Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Amit Misra Los Alamos National Laboratory
Christina Trautmann GSI Helmholtzzentrum
Brian Wirth University of California-Berkeley
BB1: Structural Materials
Session Chairs
Tuesday PM, April 06, 2010
Room 3012 (Moscone West)
9:30 AM - **BB1.1
Measuring, Modeling and Managing Helium-Displacement Damage Effects in Practical High Performance Structural Alloys.
G. Robert Odette 1 , Takuya Yamamoto 1
1 Mechanical Engineering, UC Santa Barbara, Santa Barbara, California, United States
Show AbstractDisplacement damage and high levels of helium act individually and synergistically to fuel the various scourges of radiation damage, including degradation of fracture toughness at low temperatures, void swelling at intermediate temperatures and creep embrittlement at high temperatures. There are three key ingredients in addressing these issues. The first is a suite of experiments that provide a basis to understand the transport, fate and consequences of displacement damage and helium. The second is that these experiments must be full integrated with practical multiscale theory and modeling studies that are not only physically based and predictive, including for observations before they are made. Third the insight provided by the combined experimental-modeling paradigm must be used to identify microstructures and their concomitant attributes that hold promise of mitigating and managing helium and displacement damage, including thermal and irradiation stability of far from equilibrium nanostructures. However, radiation tolerance is not enough. The imperative of an outstanding balance of processing, fabrication route and performance sustaining properties is emphasized.
10:00 AM - BB1.2
Microstructural Evolution of Nanostructured Ferritic Alloys and their Response to High Dose Irradiation.
Michael Miller 1 , Chong Fu 1 , David Hoelzer 1 , Kaye Russell 1
1 MSTD, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe materials in the next-generation of nuclear energy systems will be operated under extreme mechanical, temperature and irradiation environments. One class of material that has shown remarkable promise for use under these extreme environments is the nanostructured ferritic steels - formerly referred to as oxide dispersion strengthened (ODS) steels. Nanostructured ferritic alloys, such as MA957, 12YWT, 14YWT alloys are produced by mechanically alloying pre-alloyed metal flakes of Fe, Cr, Ti and W (or Mo and Al), with yttria powder. After suitable processing, high number densities of 2-4 nm diameter titanium-, oxygen- and yttrium-enriched nanoclusters are distributed throughout the ferrite grains and decorating the grain boundaries. Attempts to fabricate these materials with similar microstructures and properties by conventional casting or powder metallurgy routes have so far been unsuccessful. Atom probe tomography has been used to study the microstructural evolution from the original ball mill powders to the as-extruded material. Materials subjected to high temperature isothermal heat treatments up to 1400 °C, and long term creep at elevated temperatures have also been characterized. In addition, the response of these materials to high dose neutron irradiation has been investigated. Atom probe tomography of the ball milled powder has demonstrated that mechanically alloying forces all the elements into solid solution. Atom probe tomography has also revealed that the oxygen content in the ball milled powder and the as-extruded material is many times higher than the parts per million equilibrium level expected in iron alloys. First principle calculations suggest that this high oxygen level can be attributed to the interaction of oxygen, titanium and yttrium with a high concentration of vacancies. High number densities of titanium-, oxygen- and yttrium-enriched nanoclusters were observed in the as-extruded material and all the isothermally heat treated, crept, and neutron irradiated materials. Some porosity was detected in materials annealed at temperatures of 1000 °C and above. A summary of these results will be presented.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.
10:15 AM - BB1.3
Microstructure and Creep Behavior of Nanocluster-strengthened Ferritic Steels.
Michael Mills 1 , Libor Kovarik 1 , Glenn Daehn 1 , Joachim Schneibel 5 , Chain Liu 4 , Martin Heilmaier 3 , Michael Miller 2
1 Materials Science and Engineering, The Ohio State University, Columbus, Ohio, United States, 5 Institut für Werkstofftechnik und Werkstoffprüfung, Otto-von-Guericke-Universität Magdeburg, Magdeburg Germany, 4 Department of Mechanical Engineering, The Hong Kong Polytechnic University, Hong Kong China, 3 Fachgebiet Physikalische Metallkunde, Technische Universität Darmstadt, Darmstadt Germany, 2 Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractMechanically alloyed, oxide-dispersion-strengthened (ODS) ferritic steels represent a class of materials with excellent creep properties and resistance to neutron radiation. The nature of the complex, nanocluster-oxides in Fe – 12-14% Cr, 2% W, 0.22% Ti + 0.25% yttria have been studied as a function of aging time at temperature with the use of electron diffraction, probe-corrected high-resolution STEM-HAADF imaging, and electron energy loss spectroscopy (EELS). The structure of the nanoclusters that form both homogeneously within the matrix and at the grain boundaries will be presented. It will be shown that the homogeneously dispersed nanoclusters with a diameter of approximately 2 nm have distinct, faceted interfaces that are consistent with low index planes of the Fe matrix. The attractive interaction of dislocations with the nanoclusters has also been studied using bright-field STEM imaging in order to gain insight into the remarkable creep strengthening that they create. In light of these observations, a model of the creep response using a modified Rosler-Artz approach will be discussed. With this approach, it is possible to explain the low creep rates and variation of stress-exponent with stress including the small stress exponents that have been measured at applied stress in these alloys.This research was sponsored by the Division of Materials Science and Engineering, Office of Basic Energy Sciences. U.S. Department of Energy. 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.
10:30 AM - BB1.4
Irradiation Induced Solute Clustering in Under-saturated Ferritic Model Alloys.
Radiguet Bertrand 1 , Estelle Meslin 2 , Alain Barbu 2 , Philippe Pareige 1
1 , GPM-UMR CNRS 6634, Saint Etienne du Rouvray France, 2 SRMP, CEA, Saclay France
Show AbstractIn order to extend end-of-life of pressure water reactors (PWR), the reliability of the vessel must be guaranteed. However, it is well known that neutron irradiation results in hardening and embrittlement of reactor pressure vessel (RPV) steels. This evolution of mechanical properties is partly due to the formation of nanometer-size solute (Cu, Mn, Ni, Si, P) clusters. Thus, in order to predict long time evolution of the mechanical properties of RPV steels, the formation mechanism of solute clusters must be understood. This mechanism is still subject to debate (irradiation enhanced precipitation of thermodynamically stable phase or irradiation induced segregation on point defect sinks). Indeed, excepted for Cu, the solutes atoms usually present in solute clusters observed in neutron irradiated RPV steels are known to be under-saturated, at least considered separately, but solubility limits in more complex alloys remain unknown.In order to clarify this mechanism, under-saturated binary alloys (Fe-1%Mn and Fe-1%Ni) are irradiated at 400°C up to 2 dpa by 10 MeV Fe ions in JANNUS facilities. The irradiation induced microstructure is characterized at very fine scale by atom probe tomography. Thanks to this technique the presence of nanometric solute clusters can be observed and the clusters can be characterized in term of chemical composition, size and number density. In addition to experimental results, cluster dynamic model is used to estimated size and number density of mobile point defects and defect clusters. Comparison between experimental results and model prediction allows to identify relevant mechanisms of solute cluster formation.
10:45 AM - BB1.5
Multiple-beam Irradiation Effects in Ferritic Steels and EB-welded F82H Joint.
Nayouki Hashimoto 1 , Hitoshi Seto 1 , Norihito Yamaguchi 1 , Hiroshi Kinoshita 1 , Somei Ohnuki 1
1 Materials Science and Engineering, Hokkaido University, Sapporo Japan
Show AbstractReduced-activation ferritic/martensitic steel F82H has been developed as one of prime candidate materials for experimental fusion reactors. To estimate irradiation effects in fusion reactor components, multiple-scale modeling has been studied. Modeling activities for irradiation induced microstructural change is quite effective to enhance the capability to predict mechanical properties of the materials during irradiation. Defect activation energies such as vacancy and interstitial migration energies should be estimated to obtain fundamental parameters for the modeling. And also, some of the key issues are the effects of helium and hydrogen on the microstructure evolutions such as swelling, and on the mechanical properties such as fracture toughness or embrittlement. In this study, TEM specimens of electron-beam-welded F82H joint, Fe-8Cr model alloy, and Pure Fe have been irradiated by electron and helium ion beams using a High Voltage Electron Microscope (HVEM) as the experimental evaluation of the modeling and simulations. Electron irradiation experiment indicated that net migration energy of vacancy in the welded metal of F82H tended to be lower compared to that in base metal, which could be relating to difference of carbon concentration in matrix between welded and base metal. In both cases of single beam (electron) and dual beam (electron and helium) irradiation, net migration energies of vacancy were slightly higher than that of interstitial. Furthermore, as helium implantation ratio increases, vacancy migration energies become higher, while there are little differences between interstitial migration energies. This result indicates that vacancy would be trapped by implanted helium due to their strong interaction and appeared to have higher migration energy.
11:30 AM - **BB1.6
Radiation Damage Production and Evolution Near Grain Boundaries in Cu.
Blas Uberuaga 1 , Xian-Ming Bai 1 , Arthur Voter 1 , Richard Hoagland 1 , Michael Nastasi 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIt is well accepted that grain boundaries serve as effective sinks for radiation-induced defects (interstitials and vacancies). However, insight into the atomic-scale origin of this behavior is still lacking. We use molecular dynamics, temperature accelerated dynamics, and molecular statics to study radiation damage phenomena near a sigma 11 grain boundary in Cu over three different temporal regimes: the short-time damage production phase of a collision cascade, the longer-time scales over which defect annihilation and aggregation occur, and the thermodynamic-limiting behavior of the system. We find that both the production and the subsequent annealing of the radiation-induced defects are modified significantly by the presence of the grain boundary. We compare to previous experimental results and identify three regimes over which different thermally activated processes are active, resulting in different responses, both better and worse than large-grained counterparts, of the material to irradiation. Our results show that nanostructured materials have very sensitive response to irradiation and offer new insights into the design of radiation tolerant materials. In particular, we discuss the implications of these results for crystalline ceramic materials, which have been proposed as hosts for nuclear waste immobilization.
12:00 PM - BB1.7
Defect Generation, Defect Migration and Phase Segregation in Irradiated Fe-Cr Alloys.
Fei Gao 1 , Shenyang Hu 1 , Yulan Li 1 , Xin Sun 1 , Howard Heinisch 1 , Charles Henager 1 , Mohammad Khaleel 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractKnowledge of the interactions of radiation-induced point defects, such as vacancies and interstitials, with the exiting microstructures is important for developing an understanding of the kinetics and dynamics of microstructural changes in materials under irradiation. In Fe-Cr alloys experiments show that Cr atoms play an important role in radiation-induced evolution of mechanical properties. Concentrations of 2-18% Cr reduce swelling compared to pure Fe, and a minimum in ductile to brittle transition temperature appears at about 9% Cr concentration. Alloys containing more than 10% Cr exhibit α-α' phase separation resulting in the formation of the Cr-rich α' phase that is a major cause of hardening of thermally annealed materials. Modeling methods, including molecular dynamics (MD) simulations, the dimer method, and a phase field modeling have been combined to study defect generation in bulk and near surface regions, defect migration, and phase separation in Fe-Cr alloys. MD simulations of defect generation in Fe-10%Cr alloy show that the effects of Cr atoms on damage dynamics and final damage states are very small, compared to that in pure Fe, but the fraction of Fe-Cr dumbbells is around 32% of the total interstitials, which is higher than the 10% Cr concentration in the alloy. Although the size distribution and numbers of interstitial clusters in the alloy are similar to those observed in pure Fe, some large vacancy clusters are created, in contrast to pure Fe. In general, Fe interstitials and interstitial clusters are mobile, but some interstitial Fe clusters are trapped by Cr atoms, reducing their mobilities. Also, it is noted that Cr-V clusters are created within cascades with a significant population. These clusters may play an important role in microstructural evolution under irradiation. The migration mechanisms and the corresponding activation energies of these Cr-V clusters have been investigated using the dimer and nudged elastic band methods. It is found that the Cr-V clusters can lead to long-distance migration of a Cr atom by Fe and Cr atoms successively jumping to nearest neighbor vacancy positions, defined as a self-vacancy-assisted migration mechanism, with migration energies ranging from 0.64 to 0.89 eV. Thermodynamic and kinetic data of defects, such as generation rates, size and spatial distributions, and mobilities, which are obtained from atomistic simulations, have been used as inputs for a phase field model of irradiation-induced phase separation in Fe-Cr alloys. The effect of radiation conditions and structural defects including void, grain boundary, and free surface on phase separation kinetics and microstructure evolution are simulated. The preliminary results are in agreement with experimental observations.
12:15 PM - BB1.8
An Experimental Analysis of Helium-induced Hardening Effects.
Yong Dai 1 , Lei Peng 1 , Yina Huang 1
1 , Paul Scherrer Institut, Villigen Switzerland
Show AbstractThe synergetic hardening effect induced by displacement damage and helium has not been studied to a great extent. In the low temperature regime (<~400°C), the large level of irradiation hardening that is observed is dominantly induced by defect clusters and loops. The hardening effect induced by helium seems weak, particularly at lower helium concentrations (<~600 appm). Consequently, the hardening observed in ferritic/martensitic (FM) steels irradiated in spallation targets (containing helium) does not evidently differ from that of FM steels after neutron irradiation. However, in our recent studies significant hardening was observed in the specimens irradiated in STIP at temperatures >~400°C, where no hardening is observed in FM steels after neutron irradiation. The main feature of the microstructures of the STIP specimens was a high-density of small bubbles. This suggests that the observed hardening can be mainly attributed to the bubbles. With annealing at 600°C, defect clusters and loops were removed. This permits evaluating the individual hardening from helium bubbles of different sizes. In the present work, the hardening of effect of helium bubbles with sizes 1 to 2 nm has been determined.
