regarding the PWR’s nuclear fuel licensing. This behaviour is generally studied in hot cells by measuring the fission gas release out of a slice of used nuclear fuel as a function of a predetermined thermal history.Recently the fission gas release of the HBS was studied in a Knudsen cell [1]. These releases are continuous processes except the main steps at 1000 and 1500 K, which are characterised by explosive gas release evident from the pressure transients observed on the vacuum gauges. These pressure spikes are explained by burst releases of the gas contained in closed pores. Other annealing test performed in our team on a whole fuel pellet [2] confirmed that the HBS released gases at approximately the same temperature. More over some mechanical degradation of the HBS was observed after the thermal sequence.In this study we used Environmental Scanning Electron Microscope in order to characterise the change of the UO2 microstructure during temperature ramps. Two types of sample were examined. First, some UO2 pellets sintered under Ar/H2-5% high pressure was broken to get some fragments that fit into the ESEM set-up. Second, tiny parts of HBS were collected in hot cell, put on metallic holder and sent to the ESEM. The changes in the surface of the sample were observed during the same temperature transient for both type of sample. The gas release mechanism will discussed as a function of the cracks observed both at the surface and in the bulk.[1] Hiernaut, J.P. / Wiss, T. / Colle, J.Y. / Thiele, H. / Walker, C.T. / Goll, W. / Konings, R.J.M. , Journal of Nuclear Materials, 377 (2), p.313-324, Jul 2008[2] Marcet M. et al, Proceedings of Top Fuel 2009, Paris, France, September 6-10, 2009, Paper 2055
9:00 PM - V16.45
Behaviour of Nanocrystalline Silicon Carbide Under Low Energy Heavy Ion Irradiation.
Dominique Gosset 1 2 , David Simeone 1 2 , Gianguido Baldinozzi 2 1 , Lionel Thome 3 , Yann Leconte 4
1 DMN/SRMA/LA2M MFE, CEA, Gif/Yvette France, 2 SPMS-MFE, CNRS, Chatenay-Malabry France, 3 CSNSM, CNRS, Orsay France, 4 IRAMIS, CEA, Gif/Yvette France
Show AbstractSilicon carbide is one of the most studied materials for core components of the next generation of nuclear plants (Gen IV). In order to overcome its brittle properties, materials with nanometric grain size are considered. In spite of the growing interest for nano-structured materials, only few experiments deal with their behaviour under irradiation. To assess and predict their evolution under working conditions, it is important to characterize their microstructure and structure. To this purpose, we have studied microcrystalline and nanocrystalline samples before and after irradiation at room temperature with 4 MeV Au ions . In fact, it is well established that such irradiation conditions lead to amorphisation of the material, which can be restored after annealing at high temperature. We have performed isochronal annealings of both materials to point out the characteristics of the healing process and eventual differences related to the initial microstructure of the samples. To this purpose Grazing Incidence X-Ray Diffraction has been performed to determine the microstructure and structure parameters. We observe the amorphisation of both samples at the same dose but different annealing kinetics are observed. The amorphous nanocrystalline sample recovers its initial crystalline state at higher temperature than the microcrystalline one. This effect is clearly related to the initial microstructures of the materials. Therefore, the grain size appears as a key parameter for the structural stability and mechanical properties of this ceramic material under irradiation.
9:00 PM - V16.5
Mesoscale Kinetic Monte Carlo Model for Transport of Silver through TRISO Fuel Particle.