12:30 PM - BB1.9
Insights into Stress Corrosion Cracking Mechanisms from High-resolution Measurements of Crack-tip Structures and Compositions.
Stephen Bruemmer 1 , Matthew Olszta 1 , Larry Thomas 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractRecent results are presented demonstrating applications of cross-sectional analytical transmission electron microscopy (ATEM) to intergranular stress corrosion cracking of austentic stainless alloys in light-water-reactor (LWR) environments. Examinations performed at near-atomic resolutions reveal important microstructural and microchemical details of the cracks and crack tips, and provide insights into the mechanisms controlling degradation. Comparisons of crack characteristics produced during long-term service in LWRs and those in samples tested under well-controlled laboratory conditions are used to identify the critical parameters affecting cracking. Fundamental differences are identified between iron-base stainless steels and nickel-base stainless alloys in oxygenated and hydrogenated water.
12:45 PM - BB1.10
Phase-field Simulation of Inert-gas Bubble Kinetics in Irradiated Metals.
Paul Millett 1 , Anter El-Azab 2 , Michael Tonks 1
1 , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 , Florida State University, Tallahassee, 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).
BB2: Ion-Solid Interactions and Swift Heavy Ions
Session Chairs
Christina Trautmann
Yanwen Zhang
Tuesday PM, April 06, 2010
Room 3012 (Moscone West)
2:30 PM - **BB2.1
Material Modifications and Sputtering by Swift Heavy Ions:Role of the Electron-phonon Coupling.
Marcel Toulemonde 1 , Christian Dufour 1
1 GANIL, CIMAP, CAEN-cedex 5 France
Show AbstractExperimental investigations of ion tracks produced with swift heavy projectiles in the electronic energy loss regime are reviewed [1-4]. On contrary to damage induced by nuclear collisions, track and sputtering [5] shows a strong dependence on material properties. Therefore experimental results are presented for metals, oxides, and ionic crystals separately. Focusing on amorphisable insulators as target material, we present an overview of track phenomena such as the dependence of the track size on energy loss and beam velocity, the critical energy loss for track formation. In the second part, we present a theoretical description of track formation based on an inelastic thermal spike model [2,5-7]. This thermodynamic approach combines the initial size of the energy deposition with the subsequent diffusion process in the electronic subsystem of the target before its transfer to the lattice via the electron-phonon coupling. The track size, resulting from the quench of a molten phase, is determined as a cylinder in which the energy density deposited on the atoms surpassed the energy necessary to melt and the sputtering is linked to surface sublimation. [1] M. Toulemonde, C. Trautmann, E. Balanzat, K. Hjort, A. Weidinger Nucl. Instr. Meth. B 216 (2004) 1.[2] C. Dufour, A. Audourd, F. Beuneu, J. Dural, J.P. Girard, A. Hairie, M. Levalois, E. Paumier and M. Toulemonde J. of Phys.: Condens. Matter 5(1993)4573[3] A. Dunlop and D. Lesueur 1993 Rad. Eff. Def. Sol. 126(1993)123[4] C. Trautmann, M. Toulemonde, K. Schwartz, J. M. Costantini et A. Müller Nucl. Instr. Meth. B164-165(2000)365[5] Z.G. Wang, Ch. Dufour, E. Paumier and M. Toulemonde J. of Phys. : Condens. Matter 6(1994)6733[6] W. Assmann, M. Toulemonde, C. Trautmann Topics Appl. Phys. 110 (2007) 401.[7] M. Toulemonde, W. Assmann, C. Dufour, A. Meftah, F. Studer and C. Trautmann Mat. Fys. Medd. 52 (2006) 263.
3:00 PM - BB2.2
Point Defects and Swelling Induced in Yttria-stabilized Zirconia by Swift Heavy Ion Irradiations.
Jean-Marc Costantini 1 , Francois Beuneu 2
1 DMN/SRMA, CEA/SACLAY, Gif sur Yvette France, 2 LSI, CNRS, Palaiseau France
Show AbstractWe present an extensive study of point-defect creation in yttria-stabilised zirconia (ZrO2: Y) or YSZ exposed to various heavy ions (from C to U) covering an energy range from 100 MeV to several GeVs. A synthesis of results from UV-visible optical absorption spectroscopy and electronic paramagnetic resonance spectroscopy is provided with special emphasis on the respective roles of elastic collisions and electronic excitations. We provide new and deeper interpretations of all these data. The colour-centre production and recovery and ion-induced swelling (out-of-plane expansion), which is an important issue for nuclear applications, are mainly addressed in this survey. It is concluded that F+-type centres (involving singly-ionised oxygen vacancies) are produced by elastic-collision processes. The large threshold displacement energy and defect volume hint that these colour centres might actually be small paramagnetic oxygen-vacancy clusters, most likely divacancies (i.e. F2+ centres). Such a picture is consistent with the inhomogeneous broadening of the optical absorption band, the lack of hyperfine splitting, and weak spin-lattice coupling.
3:15 PM - BB2.3
Formation Mechanisms and Core Structures of Tracks Created by Swift Heavy Ions in Dielectrics.
Yehuda Eyal 1
1 Chemistry, Technion - Israel Institute of Technology, Haifa Israel
Show AbstractKnowledge of the structures of ion damage trails, latent ion tracks, created along wakes of swift heavy ions in dielectric solids, is a prerequisite for advancement of track applications in nanotechnology. Employing small-angle X-ray scattering [1], we have determined the radial electron densities within 25-125 µm long tracks created by GeV uranium, lead, and gold ions in polyimide, polycarbonate, {100} LiF, and (100) muscovite. All derived structures have clearly singled out the mechanisms responsible for track formation. Latent tracks in polyimide and polycarbonate possess ~6 nm wide highly-porous cores. These cores exhibit a specific free volume of ~17-20 nm3 per nm of track length independent of the track length [2-4]. Thus, track formation involves local radiolytic decomposition of the polymer matrix, followed by release of gaseous and volatile alteration products. Latent tracks in {100} LiF also possess highly depleted cores. These cores are ~3 nm wide, and are characterized by 49-74% electron density deficit. This morphology suggests local ion-induced disintegration of the LiF crystal to Li atoms and F2 gas, followed by removal of the gas and of at least a portion of the voluminous Li atoms [5,6]. Latent ion tracks in {100} muscovite possess ~8-10 nm diameter structurally-disordered cores (disorder derived by transmission electron microscopy) with ~4% electron density depletion (unpublished). This morphology suggests the importance of ion-induced thermal dehydration. Water, which shares 5% of the electrons in muscovite, is a vital constituent of the mica crystal. We have shown (unpublished) that a thermal spike model predicts attainment of the lattice breakdown temperature, ~1268 K, within the observed track core regions. 1. Y. Eyal, S. Abu Saleh, J. Appl. Crystal. 40 (2007) 71-76.2. S. Abu Saleh, Y. Eyal, Appl. Phys. Lett. 85 (2004) 2529-2531.3. S. Abu Saleh, Y. Eyal, Nucl. Instr. and Meth. B 236 (2005) 81-87.4. S. Abu Saleh, Y. Eyal, Nucl. Instr. and Meth. B 230, (2005) 246-250.5. S. Abu Saleh, Y. Eyal, Y. Molec. Cryst. Liq. Cryst. 448 (2006) 233/[835]-242/[844].6. S. Abu Saleh, Y. Eyal, J. Appl. Crystal. 40 (2007) s121-s125
3:30 PM - BB2.4
Changing the Phase Transformation Behaviour of ZrO2 and HfO2 by Swift Heavy Ion Irradiation at High Pressure.
Beatrice Schuster 1 2 , Maik Lang 3 , Reinhard Neumann 1 , Christina Trautmann 1
1 Materials Research Dept., GSI Helmholtz Center for Heavy Ion Research, Darmstadt Germany, 2 Solid State Physics Dept., Technical University Darmstadt, Darmstadt Germany, 3 Geological Sciences Dept., University of Michigan, Ann Arbor, Michigan, United States
Show AbstractRecent experiments combine the irradiation with GeV ions and very high pressures (> GPa). Under such conditions the materials response can be quite different compared to pressurization or radiation alone. It was demonstrated that high-pressure phases can be triggered far away from their actual stability field or usually unstable phases can be induced and stabilized to ambient conditions [1, 2]. Pressure is applied by using diamond anvil cells (DAC), where the sample is squeezed in between two 2-3 mm thick diamonds. Irradiations are performed at the heavy-ion synchrotron SIS of GSI where the kinetic energy of the ions is sufficiently large (~200 MeV/u) to pass completely through the first diamond and deposit large amount of energy inside the pressurized samples. This report focuses on zirconia (ZrO2) and its less studied homologue hafnia (HfO2). Both ceramics are known for their high thermal and mechanical resistance and excellent radiation hardness. Because of these properties, ZrO2 is considered as inert fuel matrix in nuclear reactors or as nuclear waste form.Exposure to swift heavy ions at ambient conditions showed that ZrO2 undergoes a phase transformation from monoclinic to its high-temperature polymorph (tetragonal) if the ion fluence exceeds values of about 5×1012 Pb- or U-ions/cm2, or correspondingly higher fluences for lighter ions [3]. Surprisingly, irradiations with heavy ions (Au, Pb, U) performed at pressures between 11 and 13 GPa induce the transition from the monoclinic to the tetragonal phase already at fluence more than one order of magnitude less than for irradiations at ambient conditions [2]. Irradiations at even higher pressures (20-25 GPa) produced the second high-pressure phase, orthorhombic II, far away from its known transition region (above 30 GPa). We could quench this phase to ambient conditions, which appeared to be metastable because it underwent a second phase transformation to the tetragonal high-temperature phase after performing several Raman measurements on the quenched sample. In addition, several fluence and pressure series up to 60 GPa were performed using ZrO2 of different grain size.Raman measurements, complemented by XRD and TEM, demonstrated that high-pressure irradiations influence the phase transformation behaviour of ceramics in ways neither pressure nor radiation alone could. Such experiments offer a powerful tool to induce transformations into the high-pressure and/or high-temperature phases which normally cannot be stabilized to ambient conditions.References:[1]M. Lang, F. Zhang, J. Zhang, J. Wang, B. Schuster, C. Trautmann, R. Neumann, U. Becker, and R.C. Ewing, Nature Materials 8, 793 (2009)[2]B. Schuster, M. Lang, R. Klein, C. Trautmann, R. Neumann, and A. Benyagoub, Nucl. Instr. Meth. B, 267, 964 (2009)[3]A. Benyagoub, Phys. Rev. B, 72, 094114, (2005)
3:45 PM - BB2.5
Threshold Displacement Energies in GaN and SiC: Ab initio Molecular Dynamics Study.
Haiyan Xiao 1 , Fei Gao 1 , William Weber 1
1 , pacific northwest national lab, Richland, Washington, United States
Show AbstractThreshold displacement energy (Ed), as a key physical parameter relevant to defect production under irradiation, is an important quantity for determining damage production rates under irradiation with electrons, neutrons and ions. Large-scale ab initio molecular dynamics method have been used to determine the threshold displacement energies along specific directions with ab initio accuracy and to understand the mechanisms of defect generation during low energy events in SiC and GaN. The values of Ed show a significant dependence on direction. The minimum Ed of 39 eV for Ga and 17 eV for N agree fairly well with experimental measurements. In general, the average threshold displacement energy for Ga recoils is larger than that for N recoils, in contrast to the previously classic MD simulations with empirical potentials. All the N recoil events generate a similar defect configuration, i.e., a N–N dumbbell, whereas the Ga recoil events create a variety of different defect configurations. In SiC, the displacement threshold energies along the principle directions are also determined for both C and Si recoils, and the weighted average Ed values for the four main crystallographic directions are about 25 eV for C and 46 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.
4:30 PM - **BB2.6
Swift Heavy Ion Tracks in Solids: New Insights Using Small Angle X-ray Scattering Experiments.
Patrick Kluth 1
1 Department of Electronic Materials Engineering, The Australian National University, Canberra, Australian Capital Territory, Australia
Show AbstractSwift heavy ion irradiation (SHI) of a solid can produce narrow trails of permanent damage along the ion paths, so called ion tracks. These nanometric objects are interesting in a variety of disciplines including materials science and engineering, nuclear physics, geochronology, archaeology, and interplanetary science. Though average structural properties of ion tracks can often be inferred from macroscopic measurements, the inner track structure remains extremely difficult to retrieve due to the lack of sufficient contrast inherent with most techniques. Synchrotron based small angle x-ray scattering (SAXS) provides an interesting tool to study structural details of ion tracks as it is sensitive to small density changes that often exist in the damaged regions. Furthermore, due to the short acquisition times and the intense photon beam, time resolved measurements of ion track recovery upon annealing and measurements of ion tracks under extreme conditions such as high pressure in diamond anvil cells are possible. The presentation will give an overview of the utilization of synchrotron SAXS for the study of ion track damage and outlines its strength and limitations. Our recent results on SHI damage in a variety of materials will be discussed. This includes measurements of ion tracks in amorphous silica, semiconductors such as InP, and the study of ion track annealing in apatite, a material interesting for geological dating purposes. In amorphous silica for example a fine structure in the radial density distribution comprising a cylindrical core-shell configuration is observed, consistent with a frozen-in pressure wave originating from the centre of the track. At high ion fluences, i.e. where the material experiences multiple coverage with ion tracks, measurements are consistent with a track “annihilation process”. In apatite, a significant change in the density in the track region is apparent, consistent with hollow ion tracks. Upon annealing the track radius decreases and considerable polydispersity is induced in the originally monodispers ensemble of tracks.
5:00 PM - BB2.7
Can Electronic Stopping Just Below the Track Production Threshold Enhance Nuclear Damage Production in Embedded Nanocrystals?
Marie Backman 1 , Flyura Djurabekova 1 , Olli Pakarinen 1 , Kai Nordlund 1 , Leandro Araujo 2 , Mark Ridgway 2
1 , University of Helsinki, Helsinki Finland, 2 , Australian National University, Canberra, Australian Capital Territory, Australia
Show AbstractIon irradiation can be used to modify the properties of nanoparticles that are embedded in a matrix. Depending on the energy and type of the ion, the irradiation can, for instance, cause changes in the shape of the nanoparticles,alter their size distribution, or make crystalline nanoparticles amorphous.Ion tracks are damaged regions in a material caused swift heavy ions ions that excite target electrons. At lower ion energies no track is created, but the ion may introduce damage by colliding with the target atom nuclei and thusinitiating a nuclear cascade. In the present work, we study the behaviour of embedded nanocrystals under ion irradiation with electronic stopping slightly below the track threshold. We examine the possibility of a synergetic effect between the nuclear cascade and electronic energy losses affecting the damage production in a material.By molecular dynamics simulations and experiments, we also study irradiation of embedded Ge nanocrystals in silica by Si and I ions. Due to the difference in mass of these ions, their nuclear stopping in the material differs, which allows us to investigate if the damage in a material can be directly related to the nuclear stopping of the ion in the material.
5:15 PM - BB2.8
Ion-beam Modification of Mechanical Properties of Nanoporous Solids.
Sergei Kucheyev 1 , T. Baumann 1 , A. Hamza 1 , J. Satcher 1 , M. Worsley 1
1 , Lawrence Livermore National Laboratory, Livermore, California, United States
Show AbstractOpen-cell nanoporous solids are used in inertial confinement fusion targets. Either intentional or unintentional ion bombardment will result in modification of nanofoam properties. However, the response of such unconstrained nanostructures to ion irradiation remains essentially unexplored. Here, we study the effect of ion bombardment on mechanical properties of silica and carbon-nanotube-based nanofoams. We use depth-sensing indentation with various indenter geometries (spherical, pyramidal, and flat punches) to measure Young’s modulus, crushing pressure, fracture toughness, and brittleness of nanofoams. We identify irradiation regimes where the dominant mechanism is either a change in foam connectivity or foam densification. Our results show that ion bombardment could be used to improve mechanical properties of low-density nanoporous solids. This work was performed under the auspices of the U.S. DOE by LLNL under Contract DE-AC52-07NA27344.
5:30 PM - BB2.9
AFM Study of the Fluence-Dependent Modification of the Surface of YSZ Due to Argon Ion Irradiation.
Marilyn Hawley 1 , Igor Usov 1 , Jonghan Won 1 , James Valdez 1 , Kurt Sickafus 1
1 Materials Science and Technology Division, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractYittria stabilized zirconia (YSZ) has long been of interest as a promising material for nuclear energy applications. Previous studies of radiation damage in YSZ focused primarily on microstructural changes in the bulk or in the near surface layer whereas irradiation induced changes on the surface received little attention. Here we use atomic force microscopy, AFM, to study the fluence-dependent generation of surface modifications to YSZ due to 150 keV Ar ion irradiation at fluences from 3x1015 to 1x1017 ions/cm2. The microstructural changes in the near surface region were also investigated by Rutherford backscattering spectrometry combined with channeling (RBS/C), grazing incidence X-ray diffraction (GIXRD) and transmission electron microscopy (TEM). Further, we investigated (100), (110), and (111) oriented single crystals of YSZ to explore the difference in sensitivity of crystal orientation to ion irradiation. At the highest fluence level, a dense packing of large round surface hillocks were observed for the first time, to the best of our knowledge. Ion induced surface modification of YSZ will be correlated with bulk damage and concentration of implanted Ar atoms. Evolution of the surface features and effect of surface orientation will be discussed.
Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Amit Misra Los Alamos National Laboratory
Christina Trautmann GSI Helmholtzzentrum
Brian Wirth University of California-Berkeley
BB3: Surfaces, Interfaces and Nanomaterials
Session Chairs
Sergei Kucheyev
Mike Miller
Wednesday AM, April 07, 2010
Room 3012 (Moscone West)
9:30 AM - **BB3.1
Tailoring Nanocomposites for Radiation Tolerance by Atomic-scale Design of Interfaces.
Michael Demkowicz 1 , Kedarnath Kolluri 1 , Amit Misra 2 , Michael Nastasi 2
1 Department of Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, United States, 2 Center for Integrated Nanotechnologies, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIn nanocomposites, interfaces make up such a large fraction of the total material that they may dominate its radiation response. Thus, it may be possible to tailor nanocomposites for radiation tolerance by controlling the properties of the interfaces they contain. This talk will describe a joint experimental/modeling program whose goal is to predict how different metal-metal interfaces trap, transport, and remove radiation-induced defects as well as how this information could be used to design radiation tolerant nanocomposites.This material is based upon work supported as part of the Center for Materials at Irradiation and Mechanical Extremes, an Energy Frontier Research Center funded by the U.S. Department of Energy, Office of Science, Office of Basic Energy Sciences under Award Number 2008LANL1026.
10:00 AM - BB3.2
Microstructural Stability in Nanostructured Cu Alloys During Ion Irradiation at Very High Temperatures.
Nhon Vo 1 , See Chee 1 , Robert Averback 1 , Pascal Bellon 1
1 Materials Science and Engineering, University of Illinois, Urbana Champaign, Urbana, Illinois, United States
Show AbstractThe stability of dilute nanostructured Cu90Mo10 and Cu90W10 films during irradiation with 1.8 MeV Kr+ ions at very high temperatures has been studied using X-ray diffraction and Transmission Election Microscopy. Mo and W nano-particles in Cu remained smaller than ~ 8 nm during irradiations to doses greater than 100 dpa at temperatures greater than 0.64Tm and 0.87Tm, respectively. Results from Molecular Dynamics and Monte Carlo simulations will be shown that help to explain these experimental observations. Nano-particles stability in absence of irradiation arises from the small solute diffusivities and extreme immiscibility (heat of mixing = 28 kJ/mol for Cu50Mo50 and 33 kJ/mol for Cu50W50). During irradiation, viscous displacement of Mo and W atoms and particles during the thermal spike was observed to be the only available mechanism for particle growth. This mechanism, however, limits the nano-particle size to the size of the thermal spikes, which varies in temperature from 5 nm at room temperature to 9 nm at 0.87 Tm. These sizes are in good agreement with experiments.
10:15 AM - BB3.3
Computational Studies of Point Defect Migration Near and Within Grain Boundaries in Cu.
Xian-Ming Bai 1 , Arthur Voter 1 , Richard Hoagland 1 , Michael Nastasi 1 , Blas Uberuaga 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractWe have used Temperature Accelerated Dynamics (TAD) simulations and Nudged Elastic Band (NEB) calculations to investigate the picosecond-scale interstitial and microsecond-scale vacancy migration behavior near and within the symmetric and asymmetric sigma 11 tilt grain boundaries in copper. The diffusion barriers of both defects as a function of the distance from the grain boundary are calculated. In comparison to the single crystal, we find that both boundaries reduce diffusion barriers of defects near the boundary, indicating that grain boundaries can attract and annihilate radiation-induced point defects. Within the grain boundaries, defect migration shows strong anisotropy. The differences between the two grain boundaries in modifying defect migration are also examined. Finally, we will discuss the implication of these results on the relative radiation tolerance of these two boundaries.
10:30 AM - BB3.4
Study of Nano-cluster Formation in ODS Ferritic Steel by Atom Probe Tomography.
Olena Kalokhtina 1 , Radiguet Bertrand 1 , Philippe Pareige 1 , Y. De Carlan 2
1 , GPM-UMR CNRS 6634, Saint Etienne du Rouvray France, 2 Laboritoire de Technologie des Materiaux Extremes, CEA, DEN, DMN Service de Recherches de Metallurgie Appliquee, Gif sur Yvette France
Show AbstractOxide-dispersion strengthened (ODS) alloys obtained by powder metallurgy are potential candidates for future (Gen 4, ITER) nuclear systems due to their attractive properties including high temperature creep strength and resistance to severe neutron exposure. These properties are closely connected with the presence of a high number density of nanometre scale titanium, yttrium and oxygen enriched particles. This microstructure can be obtained by milling of yttrium oxide powder and metallic powders, followed by thermo-mechanical treatments.To understand the fundamental mechanisms of formation and evolution of these nano-scale oxides it is necessary to investigate the material on each stage of the production route. In this work, an ODS ferritic steel was elaborated at CEA Saclay. It was obtained by mechanical alloying of Fe-18wt%Cr-1%W-0,3%Ni-0,3%Mn- 0,15%Si-0.3%Ti pre-alloyed powder and 0.6% Y2O3 powder followed by hot extrusion (1100°C), is studied. Its microstructure is characterised at the atomic scale by mean of atom probe tomography after each step of elaboration. First the as milled powder is investigated to study the dissolution of Y2O3 in ferritic matrix during ball-milling. Then the microstructure is characterised after hot extrusion. In addition, the effect of an annealing at 850°C during 1 hour is investigated, in order to increase understanding and knowledge of the nucleation and growth of nano-clusters.
10:45 AM - BB3.5
Nanostructures from Hydrogen and Helium Implantation of Aluminum.
Markus Ong 1 , Nancy Yang 1 , Ryan DePuit 2 , Bruce McWatters 3 , Rion Causey 1
1 Energy Nanomaterials, Sandia National Laboratories, Livermore, California, United States, 2 Department of Chemical Engineering, University of California, Santa Barbara, Santa Barbara, California, United States, 3 Radiation-Solid Interactions, Sandia National Laboratories, Albuquerque, New Mexico, United States
Show AbstractThis study investigates a pathway to nanoporous structures created by hydrogen implantation in aluminum. Previous experiments for fusion applications have indicated that hydrogen and helium ion implantations are capable of producing bicontinuous nanoporous structures in a variety of metals. This study focuses specifically on implantations of hydrogen and helium ions at 25 keV in aluminum. The hydrogen and helium systems result in remarkably different nanostructures of aluminum at the surface. Scanning electron microscopy, focused ion beam, and transmission electron microscopy show that both implantations result in porosity that persists approximately 200 nm deep. However, hydrogen implantations tend to produce larger and more irregular voids that preferentially reside at defects. Implantations of helium at a fluence of 1018 cm-2 produce much smaller porosity on the order of 10 nm that is regular and creates a bicontinuous structure in the porous region. The primary difference driving the formation of the contrasting structures is likely the relatively high mobility of hydrogen and the ability of hydrogen to form alanes that are capable of desorbing and etching Al (111) faces.
11:30 AM - **BB3.6
A Critical Experiment Using in situ Ion Irradiation to Validate Computer Modeling of Irradiation Damage.
Marquis Kirk 1 , Meimei Li 2 , Pete Baldo 1 , Donghua Xu 3 , Thibault Faney 3 , Brian Wirth 3
1 Materials Science Division, Argonne National Laboratory, Argonne, Illinois, United States, 2 Nuclear Engineering Division, Argonne National Laboratory, Argonne, Illinois, United States, 3 Nuclear Engineering, University California-Berkeley, Berkeley, California, United States
Show AbstractA combined experimental and computer modeling program of radiation damage formed in TEM thin foils of molybdenum during and following in situ ion irradiation in the IVEM-Tandem Facility at Argonne is described. The density and sizes of TEM-visible point defect clusters were measured as functions of foil thickness, ion (1 MeV Kr+) dose and dose rate, and at irradiation temperatures of 80°C and 300°C, below and above vacancy migration in stage III respectively. The effect of foil surfaces on defect density were revealed in the TEM plane view data and confirmed in 3D tomography TEM measurements. The defect evolution in TEM thin foils was simulated by defect cluster dynamic models using reaction rate theory. The computational data were compared with the experimental results and the key model parameters were refined. The findings show that in situ TEM ion irradiation provides critical input and validation data for computational modeling.
12:00 PM - BB3.7
Mechanical Properties of Ion Irradiated Nanoscale Cu/Nb Multilayers.
Nan Li 1 2 , Michael Nastasi 2 , Amit Misra 2
1 , Texas A&M University, College Station, Texas, United States, 2 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractA series of helium (He) ion irradiations with different energies and doses were performed at room temperature to produce a near constant He concentration of 6 ± 1 at% over a depth of 1 um in Cu-Nb multilayers with individual layer thicknesses ranging from 2.5 nm to 50 nm. Transmission electron microscopy shows He bubbles, approximately 1 nm in diameter, throughout the multilayered films, but the average helium bubble density, swelling and lattice expansion reduce with decreasing bilayer period. Focused-ion-beam machined micropillar specimens were used to obtain compressive stress-strain curves for Cu/Nb multilayers before and after ion irradiation. Consistent with the reduction in damage with reducing layer thickness, the magnitude of radiation hardening decreases significantly with decreasing layer thickness, without any measurable loss in deformability. Radiation hardening in nanoscale multilayers is discussed in terms of the influence of He bubbles on different unit processes that become active with decreasing layer thickness such as confined layer slip and interface crossing of single dislocations.
12:15 PM - BB3.8
In-situ Surface Characterization of Nanostructured Materials Exposed to Controlled Irradiation Fields.