Gabriel Meric 1 , Brian Wirth 1
1 Nuclear Engineering, UC Berkeley, Berkeley, California, United States
Show AbstractThis presentation describes the development of a mesoscale kinetic Monte Carlo model to investigate the diffusion of silver through the pyrolytic carbon and silicon carbide containment layers of a TRISO fuel particle. The release of radioactive silver from TRISO particles has been studied for nearly three decades, yet the mechanisms governing silver transport are not currently understood in detail. The model atomically resolves Ag, but provides a mesoscale medium of carbon and silicon carbide, which can, however, include a variety of defects including cracks, grain boundaries and point defect clusters. The key input parameters to the model (Diffusion coefficients, trap binding energies, interface characteristics) are determined from a combination of atomistic materials modeling and available experimental data. The predicted results, in terms of the time/temperature dependence of silver release during post-irradiation annealing and the variability of silver release from particle to particle have been compared to available experimental data of Minato and co-workers [1]. [1] K. Minato et al., “Fission product release behavior of individual coated fuel particles for high-temperature gas-cooled reactors”, Nucl.Technol., 131, 36 (July 2000)
9:00 PM - V16.6
Diffusion of Silver and Gold in Glassy Polymeric Carbon Used as Nuclear Fuel Coating for the Next Generation of Nuclear Reactors.
Ibidapo Ojo 1 , Claudiu Muntele 1 , Malek Abunaemeh 1 , Daryush Ila 1
1 , Alabama A&M University, Normal, Alabama, United States
Show AbstractThe TRISO fuel that is intended to be used for the generation IV nuclear reactor design consist of a fuel kernel of Uranium Oxide (UOx) coated in several layers and materials with different functions. Glassy Polymeric Carbon (GPC) is considered as a potential substitute for the pyrolitic carbon coatings. An important issue of these coatings is the retention of fission products, especially silver. GPC is known to have a virtually zero permeability, even for helium. However, radiation damage to the material changes the permeability of the non irradiated material. Here we are presenting experimental estimates of diffusion coefficients for silver and gold ions implanted into glassy polymeric carbon. We implanted 5 MeV silver and gold ions and used Rutherford Backscattering Spectrometry for measuring the re-distribution of the two elements into GPC as a function of time at temperatures of 600 and 1000 deg. C.
9:00 PM - V16.7
Oxide Nano-clusters and Precipitates in Mechanically Alloyed Nickel.
James Bentley 1 , D. Hoelzer 1 , C. Fu 1 , X. Chen 1 , D. Coffey 1
1 Mater. Sci. & Technol. Division, Oak Ridge National Lab, Oak Ridge, Tennessee, United States
Show AbstractFollowing development of a new class of mechanically alloyed (MA) nano-structured ferritic alloys (NFA) with exceptional mechanical properties that are largely due to the presence of high concentrations (>1023 m-3) of small (<5 nm diameter) Ti-Y-O nano-clusters (NC), an initial effort to produce NC in a face-centered cubic Ni matrix has been undertaken. As with NFA, there is major interest in potential applications to fission and proposed fusion reactors because NC-containing materials may be highly resistant to neutron radiation damage (swelling and embrittlement). In the present work, Ni-1.00wt.%Ti-0.60%Y2O3 (NTYO) and Ni-1.89wt.%Zr-0.86%La2O3 (NZLO) alloys were fabricated by attritor ball milling metal and oxide powders in Ar. Specimens for transmission electron microscopy (TEM) were prepared from milled powders by focused ion beam (FIB) lift-out methods. Compositional information at a resolution <2nm was obtained by energy-filtered TEM (EFTEM), including Ni-M jump-ratio images to reveal small precipitates and NC, and energy-dispersive X-ray spectroscopy (EDS) spectrum imaging. The grain size of the as-milled powders was only ~20 nm and, surprisingly, EFTEM revealed that high concentrations of small particles (<4 nm) were present. This is quite different behavior from that of NFA where NC are present only following high-temperature processing. Vacuum annealing ball-milled powders for 1 h at 750 and 800°C (potential extrusion temperatures) resulted in grain growth and considerable coarsening of the particles. Anneals at higher temperatures were performed to explore the stability of the particles since NC in NFA survive annealing at 1400°C. Ball-milled powders were consolidated by hot pressing at 950°C for 600 s to facilitate subsequent anneals at 1000, 1100 and 1200°C for 1 and 10 h. Grain growth and oxide-particle coarsening increased with increasing temperature. Particles as large as 100 nm were present after annealing at 1200°C. It is likely that most particles in the annealed specimens are conventional oxide phases but this is yet to be determined. Interestingly, the larger particles tend to be Zr- and Ti-rich while the smaller particles tend to be La- and Y-rich in NZLO and NTYO, respectively. First-principles calculations using the local-density-functional approximation reveal that Ti-Y-O-vacancy or Zr-La-O-vacancy cluster nuclei in Ni are never more energetically stable than TiO2 and ZrO2 phases. Again this is contrary behavior to that of cluster nuclei in NFA. In summary, the oxide particles in NTYO and NZLO alloys do not appear to share the unusual and attractive behavior of NC in NFA, an unfortunate but, nevertheless, important result that is supported by first-principles calculations. Research supported by the Laboratory Directed Research and Development Program of ORNL and at the ORNL SHaRE User Facility by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.