Jean Paul Allain 1 2 3 , Osman El-Atwani 3 2 , Daniel Rokusek 1 , Chase Taylor 1 2 , Bryan Heim 1 2 , Zhangcan Yang 1
1 Nuclear Engineering, Purdue University, West Lafayette, Indiana, United States, 2 Birck Nanotechnology Center, Purdue University, West Lafayette, Indiana, United States, 3 Materials Engineering, Purdue University, West Lafayette, Indiana, United States
Show AbstractThe study of surfaces and interfaces and their interaction with radiation and plasma is of great importance to tailor function in low-dimensional systems such as ultrathin multilayers and ion-induced patterned nanostructures. In particular controlling matter at spatial scales the order of the first layer of atoms at the vacuum/film interface is important for applications in materials exposed to extreme environments such as high-density plasma, nuclear fusion edge plasmas and high current density ion-beams. Directed irradiation synthesis enables the modification of multi-component surfaces self-organizing into regular nanoscale patterns. Recent work has demonstrated the importance of metal seeding and compositional modulation of nanopatterns [1,2]. In this work design of functional properties for irradiation-driven surfaces is enabled by in-situ characterization during exposure to well-defined and controlled irradiation fields. The resulting nanopatterns can induce changes in the electronic and mechanical properties of these materials. We present in this work the study of two distinct material systems exposed to controlled irradiation environments during surface characterization in-situ. The first, ultra-thin coatings (e.g. Li, Sn) irradiated by low-energy ions near the sputtering threshold. The second, irradiation of III-V compound semiconductor surfaces (GaSb and InP). We study the effect of compositional and chemical states of these multiple-component heterogeneous surfaces on ion-induced nanoscale processes (e.g. erosion dynamics, redeposition, surface diffusion kinetics) and physical properties (i.e. optical, electronic, hydrogen recycling). A recently built experimental facility at Purdue University can perform in-situ atomic scale characterization of elemental, chemical, and electronic properties using complementary surface-sensitive techniques. In-situ techniques used include: low-energy ion scattering spectroscopy (LEISS), angular-resolved photoemission spectroscopy (ARPES), EUV reflectivity, X-ray photoelectron spectroscopy, erosion measurements and ultraviolet photoelectron spectroscopy (UPS). UPS/ARPES combined with LEISS can give chemical state and elemental information at the first 2-3 monolayers, respectively. The facility is equipped with ion sources capable of delivering hyperthermal and energetic heavy-ions at energies between 10-1000 eV with current densities above 100 uA/cm^2. Both ion and electron spectroscopies are conducted using a highly sensitive hemispherical sector energy multi-channel analyzer. Results elucidate on the synergistic effect between multiple particle-beam exposure on the electronic and optical properties of transition-metal EUV reflective coatings and surface nanopatterning of III-V compound semiconductors.[1] V.B. Shenoy et al. PRL 98, 256101 (2007)[2] G. Ozaydin-Ince et al. J. Phys. Condens. Matter. 21, 224008 (2009)
12:30 PM - **BB3.9
Response of Nanostructured Ceramics Under Extreme Radiation Environments.
Jie Lian 1 , Jiaming Zhang 2 , Fengyuan Lu 1 , Antonio Fuentes 3 , Alexandra Navrotsky 4 , Fereydoon Namavar 5 , Rodney Ewing 2
1 , Rensselaer Polytechnic Institute, Troy, New York, United States, 2 , University of Michigan, Ann Arbor, Michigan, United States, 3 , Unidad Saltillo, Saltillo Mexico, 4 , University of California at Davis, Davis, California, United States, 5 , University of Nebraska Medical Center, Omaha, Nebraska, United States
Show AbstractAtomic-level understanding of radiation interactions with surfaces and interfaces, including grain boundaries and precipitate-matrix interfaces, is crucial for the development of advanced nuclear materials that can withstand the extreme radiation environments in reactors, accelerators, and even geologic repositories. Materials design at the nanoscale will play a key role in mitigating radiation damage and for developing radiation tolerant materials. In this talk, we highlight recent progress in understanding response of nanostructured ceramics under extreme radiation environments by focusing on radiation-induced amorphization, grain coarsening and phase transition among different polymorphs. The nanostructured ceramics under investigation include fluorite-related structure such as pyrochlore and zirconia, potential host phases for actinides incorporation, and nanocrystalline titanate. Systematic ion beam irradiation studies suggest that nanostructured materials are not inherently radiation tolerant and there is a critical length scale (size) at which nanostructured materials are optimally radiation tolerant. For nanocrystalline pyrochlores, 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 correlation among the tendency of phase transformation, crystal size and structure, defect production and dynamic annealing, and the thermodynamic properties for nanostructured zirconia and titania will also be discussed.
BB4:Nuclear Fuels
Session Chairs
Wednesday PM, April 07, 2010
Room 3012 (Moscone West)
2:30 PM - **BB4.1
An Atomic Insight of the Primary Irradiation Damages in Nuclear Fuel UO2 With Molecular Dynamics Simulations.
Laurent Van Brutzel 1 , Alain Chartier 1
1 DPC/SCP/LM2T, CEA, Gif-sur-Yvette France
Show AbstractNuclear fuel, UO2, undergoes during is lifetime different type of irradiation which produce atomic point defects. Understanding at the atomic scale the creation of the primary damage is then of great interest. Molecular dynamics (MD) because of its intrinsic length and time scales is particularly suitable to study point defects creation under irradiation events. We will present herein a review of several MD simulations of collision cascades, thermal spikes in UO2 single- and poly-crystals which have been carried out on massive parallel computers and give an insight of the primary irradiation damages.First, collision cascades with initial energies range from 5 to 80 keV have been carried out in single-crystals UO2 to model the effects of a recoil nucleus. Only few Frenkel pairs are created and their number is lower than the number predicted by NRT law due to numerous recombinations. Clustering of vacancies has been found in the core cascades for high energy cascades. Study of cascade overlap sequence shows saturation on the number of point defects created as the irradiation dose increases.Influence of grain boundaries on the production of point defects has also been studied. Collision cascades up to 80 keV have been initiated near different symmetrical tilt grain boundaries. In all the cases numerous point defects are created along or near the interfaces. However, for the grain boundaries with Schottky defects along the interface (misorientation angles superior to 15°) cascades seem stopped by the interface. Damage created by in poly-crystal UO2 in which orientations and shapes of grains are randomly distributed has been also studied.Simulations of thermal spike in single- and poly-crystals UO2 have been as well carried out in order to model fission tracks. The results show different primary damage type than for collision cascades. In the case of the single-crystal, for small linear energy transfer (4 kev/nm) no defects are created. For higher linear energy transfer (16 - 66 keV/nm) dislocation loops are found around the fission tracks. In the case of the poly-crystals, fusion and recrystallization of grains occurs.Stability and the formation of xenon clusters in several extended defects such as grain boundaries and nanocavities in single-crystal UO2 have been calculated. Static calculations show that xenon atoms are more likely to aggregate than staying homogeneously distributed. For xenon density higher then 0.12 mol/cm3 in bubbles, the xenon crystallizes into fcc structure and small dislocation loops appear in the UO2 matrix.Finally, studies of crack initiation in irradiated and unirradiated single- and poly-crystal UO2 have been investigated. Cracks initiate at lower strain for the irradiated UO2 due to the presence of point defects and intergranular propagation occurs in the poly-crystal.
3:00 PM - BB4.2
The Atomistic Modelling of Realistic Displacement Cascades Within Uranium Dioxide Fuel.
Clare Bishop 1 , Boris Dorado 2 , Robin Grimes 1 , Simon Middleburgh 1 , David Parfitt 1
1 materials, imperial college london, London United Kingdom, 2 , CEA DEN, DEC, Centre de Cadarache, Saint-Paul-Lez-Durance, France
Show AbstractIntra-granular gas bubbles are observed to form in uranium dioxide as part of the fission process within a nuclear fuel reactor. It has been postulated that these bubbles (which degrade the thermal conductivity of the fuel) form in highly defective regions occurring within the wake of fission track events. Therefore, it is of great importance to further understand how high energy cascades such as these determine the evolution of the structure of the fuel (in terms of point defects and defect clusters) when a cascade takes place. The method that we employ is molecular dynamics simulations in which the velocity of a primary knock-on atom (PKA) is rescaled to a high velocity and the structure of the uranium dioxide is monitored to long time scales. This technique relies on using accurate pair potentials which are capable of modelling non equilibrium events. The suitability of the current uranium dioxide pair potentials, with regard to high energy cascade simulations, will be discussed. Modification of the current techniques is explored with the aim of improving the description of the evolution of the defective region in comparison to the existing simulation methods.
3:15 PM - BB4.3
Mesoscale Simulations of Pore and Bubble Migration in a Nuclear Fuel.
Bala Radhakrishnan 1 , Sarma Gorti 1
1 , Oak Ridge National Laboratory, Oak Ridge, Tennessee, United States
Show AbstractThe evolution of microstructure in a nuclear fuel is strongly influenced by the local fuel temperature, irradiation conditions and the temperature gradient. The evolution of the fission gas and the fabrication porosity under such conditions is simulated using a Potts model approach that allows for porosity/bubble evolution by surface and volume diffusion as well as the concurrent migration of grain boundaries that are pinned by the bubbles and pores. The simulations realistically capture the random migration of the gas bubbles in the fuel , the directed migration of the pores and bubbles in a temperature gradient as a function of the pore/bubble size, and also the interaction between pores/bubbles and grain boundaries. The overall evolution of the pore/bubble structure is simulated at the length scale of the cross-section of the fuel pin with a high spatial resolution using high performance computing. The evolution of the fission gas is modeled by quantifying the percolation path for the gas bubbles in the microstructure encompassing the entire fuel cross section as function of the operating parameters. Research funded by the DOE Office of Nuclear Energy, Advanced Fuel Cycle Initiative.
3:30 PM - BB4.4
Steam and Air Oxidation Behavior of Nuclear Fuel Claddings at Severe Accident Conditions.
Mirco Grosse 1 , Martin Steibrueck 1 , Juri Stuckert 1
1 Institute for Material Research, Karlsruhe Institute of Technology, Karlsruhe Germany
Show AbstractHypothetical scenarios of severe accidents in nuclear reactors are studied experimentally in the KIT QUENCH program. The accident starts with loss of the coolant and results in a overheating of the reactor core. Reflood of the reactor core is the main measure to terminate such an accident. The combination of water and the overheated core results in steam oxidation of the nuclear fuel claddings made of zirconium alloys. The oxidation is connected with a strong heat production which can result in a serious temperature escalation. In some hypothetical cases the accident occurs not only in steam but in a mixed atmosphere containing air or nitrogen. In the paper the isothermal steam and air oxidation behaviour of different commonly used cladding materials is compared in the temperature range between 800 and 1400°C. The steam oxidation reactions show a parabolic kinetics, at least at the beginning. The temperature dependence is of Arrhenius type. At about 1000°C for Zr-Sn alloys and at about 1050°C for Zr-Nb alloys the activation energies of the processes changes due to the different crystallographic structures of the oxides. Whereas at temperatures of 1050°C and above the kinetics keep parabolic, at lower temperatures the reaction can be accelerated by a tetragonal to monoclinic phase transformation in the oxide layer which results in the formation of a certain crack structure, a spalling of oxide parts and with it an acceleration of the oxidation. The temperature ranges in which this breakaway effect occurs and the thickness of the oxide parts separated by the formed cracks differ between the various alloys. The differences are caused not only by the chemical composition but also on the cladding production process.The oxidation in pure oxygen is similar to the oxidation in steam. In contrast to this, the reaction in air strongly differs from the behaviour given above. After a certain time, the reaction kinetics is fastened. The reason is the reaction of nitrogen with oxygen vacancies at positions were local oxygen or steam starvation takes place. Zirconium nitrides are precipitated in the oxide. If the starvation conditions are finished the nitrides are oxidized again. Both transformations are connected with a volume change and with the formation of cracks in the oxide layer. The oxide looses its protective behaviour and the reaction kinetics changes to a nearly linear one. The differences between oxidation in steam or pure oxygen at the one side and in air at the other side are as strongest at the intermediate temperatures. At temperatures below 800 °C and above 1400 °C the effect of nitrogen is smaller.
3:45 PM - BB4.5
Modeling the Effect of Stress on Defect Migration and Void Formation With a Finite Element-based Phase Field Method.
Michael Tonks 1 , Paul Millett 1 , Anter El-Azab 2 , Dieter Wolf 1
1 Fuels Modeling and Simulation , Idaho National Laboratory, Idaho Falls, Idaho, United States, 2 Scientific Computing, Florida State University, Tallahassee, Florida, United States
Show AbstractMicrostructural defects play an important role in determining the internal stress in materials. They can also alter the applied stress distribution, e.g. stress concentrations due to voids. The internal stress, in turn, affects the defect dynamic within the material. In this work, we model the internal stresses in irradiated materials, taking into consideration the intrinsic defect stress as well as their interactions with each other and with the applied stress. To capture these phenomena, we solve the phase field equations with the finite element method (FEM). This FEM phase field model is applied to the problem of point defect diffusion and microstructure evolution in irradiated materials under an applied load.
4:30 PM - **BB4.6
Properties and Irradiation Behaviour of High Burnup Nuclear Fuels.
Vincenzo Rondinella 1 , Thierry Wiss 2 , Dragos Staicu 2 , Jean-Pol Hiernaut 2 , Jose Spino 3 , Dimitrios Papaioannou 1
1 Hot Cells, JRC-ITU, Karlsruhe Germany, 2 Materials Research, JRC-ITU, Karlsruhe Germany, 3 Nuclear fuels, JRC-ITU, Karlsruhe Germany
Show AbstractDuring irradiation in nuclear reactor, each atom in the fuel experiences several thousand displacements from its initial lattice position. The properties of the fuel change significantly as a function of burnup, due the effect of radiation damage, power and temperature profiles, and of accumulation of fission and neutron absorption products. Defects generated in the fuel structure (point and extended defects, micro- and macro-bubbles, solute and segregating impurities) will alter key properties which determine the performance of the fuel, like e.g. thermal conductivity, density and mechanical properties. Fission gas production will contribute to fuel swelling and eventual pressurization of the fuel rod. At medium-high burnup, the fuel undergoes a restructuring of its structure, through grain subdivision and redistribution of gases and extended defects. Future reactor concepts, as pursued by international efforts like e.g. GenIV, envisage the use of fuel (and materials) up to higher burnup and more severe irradiation conditions compared to Light Water Reactors (LWR) currently in operation. In the perspective of future developments of advanced reactors and related fuel cycles the focus of fuel studies is shifting from standard LWR UO2 fuel towards Pu- and minor actinide-containing fuel, starting with mixed U-Pu oxide, and including also non-oxide systems.This presentation shows highlights from studies performed at the Institute for Transuranium Elements (ITU) on thermal transport, fission gas behaviour, mechanical properties and microstructural evolution of nuclear fuel as a function of burnup and irradiation conditions, with particular emphasis on high and very high burnup features. The main challenges facing the experimental characterization of this type of materials will be highlighted.
5:00 PM - BB4.7
Helium Solubility Determination in UO2 Single Crystals.
Thierry Wiss 1 , Emilio Maugeri 1 , Vincenzo Rondinella 1 , Rudy Konings 1
1 , European Commission - JRC - ITU, Karlsruhe Germany
Show AbstractAfter several centuries of disposal the radioactivity of the nuclear fuel used in today's reactors will be mostly due to α-decaying actinides. This will generate large quantities of helium in the spent fuel matrix. The extreme consequence of helium formation and subsequent release from the fuel could be the overpressurization of the fuel rod followed by its failure. If helium is retained within the fuel matrix, microstructure alteration ultimately affecting the mechanical stability of the fuel could occur. Moreover, interaction of the fuel matrix with external helium may be relevant in special cases, such as for fuels envisaged for GEN IV He-cooled reactors, namely VHTR and GFR, which will operate at high temperature. The knowledge of thermodynamic quantities like helium solubility is particularly important as a boundary condition to assess kinetic properties like e.g. the diffusion coefficient of helium in the UO2 matrix. High temperature and high pressure helium infusion followed by thermal desorption spectrometry has been used to determine the solubility as well as to investigate the mobility of He in UO2. The role of intrinsic defects as a function of temperature has been taken into account when assessing the different diffusion mechanisms. Parallel studies on alpha-damaged samples allowed obtaining deeper insight on the specific effect associated to various types of defects, including single and extended defects.
5:15 PM - BB4.8
Influence of Irradiation Temperature on HfO2/MgO/HfO2 Multi-layer Structures.
Igor Usov 1 , James Valdez 1 , Jonghan Won 1 , David Devlin 1 , Robert Dickerson 1 , Marilyn Hawley 1 , Yongqiang Wang 1 , Gordon Jarvinen 1 , Kurt Sickafus 1
1 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractIt may be possible to improve the performance of nuclear fuels by employing composite materials. A simple example of such a material is a two-phase ceramic structure consisting of fissile and non-fissile components. By choosing an appropriate non-fissile component, important properties of the composite such as thermal conductivity, release of fission products, corrosion resistance, hardness and fracture toughness can be improved. However, these properties inevitably degrade during operation in a nuclear reactor environment. One reason for this degradation is radiation damage accumulation and attendant microstructural changes. For this study, we fabricated a multi-layer structure composed of MgO and HfO2 thin films. This multi-layer structure represents an idealized, CERCER (ceramic-ceramic) composite fuel form. The MgO is intended to represent a non-fissile component, while the HfO2 is a surrogate for a phase containing fissile species (U, Pu, etc.). In this study, we examined radiation damage evolution over a wide range of irradiation temperatures, from very low temperatures (~100 K), wherein thermally-activated processes are suppressed, to high temperatures (1300 K), corresponding to conditions in the core of a nuclear fuel pellet. To simulate radiation damage conditions in a nuclear fuel pellet, we irradiated our multi-layer structures with 10 MeV Au ions to a fluence of 5x1015 Au/cm2. The temperature during irradiation was varied from 100 K to 1300 K. Microstructural changes were examined using transmission electron microscopy (TEM) and grazing incidence X-ray diffraction (GIXRD). Inter-mixing between the composite layers was analyzed using Rutherford backscattering spectrometry (RBS) and scanning-TEM (STEM) elemental mapping. Surface topography changes were observed using Nomarsky optical microscopy and atomic force microscopy (AFM).
5:30 PM - **BB4.9
Advanced Measurement Techniques for Highly Radioactive Materials.
J. Rory Kennedy 1 , Matthew Fig 1 , James Cole 1 , Dawn Janney 1
1 AFCI Program, Idaho National Laboratory, Idaho Falls, Idaho, United States
Show AbstractIn the realm of radioactive materials, a major challenge to the development of materials is the measurement of the properties for which the material is being developed. For example, the phenomenon of microstructure evolution of a nuclear fuel in reactor is well known but the details of the effects of the change on the behavior of such important subjects as thermal conductivity, mechanical properties, and phase formation have not been quantified at the grain size level. There is a strong need to develop or adapt advanced instrumentation for measurements on radioactive materials. Idaho National Laboratory has an ongoing effort to develop or adapt a variety of measurement techniques to highly radioactive materials. A laser based device called the Scanning Thermal Diffusivity Microscope, conceived and developed over the past few years, has recently been installed in a hot cell where examinations of fresh and irradiated fuel samples have begun in order to profile the thermal diffusivity of fuels and materials at 50µm spatial resolution. Further development effort is the unique application of dual-beam focused ion beam (FIB) microscopy to highly radioactive materials, which will be an exceedingly useful tool to prepare samples for study by nano/micro indentation or compression testing (also under development), transmission electron microscopy, or even atom probe tomography. Finally, the application of micro-focus X-ray diffraction will allow crystal phase identification at spatial resolution of 10-100 µm. This contribution will present the current state of the implementation plan of these instruments to highly radioactive materials. Material study examples will be given and the challenges met to implementing these measurement techniques to highly radioactive materials will be discussed.
BB5: Poster Session: Radiation Effects
Session Chairs
Thursday AM, April 08, 2010
Salon Level (Marriott)
9:00 PM - BB5.1
Direct Energy Conversion From Gamma Ray to Electricity Using Silicon Semiconductor Cells.
Kenichi Hashizume 1 , Hiroki Kimura 1 , Teppei Otsuka 1 , Tetsuo Tanabe 1 , Tomio Okai 2
1 Interdisciplinary Graduate School of Engineering Sciences, Kyushu Univ., Fukuoka Japan, 2 Applied Quantum Physics and Nuclear Engineering, Kyushu Univ., Fukuoka Japan
Show AbstractSpent fuels and high level radioactive wastes which emit high doze of gamma rays could be a promising and long-lasting power source, if the gamma ray energy was effectively converted other forms of energy. In the present study, we have tried to convert gamma ray to electricity directly, with using silicon semiconductor cells made of p-type Si single crystal wafers with various specific resistivities ranging from 0.01 to 1000 Ohm-cm. On both surfaces of the cell (20x20x0.5mm3), Al and Sb were deposited in vacuum to make electrodes at room temperature. The voltage-current measurement of the cells showed a rectification effect, and Al side was found to work a cathode. This suggests a Schottky junction was formed at the interface between the deposited Al and Si wafer. The cell irradiated by gamma ray in Co-60 irradiation facility in Kyushu Univ. with an absorbed dose of about 200Gy/h, and output voltage and current generated by the irradiation with external resistances varying from 200 to 100,000 Ohm were measured. The maximum electric power obtained for each cell ranged from 0.002 to 200 micro-W/m2, and clearly increased with increasing the specific resistivity of Si wafers. For comparison, a single crystal Si solar cell (2400mm2x0.5m, 0.5Vx450mA in AM1.5 condition) was also exposed to the gamma ray, and its maximum electric power was 2 micro-W/m2. The output power of the present cell with high resistivity was two orders of magnitude higher than that of the Si solar cell.Energy deposition in the Si cell during gamma irradiation was evaluated with the Monte Carlo Simulation for N Particles (MCNP) code. For Si with 0.5 mm thickness, the deposited energy was calculated to be 17000 micro-W/m2 for 200Gy/h. Comparing the output energy by the gamma irradiation, the energy conversion efficiency of the present Si cells reached about 1%. Unfortunately, the present cells were unstable even in ambient atmosphere, the conversion ratio of which decreased to less than one tenth in six months. Further development of the cells with higher conversion ratio and improvement of its stability will be discussed.
9:00 PM - BB5.11
Radiation Effects in Mn+1AXn Phases: Ti3AlC2 and Ti3SiC2.
Karl Whittle 1 , Daniel Riley 2 , Mark Blackford 1 , Robert Augtherson 1 , Sam Moricca 1 , Nestor Zaluzec 3 , Gregory Lumpkin 1
1 Materials Engineering, ANSTO, Menai, New South Wales, Australia, 2 School of Engineering, Unviersity of Melbourne, Melbourne, Victoria, Australia, 3 Electron Microscopy Center, Argonne National Laboratory, Chicago, Illinois, United States
Show AbstractFuture nuclear power plants, whether they are fission (Gen IV) or fusion (ITER/DEMO), represent viable options to reduce the carbon footprint of energy generation. However, these new designs require materials that are not only tolerant of the high temperature but also resistant to radiation damage, in some designs chemical attack, and have low neutron absorptions. Ternary carbide materials, i.e. Mn+1AXn have been proposed as meeting these criteria with applications within these future technologies. In order for any material to be applied the ability of the materials to tolerate radiation damage needs to be understood and in particular predicted.Two systems, specifically Ti3AlC2 and Ti3SiC2 have been studied to determine their radiation tolerance, using in-situ ion beam irradiation of 1 MeV Kr and Xe ions, coupled with transmission electron microscopy. Experiments have shown that Ti3AlC2 in particular shows no appreciable amorphisation at 300K up to doses of at least 6.25x1015 ions cm-2 (~25-30 dpa). Interestingly there is an observed difference between Ti3AlC2 and Ti3SiC2.Explanations and possible mechanisms for recovery from damage are presented, along with implications for future potential uses.
9:00 PM - BB5.12
Interaction Barrier Effect of Zr-O-N Thin Film Deposited by Hydrothermal Deposition for Sodium-cooled Fast Reactor.
Seung Hyun Jee 1 , Su Hyun Kim 1 , Jun Young Oh 1 , Nitul Kakati 1 , Young Soo Yoon 1
1 Department of Materials Science and Engineering, Yonsei University , Seoul, Seodaemoon gu, Korea (the Republic of)
Show AbstractThe unique properties of U-Zr metal fuel in nuclear power plant, such as high thermal conductivity, proliferation resistance, ease of fabrication, and a good compatibility for sodium have stimulated research in Sodium-cooled Fast Reactor (SFR) for burning the long-lived fission products. It is very hard to use U-Zr metal fuel directly since actinide elements have a fuel-clad chemical reaction (FCCI) or eutectic reaction with stainless steel based cladding material even at just above 650 oC. The most common steady-state FCCI reaction is interchanged by diffusion of cladding components (Fe, Ni, etc.) and rare earth fission products. (Nd, Ce, Sm, etc.) The interaction layer in the cladding is very brittle and considered as having no strength. The barrier concept is needed to prevent the fission product interdiffusion as well as to prevent 'eutectic' melting during transient events.Hydrothermal deposition of Zr-O-N (ZON) thin films onto a HT9 clad was carried out to prevent a FCCI and a eutectic reaction of a U-10wt% Zr metal fuel with clad material of nuclear power plant. ZON thin films were deposited with various pH values(2, 5, 8) using NH3OH at 140 oC for 5 hours. X-ray diffraction (XRD) patterns indicated that an amorphous structure of the hydrothermal deposition grown ZON thin films regardless of growth temperature and ph variation even though typical crystal-like grain structures were found in plane and cross sectional images of the as deposited ZON thin films with thickness range of 400 ~ 600 nm. A depth profile of auger electron microscopy (AES) revealed that carbon, nitrogen and oxygen atoms were found in the as deposited ZON thin film. Diffusion couple tests at 800 oC for 25 hours showed that the as deposited ZON thin films with high nitrogen content confirmed by Rutherford backscattering spectrometry (RBS) showed a barrier behavior for FCCI between the metal fuel and the clad. This result suggested that nitride form of ZON thin film, even with the thickness below um level, has very high possibility as an effective barrier against FCCI.
9:00 PM - BB5.13
Damage Induced by Carbon Irradiation in Silicon Carbide.
Nihed Chaabane 1 , Patrick Trocellier 2 , Sandrine Miro 2 , Gael Sattonnay 3 , Yanwen Zhang 4 , William J Weber 4 , Constantin Meis 1
1 INSTN/UESMS, CEA/Saclay, Gif sur Yvette France, 2 DMN/SRMP, CEA/Saclay, Gif sur Yvette France, 3 ICMMO/LEMHE, Université Paris Sud, Orsay France, 4 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show Abstract6H-SiC and 3C-SiC carbon were implanted at room, 670 K and 160 K temperatures and high energy with fluencies ranging between 1015 and 1016 cm-2 at JANNUS facilities. Damage accumulation was studied by combining X-ray diffraction, Raman spectroscopy and cross-sectional transmission electron microscope. A long the ion path X-ray diffraction curves show a lattice strain gradient perpendicular to the surface. This dilatation of the c axis of the lattice disappears after high temperature annealing implying the structural recovery of all the crystalline perturbed regions. Moreover the Raman spectroscopy observations show the formation of homonuclear bonds Si-Si and C-C within the SiC network. Several Si-C bands disappear and the bands become broader. The increase of the irradiation temperature induces a slowing down of the disorder production due to an enhancement of the dynamic annealing.