9:00 PM - V16.8
Fabrication and Characterization of Monolithic Nuclear Fuels by Hot Isostatic Pressing.
Jan-Fong Jue 1 , Blair Park 2 , Dennis Keiser 1 , Glenn Moore 2
1 Fuel Performance & Design, Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Fuel Fabrication, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractThe RERTR (Reduced Enrichment for Research and Test Reactors) program is developing a monolithic fuel type in order to convert US high performance reactors from highly enriched uranium fuels to low enriched metallic fuels. Hot isostatic press (HIP) bonding technique has been investigated extensively at the Idaho National Laboratory to bond uranium-molybdenum fuel foil to aluminum alloy cladding. More than twenty fuel plates have been fabricated by hot isostatic pressing and irradiated in the ATR (Advanced Test Reactor). Early experimental results indicate that the high processing temperature used during hot isostatic press bonding can cause extensive interaction between the fuel foil and cladding. This interaction layer usually contains more than one phase and is brittle in nature. The parameters that control the fuel/cladding interaction were found to be temperature, bonding time and composition of uranium-molybdenum fuel foil. To address the extensive fuel/cladding interaction concern, a diffusion barrier was introduced between the fuel and cladding. Experimental results show that such interfacial modification effectively minimized/eliminated the fuel/cladding interaction. Scanning electron microscopy with EDS/WDS capability was used to characterize the interfaces in bonded fuel plates and the results will be presented in this paper. Parameters controlling the bonding process and mechanisms of diffusion barrier approach will also be discussed.
9:00 PM - V16.9
Direct Quantitative Measurement of Trace Level Impurities in High Purity Graphite Materials by High Resolution Fast-Flow Glow-Discharge Mass Spectrometry.
Karol Putyera 1 , Gaurav Bhagat 1 , Changhsiu Liu 1 , Richard Hockett 2
1 , EAG NY, Syracuse, New York, United States, 2 , EAG CA, Sunnyvale, California, United States
Show AbstractThe calibration factors are examined for direct determination of element impurities in variety of graphite grades in the new generation of high resolution fast-flow glow-discharge mass spectrometry (FF-GDMS). It is shown that using the generalized calibration factors from the instrument’s Standard RSF table, the relative errors observed in the determination of the majority of important analytes is not acceptably small. For instance, sensitive and accurate determination of neutron adsorbing elements or exact assessments of equivalent boron content (EBC) in nuclear grade materials is very important, since the presence of impurities may significantly affect the graphite properties even in trace levels. Since there is no solid graphite certified reference material on the market today with known reference values, control of the accuracy of graphite measurements is very difficult for any technique. Thus, several metallic carbides and solid graphite samples with known impurity contents were used for calibrations and for establishing good analytical procedures for measurements of wide variety of samples. The obtained results for the large majority of important analytes using the new FF-GDMS procedure agreed well with other characterization techniques commonly used in this industry. In conclusion the FF-GDMS method gives very satisfactory results for almost all elements in the periodic table.