9:00 PM - BB5.14
Modelling and Experimental Characterization of a Residual Stress Field in a Ferritic Compact Tension Specimen.
Mark Wenman 1 , Alexander Price 2 , Paul Chard-Tuckey 3
1 Materials, Imperial College London, London United Kingdom, 2 , Fraser Nash Consultancy, Risley, Lancashire, United Kingdom, 3 Nuclear Department, Defence Academy, Gosport, Hants, United Kingdom
Show AbstractThe aim of the work is to elucidate the influence of plasticity behaviour on the residual stress field in a ferritic reactor pressure vessel steel. To this end, we investigate two compressively pre-loaded compact tension (CT) specimens to generate a mechanical residual stress field. One specimen was subsequently pre-cracked by fatigue before both specimens were measured using high-energy synchrotron X-ray diffraction. A fine grain size microstructure (~5-10 μm grain size) allowed a small X-ray beam slit size and therefore gauge volume. The results provide an excellent data set for validation of finite element (FE) modelling predictions against which they have been compared. The results of both mechanical testing and modelling suggest that the use of a combined hardening model is needed to accurately predict the residual stress field present in the specimen after pre-loading. Some discrepancy between the modelled crack tip stress values and those found by X-ray diffraction remain which can be partly explained by volume averaging effects in the presence of very high stress/strain gradients.
9:00 PM - BB5.15
Microstructure Evolution and Phase Transformation Induced by Plasma Processing Techniques Used for Improvement of Structural Nuclear Materials Properties.
Victor Andrei 1 , Cristian Lungu 2 , Oana Rusu 1 , George Oncioiu 1 , Florea Silviu 1 , Mihai Chivoci 1
1 Surface Analysis Laboratory, Institute for Nuclear Research, Pitesti Romania, 2 , National Institute for Laser, Plasma and Radiation Physics, Bucharest Romania
Show AbstractThe production of nanostructured surface films with controlled properties is crucial for the development of materials necessary for the Advanced Systems for Nuclear Energy. Since the surface of materials is the zone through which materials interact with the environment, the surface science and surface engineering techniques plays an essential role in the understanding and control of the processes involved. Complex surface structures were developed on stainless steels used as structural nuclear materials: austenitic stainless steels based on Fe (304L, 316L), austenitic steels with high content of Cr, ferrites resistent to corrosion, by various Plasma Processing methods which include:-Plasma Electrolytic (PE) treatments: the steel substrates were modified by nitruring and nitro-carburizing plasma diffusion treatments;-carbonic films deposition in Thermionic Vacuum Arc Plasma.Microstructure evolution and phase transformation induced by nitruring and nitro-carburizing plasma diffusion treatments are studied.The results of the characterization of surface structures obtained in various experimental conditions for improvement of the properties (corrosion resistance, hardness, wear properties) are reported: the processes and structures were characterized by correlation of the results of the complementary techniques: SEM, XRD, LAXRD, XPS, “depth profiling”, EIS. An overall description of the processes involved in the surface properties improvement, and some consideration about the new materials development for energy technologies are presented.
9:00 PM - BB5.16
Radiation-Induced Amorphization and Recrystallization in Oxides.
Yanwen Zhang 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, 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. The effects of displacement damage in Sr2Nd8(SiO4)6O2, SrTiO3, Ho2Ti2O7 and Sm2Ti2O7, as representative materials, are studied under energetic Au irradiations. The disorder, due to atomic collisions processes, increases nonlinearly with irradiation dose and can be well described by a disorder accumulation model that includes contributions from point defects, defect clusters and amorphous domains. Ionization from 200 keV electrons induces recrystallization at the amorphous/crystallineinterface that exhibits several distinct stages associated with residual defect annihilation near the interface, epitaxial regrowth at the interface, and a surface-stabilized amorphous state. The results suggest that understanding ionization effects and the coupled effects of electronic and atomic dynamics on material behaviour is a critical area for nuclear applications.
9:00 PM - BB5.17
Enhanced Ion Radiation Tolerance Properties in TiN/AlN Nanolayer Films.
Ickchan Kim 1 , Michael Martin 2 , Lin Shao 2 , Xinghang Zhang 3 , Greg Swadener 4 , Yongqiang Wang 4 , Haiyan Wang 1
1 Electrical and Computer Engineering, Texas A&M University, College Station, Texas, United States, 2 Nuclear Engineering, Texas A&M University, College Station, Texas, United States, 3 Mechanical Engineering, Texas A&M University, College Station, Texas, United States, 4 , Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractRecently nitride ceramics are considered to be one of the advanced radiation tolerance materials with superior thermal/mechanical and diffusion barrier properties. As the demand for enhanced tolerance property exceeds the capabilities of existing single phase bulk materials, new materials with superior properties are needed. In this study, we investigated interfacial effect on ion-irradiation tolerance properties in nanolayered films. A set of nanolayer TiN/AlN samples composed of each layer thickness varied from 5 nm to 50 nm, was prepared by pulsed laser deposition (PLD). Detailed TEM and nanoindentation studies were conducted on the samples before and after ion-implantation with He+ ions to high fluences (4 × 1016 cm-2) at 50 KeV on. A surprising suppression of amorphization of AlN film was observed in all the nanolayer samples. TEM and nanoindentation study of the ion-irradiated TiN/AlN multilayer films showed superior radiation tolerance in 10 nm and 20 nm nanolayer specimens, whereas obvious hardness reduction were observed in 5 nm and 50 nm nanolayer specimens and 5 nm specimen showed significant nanolayer intermixing. This study indicates that stable interface could substantially reduce radiation-induced damage, and there are nanolayer thickness criteria for stable interface in TiN/AlN multilayer, which may be varied depending on material.
9:00 PM - BB5.18
Evolution of the Nano-structure of French Reactor Pressure Vessel Steels With Neutron Fluence.
Hefei Huang 1 , Radiguet Bertrand 1 , Patrick Todeschini 2 , Guillaume Chas 3 , Philippe Pareige 1
1 , GPM-UMR CNRS 6634, Saint Etienne du Rouvray France, 2 MMC, EdF, Les renardières France, 3 CEIDRE, EdF, Chinon France
Show AbstractIt is well established that vessel steels of pressure water reactors (PWR) harden and embrittle during neutron irradiation, due to the formation of a high number density of nano-scale features: point defect clusters and solute clusters. However, the long term structural evolution of these materials under the specific ageing condition of PWR need some highlights.In this work, different materials from surveillance program of different French nuclear power plants are studied. These different steels (i.e. different nominal compositions) aged under different fluences will give information on the chemical composition and irradiation conditions influence on solute behaviours. The production of irradiation defects is simulated from state of the art numerical tools and models. The solute behaviour under this point defect production condition is characterized using the Atom Probe Tomography technique. The coupling of nanoscale modelling and nanoscale observation allows to give key information concerning the basic mechanisms at the origin of the phase transformation and thus to estimate the long term behaviour of the materials.The aim of this study is to improve knowledge on structural evolution under irradiation of French RPV steels up to high fluence and to support hypotheses for the reviewing of prevision formula.
9:00 PM - BB5.19
Grain Boundary Structure Effects on Damage Accumulation in Reactor Steels.
Zhe Leng 1 , David Field 1
1 , Washington State Univ, Pullman, Washington, United States
Show AbstractFerritic/martentic steels are attractive materials for use as components in nuclear reactors because of their high strength and good swelling resistance. Grain boundary specific phenomena (such as segregation, voiding, cracking, etc) are prevalent in these materials so grain boundary character is of primary importance. Certain types of boundaries are more susceptible to damage whereas others tend to resist radiation damage. If more damage resistant boundaries can be introduced into the structures, this will result in steel that is more resistant to the processes of degradation that prevail in high-temperature, radiation containing environments. We have characterized the grain boundary structure in HT9 and T91 steels by electron backscatter diffraction to identify boundaries that are resistant to degradation and those that are more susceptible to damage during high temperature fatigue. It is found that intergranular damage is mitigated by a high fraction of low energy boundaries due to lower diffusivity and less segregation.
9:00 PM - BB5.2
The Gamma Irradiation Effects on the Reliability of High-k Hafnium Oxide.
Ercan Yilmaz 1 , Aliekber Aktag 1 , Belgin Ergin 1 , Rasit Turan 2
1 Physics, Abant Izzet Baysal University, Bolu Turkey, 2 Physics, Middle East Technical University, Ankara Turkey
Show Abstract Radiation effects on MOS based devices with high-K dielectrics including MOS capacitors are of interest both from space and microelectronic applications point of view. MOS capacitors having a dielectric layer between semiconductor substrate and a metal electrode are sensitive to irradiation such as gamma ray, neutrons, electrons, x rays. Traditionally SiO2 is used as the dielectric layer in MOS based devices. This study focuses on the irradiation effects on the MOS devices with high-K HfO2 as the dielectric layer. In this work we carried out experimental studies on MOS capacitors with HfO2 layer with a thickness of 7, 9, 15 nm. The study includes the effect of gamma radiation with doses up to 16 Gray. The effect of radiation was determined from the valance band shift in the C-V curves upon exposing the device to the radiation. Variations of the device response with different HfO2 thicknesses have been determined and discussed. The midgap and flatband voltage shifts in these devices were analyzed. The amount of charge induced by the ionization radiation has been calculated and the results are compared with the response of MOS capacitors with SiO2 with 250 nm thickness. Results show that the radiation effects on the devices with HfO2 suggest that alternative dielectrics to SiO2 and could be integrated into future electronic technologies for not only many space applications at high doses but also for medical applications particularly at low radiation doses."This work is supported by Abant Izzet Baysal University under contract number AIBU, BAP. 2009, 03.02.319. (Scientific Research Projects)".
9:00 PM - BB5.20
Fracture Toughness of Nanostructured Ferritic and Conventional Ferritic-Martensitic Alloys.
Mikhail Sokolov 1 , David Hoelzer 1
1 , ORNL, Oak Ridge, Tennessee, United States
Show AbstractThe ferritic-martensitic (FM) steels are the primary candidates for a fusion power plant first wall and blanket structure. However, their utilization is limited to temperatures of approximately 600oC due to inferior tensile and creep strength at higher temperatures. Oxide-dispersion strengthening (ODS) is a well known process to improve the strength and creep properties of ferritic steels. Moreover, recent advances in manufacturing and understanding of the strengthening mechanisms have resulted in the creation of a new type of materials, namely, nanostructured (NS) ferritic steels. Significant improvement in strength and creep properties have been achieved of these NS steels compared to the conventional ODS steels. Elevated temperature strength in these steels is obtained by a high number density of ultra-fine, nanometer-scale complex titanium-yttrium-oxygen particles dispersed in a ferritic matrix. However it was anticipated that improvements in strength and creep properties of these NS steels should come in price of low fracture toughness. In this study, fracture toughness of the NS alloys (12YWT and 14YWT) is compared to fracture toughness of conventional FM steels like F82H and EUROFER and conventional ODS steels like EUROFER ODS and MA967. Fracture toughness characterization was performed in the transition region as well as in the ductile region at elevated temperatures (300C and above). It is shown that 14YWT alloy exhibited relatively low transition temperature and decent ductile toughness in LT orientation. However, fracture toughness of this alloy suffered in TL orientation for both, transition and ductile regions. The limited data of fracture toughness of these NS alloys after irradiation in HFIR reactor at 300C will be compared to shifts of fracture toughness of conventional FM steels like F82H and EUROFER at comparable doses.
9:00 PM - BB5.21
Effects of Structure and Grain Size on the Response of Titania to Ion Beam Irradiation.
Jiaming Zhang 1 , Fengyuan Lu 2 , Jie Lian 2 , Fereydoon Namavar 3 , Jianwei Wang 1 , Rodney Ewing 1
1 , University of Michigan, Ann Arbor, Michigan, United States, 2 , 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. In the present study, two different nanocrystalline phases of Titanium dioxides, anatase and rutile, have been synthesized in different ways to study the effects of structure and grain size on the radiation response of the titania. Nanocrystalline rutile, with grain size ~5 nm in dense thin-film, was synthesized on Si substrate by ion beam assisted deposition (IBAD). A crystal-to-amorphous transformation was observed 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. However, nanocrystalline anatase, which is synthesized by mechanochemical methods, shows enhanced radiation resistance in the grain size ~20 nm. However, this enhanced tolerance was mitigated when the crystal sizes of anatase phase are ~10 nm and ~50 nm. The dependence of radiation resistance on the temperature will also be discussed.
9:00 PM - BB5.22
Study of GPC as Potential Material for Heat Exchanger for Nuclear Reactor.
Bopha Chhay 1 , Daryush Ila 1
1 , Alabama A&M University, Normal, Alabama, United States
Show AbstractGlassy Polymeric Carbon (GPC) is a unique form of carbon that can withstand high temperature and corrosive environments, with applications in diverse areas including crystal growth, space industry, and biomedical research. In addition to very high chemical and thermal stability, GPC distinguishing properties include good mechanical performance (no significant mechanical deformation when tested up to 3000 °C).In this research we fabricated a heat exchanger for nuclear reactor made of GPC material. We studied the mechanical properties of GPC with an ingenious instrument designed at AAMU and the chemical properties with Raman analysis and XPS.
9:00 PM - BB5.24
Characterization of Changes in Properties and Microstructure of Glassy Polymeric Carbon Following Ag Ion Irradiation.
Malek Abunaemeh 1 , Mohammad Seif 2 , Abdulla Elsamadicy 3 , Ibidapo Ojo 1 , Young Yang 4 , Claudiu Muntele 1 , Daryush Ila 1
1 , Alabama A&M University, Madison, Alabama, United States, 2 Mechanical Engineering Department, Alabama A&M University, Normal, Alabama, United States, 3 Physics Department, University of Alabama in Huntsville, Huntsville, Alabama, United States, 4 Engineering of Physics Department., University of Wisconsin, Madison, Wisconsin, 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 fluence of 5 o MeV Ag equivalent to a 1displacment per atom (DPA) at samples prepared at 1000, 1500 and 2000οC. For surface analysis we are using scanning electron microscopy, nano-indentation, X-ray photoelectron spectroscopy, Raman spectroscopy and transmission electron spectroscopy (TEM). The GPC material is manufactured and tested at the Center for Irradiation Materials (CIM) at Alabama A&M University.
9:00 PM - BB5.4
He-implanted Nuclear Ceramics: Microstructural Modifications Upon Thermal Annealing.
Gihan Velisa 1 , Aurelien Debelle 2 , Laetitia Vincent 2 , Lionel Thome 2 , Alain Declemy 3 , Dan Pantelica 1
1 Department of Nuclear Physics-Tandem, Horia Hulubei National Institute for Physics and Nuclear Engineering , Magurele, Ilfov, Romania, 2 , Centre de Spectrométrie Nucléaire et de Spectrométrie de Masse, Univ. Paris-Sud 11, Orsay France, 3 , Laboratoire de Physique des Matériaux, CNRS-Université de Poitiers, Poitiers France
Show AbstractNuclear ceramics are refractory solids which possess interesting physico-chemical properties such as high chemical stability and strong resistance against irradiation. For specific applications such as the safe and long term disposal of radioactive waste and the development of inert fuel matrices for actinide transmutation, it is of prime importance to study in this type of materials both (i) the production and recovery of radiation damage and (ii) the behavior of foreign species introduced in the matrix due to the fission reactions or to the presence of radioactive elements. In this framework, the effect on the ceramic matrices of the alpha decay of actinides, leading to the production of He gas and alpha recoil nuclei, appears as a mandatory issue to study. This paper presents results on this topic. The microstructural modifications in crystalline yttria-stabilized zirconia (YSZ) and magnesium aluminate spinel (MAS) implanted with He ions and subsequently annealed at 800°C were examined. The combination of three complementary analysis techniques, namely Rutherford backscattering spectrometry in channeling geometry, X-ray diffraction and transmission electron microscopy, allowed to get a good insight of the behaviour of these matrices in the presence of a large He concentration. The damage and elastic strain depth distributions were determined and compared to SRIM-predicted He and dpa profiles, which allowed determining the nature of radiation defects in the as-implanted crystals. After annealing, a partial strain relaxation was measured in YSZ whereas a complete one was measured in MAS. Furthermore, the occurrence of huge microstructural modifications was observed. Actually, TEM micrographs show that He bubbles are formed in both materials. However, spherical He bubbles developed in YSZ while He platelets formed in MAS. In addition, large elongated cracks were visible in the YSZ crystals. The difference in the behaviour of both materials upon annealing are discussed in the light of He transport and clustering property differences.
9:00 PM - BB5.6
The Effects of Gamma Radiation on Giant Magneto Resistance (GMR) Devices.
Sarah Parks 1 , Turhan Carroll 1 , Adam Hauser 1 , Jeremy Lucy 1 , Cole Robinette 1 , James Mathis 3 , Joseph Talnagi 2 , Denis Pelekhov 1 , Thomas Blue 2 , P. Hammel 1 , Fengyuan Yang 1 , Ezekiel Johnston-Halperin 1
1 Physics, The Ohio State University, Columbus, Ohio, United States, 3 Physics, Norfolk State University, Norfolk, Virginia, United States, 2 Mechanical Engineering, The Ohio State University, Columbis, Ohio, United States
Show AbstractCurrent information technology relies heavily on magnetic memory in the form of giant magnetoresistive (GMR) read heads coupled with magnetic recording media and the more recent implementation of magnetic random access memory (MRAM). Among the many potential benefits of this paradigm is the presumption that these materials are radiation hard with respect to both photons and particles, potentially increasing their utility for nuclear energy and space based applications. However, to date there are few detailed studies of magnetism and GMR devices in high radiation environments. This project utilizes the facilities at The Ohio State University Nuclear Reactor Laboratory to study the effects of Co-60 gamma radiation on GMR samples. The structure used in this experiment is a standard GMR trilayer consisting of a thin, non-magnetic layer placed between thin ferromagnetic layers, with one of the two magnetic layers in contact with an exchange bias antiferromagnet (Py/Cu/Py/FeMn/Ge). To study the effects of radiation three types of magnetic measurements, vibrating sample magnetometery (VSM), magnetoresistance (MR), and magneto-optical Kerr effect (MOKE), are taken and correlated pre and post gamma radiation. We present characterization of the devices pre and post gamma irradiation for multiple device geometries and radiation doses up to 50Mrad.
9:00 PM - BB5.8
Surface Deformation Nanostructures and Stress Corrosion Crack Precursors Originating from Surface Grinding.
Matthew Olszta 1 , Larry Thomas 1 , Stephen Bruemmer 1
1 , Pacific Northwest National Lab, Richland, Washington, United States
Show AbstractStress corrosion cracking (SCC) continues to be a dominant materials degradation mechanism in light-water reactor (LWR) components. While considerable understanding has been gained concerning SCC through crack growth measurements and crack characterizations, limited work has been performed on processes controlling crack initiation in service. Surface and near-surface microstructures and microchemistries have been investigated in a variety of cracked stainless steel components after long-term LWR exposure. A common feature in most examples was the presence of surface damage, typically grinding, before high-temperature water exposure. Analytical electron microscopy examinations reveal localized oxidation through a fine-grained, surface layer and into a heavily deformed subsurface region. In order to better precursor damage structures, controlled polishing and grinding has been performed on representative stainless steels and characterized by focused ion beam, scanning electron microscopy, electron backscatter diffraction and transmission electron microscopy. Surface recrystallization was found to occur in lightly ground to heavily abraded materials producing nanocrystalline grains within the first ~0.5-10 μm along with high dislocation densities, twinning and lath structures. These near-surface damage characteristics will be linked to the service sample results and discussed with respect to SCC initiation precursors in LWR components.
9:00 PM - BB5.9
He Ion Irradiation Damage in Ag-V Multilayer Nanocomposites.
Qiangmin Wei 1 , Yongqiang Wang 1 , Amit Misra 1 , Michael Nastasi 1
1 , Los Alamos National Lab, Los Alamos, New Mexico, United States
Show AbstractMicrostructure and hardness of He implanted polycrystalline Ag-V multilayers with different individual thickness ranging from 1nm to 50 nm were investigated. Irradiation experiments were performed using 33 keV 3He+ at room temperature up to a dose of ten to the seventeenth per square centimeter. It was found that bubbles generated in the Ag layer are larger than those in the V layer. At 1 nm layer, bubbles are confined in the Ag layer by the V-Ag interfaces, leading to the elongated bubble formation. The minimum He concentration at which bubbles are detected in TEM increases from 2~3 at.% at 5 nm to 4~5 at.% at 1 nm layer thickness, up to an order of magnitude higher than bulk materials. More importantly, radiation hardening decreases with decreasing layer thickness and radiation softening was observed at thickness of 1 nm. The He solubility and inverse hardening were discussed on the basis of microstructure of ion-irradiated multilayers.
Symposium Organizers
Ram Devanathan Pacific Northwest National Laboratory
Amit Misra Los Alamos National Laboratory
Christina Trautmann GSI Helmholtzzentrum
Brian Wirth University of California-Berkeley
BB6: Theory, Modeling and Simulation
Session Chairs
Kai Nordlund
Michael Tonks
Thursday AM, April 08, 2010
Room 3012 (Moscone West)
9:30 AM - **BB6.1
Electronic Effects in Radiation Damage Simulations of Metals and Insulators.
Dorothy Duffy 1 2 , Sascha Khakshouri 1 , Jack Mulroue 1 , Szymon Daraszewicz 1
1 Physics and Astronomy, University College London, London United Kingdom, 2 EURATOM/UKAEA Fusion association, Culham Science Centre, Culham, Oxfordshire, United Kingdom
Show AbstractCascade simulations, where one atom of a material is given a high velocity to represent a radiation event, have made significant contributions to the understanding of radiation damage in a wide range of materials. The energy deposited by the radiation event creates a localised region of dynamic disorder in the lattice, which recrystallises as the energy dissipates via thermal diffusion through the lattice. The residual damage (point defects and defect clusters) remaining after recrystallization evolves over time and results in changes to the macrpscopic material properties, such as swelling and embrittlement.Classical cascade simulations neglect the effects of the interchange of energy between the atomic nuclei and the electrons. These effects include the inelastic energy losses due to electronic stopping, energy storage and transport by the electrons and redistribution of electronic energy to the lattice. We have developed a method for including these effects in classical molecular dynamics simulations of metallic materials [1]. We couple the atomistic simulation to a coarse-grained model for the electronic energy, which evolves via the heat diffusion equation. Energy is exchanged between the atomistic simulation and the coarse-grained model at each time step.The model has been used to model low energy radiation events, swift heavy ion irradiation and sputtering events. For low energy events the energy storage was found to enhance defect annealing, caused by slower cooling of the molten region. For very high energy radiation, such as swift heavy ion irradiation, energy is deposited primarily in the electronic system. This energy is transferred to the electrons via electron-phonon coupling and, above a threshold stopping power, a cylindrical molten region is formed in the crystal, which results in an elongated track of defects. Results for cascade [2], swift heavy ion [3] and sputtering simulations [4] will be presented. Radiation damage insulating materials is a major concern for the nuclear industry, particularly in the fields of nuclear fuel and radioactive waste disposal. Work is in progress to extend the methodology developed for metals to insulating materials. In this case materials some of the energy deposited by a radiation event may excite electrons to create free carriers (electrons, holes, excitons). These carriers diffuse through the lattice and eventually decay. The decay can be radiative, in which case the energy is lost to the system, or non-radiative, when the energy will be deposited in the lattice. A methodology for including this effect will be described and the results of preliminary simulations will be presented.1. D.M.Duffy and A.M. Rutherford, J. Phys: Cond. Matt. 19, 016207 (2007) 2. A.M. Rutherford and D.M. Duffy, J. Phys: Cond. Matt. 19, 496201 (2007)3. D.M Duffy, N Itoh, A.M. Rutherford and A.M. Stoneham, J.Phys. Cond Matt. 20, 082201 (2008)4. S. Khakshouri and D.M. Duffy, Phys. Rev. B 80, 035415 (2009)
10:00 AM - BB6.2
Kinetics Versus Thermodynamics in Materials Modelling: The Case of the Di-vacancy in Iron.
Flyura Djurabekova 1 , Lorenzo Malerba 2 , Roberto Pasianot 3 4 , Par Olsson 5 , Kai Nordlund 1
1 , Helsinki Institute of Physics, Helsinki Finland, 2 , SCK-CEN, Mol Belgium, 3 , Depto. Materiales, CAC-CNEA, Buenos Aires Argentina, 4 , CONICET, Buenos Aires Argentina, 5 , Dept. MMC, EDF-R&D, Moret-sur-Loing France
Show AbstractThe in-service and off-normal behaviour of structural materials is largely determined by the kinetics of thermally activated atomic-scale processes. Monte Carlo models are widely used for the study of microstructural and microchemical evolution of these materials under irradiation. However, they often link explicitly the relevant activationenergies to the energy difference between local equilibrium states. By combination of Monte Carlo and Molecular Dynamics techniques we show that the explicit link between two energies may lead to the misinterpreting of the actual path the object chooses for its motion. We provide a simple example (di-vacancy migration in iron) in which we use a rigorous activation energy calculation by meansof both empirical interatomic potentials (drag method by applying MD simulations) and density functional theory methods (nudged elastic band method, implemented in VASP). Both cases clearly show that such a link is not granted, revealing a migration mechanism that a thermodynamics-linked activation energy model cannot predict. Such a mechanism is, however, fully consistentwith thermodynamics. This example emphasises the importance of basing Monte Carlo methods on models where the activation energies are rigorously calculated, rather than deduced from widespread heuristic equations.
10:15 AM - BB6.3
Palladium in Cubic Silicon Carbide: Stability and Kinetics.
Guido Roma 1
1 DEN/DMN/SRMP, CEA, Gif sur Yvette France
Show AbstractSeveral technological applications of silicon carbide are concerned with the introduction of palladium impurities. Be it intentional or not, this may lead to the formation of silicides. Not only this process is not well understood, butthe basic properties of palladium impurities in silicon carbide, like solubility or diffusion mechanisms, are far from being known. Here the stability and kinetics of isolated Pd impurities in cubic silicon carbide are studied by first principles calculations in the framework of density functional theory. Pd turns out to be preferentially substitutional in cubic SiC, but diffusing mainly in interstitial form. This gives rise to a variety of intrinsic defect-assisted mechanisms some of which have not so high migration barriers. We will discuss them as well as their role in the diffusion of palladium in SiC in various conditions, in particular under irradiation. This finding is relevant for the application of silicon carbide as a cladding material for gas-cooled fission reactors, where Pd, as a fission product, can be detrimental to tightness of the cladding. The early stages of nucleation will also be adressed on thebasis of the results obtained for isolated impurities and the smallest clusters.
10:30 AM - BB6.4
Phase-field Modeling of Void Migration and Swelling in Materials Under Irradiation and a Temperature Gradient.
Yulan Li 1 , Shenyang Hu 1 , Xin Sun 1 , Fei Gao 1 , Charles Henager 1 , Moe Khaleel 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractVacancy and interstitial are two major defects produced by fission fragment and neutron irradiations in nuclear reactor components such as nuclear fuels and cladding materials. With the accumulation of these defects and their interactions with microstructures, void nucleation, growth, and volume swelling take place. At the presence of temperature gradient, voids might migrate. Experiments show that the migration of voids in nuclear fuels results in the formation of a central hole and interconnected tunnels on grain boundaries. In cladding materials, the formation and evolution of void lattices are often observed in body-centered cubic (BCC), face-centered cubic (FCC) and hexagonal-close-packed (HCP) structural materials. These microstructure changes result in changes of thermo-mechanical properties such as thermal conductivity, elasticity, plasticity as well as structure instability including volume swelling and cracking. Therefore, a fundamental understanding of the effects of temperature gradient and irradiation conditions on void evolution is crucial for the scientific design of nuclear reactor materials, optimizing plant operation, and reducing uncertainty in operational and safety margins. We employed the phase-field approach to investigate the migration of vacancies, interstitials, and voids as well as volume swelling in materials under irradiation and a temperature gradient. A phase-field model has been developed to take into account the generation rates of irradiation-induced vacancies and interstitials, their recombination between vacancies and interstitials, and the different mobilities of vacancies along void surface and in bulk. The dependence of single void migration velocity on surface diffusion and recombination rates, the volume swelling on interstitial mobilities and the generation rate of both vacancies and interstitials will be presented. Comparison with experimental observations will be discussed.
10:45 AM - BB6.5
Multiscale Modeling of Irradiation Induced Hardening in a-Fe, Fe-Cr and Fe-Ni Systems.
Ioannis Mastorakos 1 , Ngoc Le 1 , Hussein Zbib 1 , Mohammad Khaleel 2
1 , Washington State University, Pullman, Washington, United States, 2 , Pacific Northwest National Laboratory, Richmond, Washington, United States
Show AbstractThe development of the new Generation IV reactors depends on ensuring structural materials integrity.Structural materials in the new generation of reactors will operate in harsh radiation conditions coupled with high levels of hydrogen and helium production,thus experiencing severe degradation of mechanical properties.The development of structural materials for use in such a hostile environment is predicated on understanding the underlying physical mechanisms responsible for microstructural evolution along with corresponding dimensional instabilities and mechanical property changes. The purpose of this work is to study the behavior of a-Fe, Fe-Cr and Fe-Ni systems under irradiation using both Molecular Dynamics(MD)and Dislocation Dynamics(DD)simulations. Critical information is passing from the atomistic (MD)to the microscopic scale(DD)in order to study the degradation of the material under examination. In particular, information pertaining the dislocation-defects(such as voids, helium bubbles and prismatic loops)interactions is obtained from MD simulations. Then this information is used by DD to simulate large systems with high dislocation,and defect densities. The results reveal the multiscale nature of the degradation phenomena in extreme environment.
11:30 AM - **BB6.6
Atomistic Simulation of Steels: Radiation Effects in Fe, FeCr, FeC and FeCrC.
Kai Nordlund 1 , Carolina Bjoerkas 1 , Katharina Voertler 1 , Niklas Juslin 1 , Krister Henriksson 2
1 Department of Physics, University of Helsinki, Helsinki Finland, 2 Department of Chemistry, University of Helsinki, Helsinki Finland
Show AbstractKnowledge of the nature of radiation damage in steels is ofgreat importance for extreme nuclear applications environments, such as Gen IV fission and fusion reactors. The steels are subject to a continuous flux of neutrons, producing damage in the materials which eventually leads to both embrittlement and swelling.The theoretical study of damage in steels focused for a long time on pure Fe, under the assumption that the damage in a real steel is similar to that in pure Fe. However, experiments have shown that at least the presence of Cr has a profound effect on the damage in steels. It is only during the last about 5 years that the studies have been extended from pure Fe to the FeCr alloy, when suitable interatomic potentials have become available. The studies of FeCr have indicated that the primary damage stateis essentially identical in FeCr as in Fe, but that Cr affects dislocation mobility strongly. However, real steels always contain also carbon as an alloying element. Even though Cr does not affect the primary state of damage, C might since it is a much lighter element. Using very recently developed bond-order interatomic potentials for the FeCrC system, we simulate the primary state of radiation damage in FeCrC, at compositions corresponding to stainless steel, and compare it with damage in Fe, FeCr and FeC at similar irradiation energies.
12:00 PM - BB6.7
Atomic Scale Clustering of Fission Products at Dislocations in Uranium Dioxide.
David Parfitt 1 , Clare Bishop 1 , Mark Wenman 1 , Robin Grimes 1
1 Materials, Imperial College London, London United Kingdom
Show AbstractThe structure and movement of dislocations in uranium dioxide is of significance for both the bulk properties of nuclear fuel and the mobility of defects on an atomic scale. We examine here atomistic models of typical edge and screw dislocations with the aim of quantifying the influence of dislocations upon the distribution of fission products within a fuel grain. We show that the dislocation cores and associated strain fields may provide energetically favourable trapping sites for a variety of fission products. We also consider the effect these defects may have in pinning the dislocations themselves.
12:15 PM - BB6.8
First Passage Monte Carlo Simulation of Intragranular Fission Gas Bubble Population Evolution in UO2 Fuels.
Daniel Schwen 1 , Robert Averback 1
1 Materials Science and Engineering, University of Illinois at Urbana-Champaign, Urbana, Illinois, United States
Show AbstractUnderstanding and possibly control of the evolution of an intragranular fission gas bubble population is a key challenge in the development of the high burn-up generation IV reactor technology.Distribution and size of fission gas bubbles in uranium dioxide and mixed oxide fuel elements determine the gas transport properties and the fuel performance. Fission gas retention can adversely influence the thermal and mechanical properties of reactor fuels, while release of gaseous species from the fuel can lead to cladding failure at high burn-up. We developed a code to track the diffusive motion of individual atoms and gas bubbles over long timescales. The core idea, presented in a paper by Opplestrup et al. [1], is the partitioning of the sample domain into zones containing single bubbles or fission gas atoms and analytically calculating the diffusion times to the zone edges, thereby replacing many single diffusive jumps by a few super-jumps.In contrast to mean field theories spatial correlations arising from re-solution distances and proximity to grain boundaries (sinks) or free surfaces, as well as local fluctuations, can be readily incorporated. A variety of gas re-solution scenarios (homogeneous and heterogeneous re-solution) and diffusion models are explored. Of particular interest is the existence of patterning regimes and their stability with respect to the various input parameters.[1] T. Opplestrup, V. Bulatov, G. Gilmer, M. Kalos, B. Sadigh, Phys. Rev. Lett. 97 (2006) 10-13.
12:30 PM - BB6.9
Atomistic Simulation of Thermal Spikes in Zircon.
Ram Devanathan 1 , Pedro Moreira 2 , William Weber 1
1 Chemical & Materials Sciences Division, Pacific Northwest National Laboratory, Richland, Washington, United States, 2 Instituto de Fisica “Gleb Wataghin”, Universidade Estadual de Campinas, Campinas, SP, Brazil
Show AbstractZircon (ZrSiO4) has been proposed as a host material for the disposition of nuclear waste and the immobilization of excess weapons plutonium. Fission tracks in zircon are of interest from the standpoint of determining the effect of radiation on the structural and mechanical integrity of ceramic waste forms. Fission tracks are also used in radiometric dating to understand the thermal history of minerals. We have used classical molecular dynamics simulations with a recently developed empirical potential to model fission tracks as cylindrical thermal spikes. The results show that there is a critical electronic stopping (dE/dx) value below which there is no track formation. Above this value, the radius of the track increases with deposited energy. The tracks show density and composition variation from center to the periphery. These results will be contrasted with the response of UO2 to fission track damage and conclusions will be drawn about radiation tolerance of ceramics.
12:45 PM - BB6.10
Predicted Structure and Stability of Radiation Tolerant A4B3O12 δ-phase Compositions.
Christopher Stanek 1 , Chao Jiang 1 , Blas Uberuaga 1 , Kurt Sickafus 1 , Robin Grimes 2
1 Material Science and Technology, Los Alamos National Laboratory, Los Alamos, New Mexico, United States, 2 Department of Materials, Imperial College London, London United Kingdom
Show AbstractRecently, A4B3O12 δ-phase compounds have been shown to be tolerant to amorphization under irradiation. In this talk, we discuss the results of atomistic simulations employed to predict ordered structures for a series of A4B3O12 δ-phase compounds, where A is a 3+ cation ranging in size from Sc3+ to Ho3+ and B is a 4+ cation ranging from Ti4+ to Zr4+. Experimentally, a fully-ordered cation structure has yet to be resolved for any of these compounds. Monte Carlo energy minimization calculations using short-range pair potentials identifed three low energy arrangements of A3+ and B4+ cations. The details of these three structures were analyzed with the layer motif method. To quantitatively determine the δ-phase structure of each composition, the three configurations were re-evaluated with density functional theory. We also used special quasirandom structures to compare the ordered low energy configurations to cation disorder. For all compositions considered, we find that at least one of the three ordered structures is lower in energy than the disordered structure, suggesting the thermodynamic stability of an ordered phase. Of the three ordered structures identified by this approach, one has not been identified previously in the literature for any composition. In addition, we discuss the stability of δ-phase compounds with respect to other ``ABO4-x" fluorite derivative compositions and predict the structure of compositions for which none has been reported.
BB7/AA8: Joint Session: Radiation Effects in Wasteforms and Ceramics
Session Chairs
Ram Devanathan
Karl Whittle
Thursday PM, April 08, 2010
Room 3010 (Moscone West)
2:30 PM - **BB7.1/AA8.1
Radiation Effects in Actinide-containing Ceramics for the Advanced Fuel Cycle.
William Weber 1 , Ram Devanathan 1
1 Fundamental & Computational Sciences Directorate, Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractAdvanced nuclear fuel cycles may include a separate waste stream for minor actinides that may be immobilized directly in waste forms or in inert matrices for in-reactor transmutation and subsequent geologic disposition. Since some minor actinides have relatively high spontaneous fission rates or neutron fission cross sections, radiation effects from both alpha decay and fission may need to be considered. Self-radiation from alpha decay and fission of actinides in crystalline ceramic host phases generally leads to track formation, point defect accumulation, and crystalline phase transformations or amorphization, which may be accompanied by macroscopic swelling and increases in dissolution rates. The results of computer simulations, new models, and experimental studies in several relevant materials using short-lived actinides and ion-beam irradiation methods, along with comparisons to natural mineral data, will be presented to highlight the fundamental understanding and models of radiation effects in actinide-containing ceramics developed over the past 30 years. Low-energy (~MeV) heavy-ion irradiation techniques have been demonstrated to be very effective in simulating radiation effects from alpha decay over a large range of experimental conditions in order to develop more detailed scientific understanding and predictive models. Swift-heavy ion irradiations (~0.1 to 2 Gev) provide a means to better understand the nature of fission tracks. The integration of computer simulations into these studies have advanced the interpretation of experimental results and led to the development of a comprehensive atomic-level understanding of radiation damage processes and predictive models of the complex evolution of radiation damage in actinide-containing waste forms or inert matrix fuels with time and temperature.
3:00 PM - BB7.2/AA8.2
Quantification of α-type Radiation Damage in Zircon.
Katie Gunderson 1 , Clive Brigden 1 , Eric Vance 2 , John Hanna 2 , Ian Farnan 1
1 Department of Earth Sciences, University of Cambridge, Cambridge United Kingdom, 2 , Australian Nuclear Science and Technology Organisation, Menai, New South Wales, Australia
Show AbstractQuantification of α-radiation damage in materials is important because of the large number of heavy α-emitters in nuclear waste and nuclear materials. Each α-decay event causes radiation damage by two processes: the emission of the α-particle itself, and the recoil of the heavy nucleus. The majority of localized structural damage leading to amorphization is caused by the recoil of the heavy actinide nucleus (70-100 keV), such as that of plutonium, americium or curium. The higher energy α-particle (4.5-5.5 MeV) mainly causes ionizations, but is also known to cause some atomic displacements as it is slowed by collision with atomic nuclei. Nuclear magnetic resonance (NMR) has previously been used to quantify the cumulative amount of radiation damage in material due to the total damage event (heavy recoil + α). This study attempts to quantify the damage due to α-particles alone and provides precise constraints on the number of damage events through 7Li spin counting. Here we present results from zircon synthesized with natural boron as B2O3 inclusions. The natural abundance of 10B is 19.9%, and it undergoes the reaction 10B + n → 7Li + α. Samples were neutron-irradiated at HIFAR reactor at Australian Nuclear Science and Technology Organisation for one day, one week, and one month at a neutron fluence of 1013 cm2/s. The amorphous fraction of the sample is quantified using 29Si NMR by comparing the area of a narrow line corresponding to undamaged zircon with the area of a broad resonance at more negative chemical shifts corresponding to amorphized material. A calibration curve is constructed for 7Li NMR signal as a function of the number of 7Li atoms, and is used to quantify the number of 7Li atoms in each irradiated sample. Complete fission occurred in the sample that had been irradiated for one month, and analysis of the 29Si NMR spectrum indicates that 30% of the structure had been damaged. Each 10B + n → 7Li + α reaction creates two energetic light particles that we consider together. The amount of 7Li detected in the 1 month sample is 4.08 x 1018 atoms, which translates to 1.95 x 1019 α events / g material. In the same sample the 29Si NMR spectrum shows a damage level of 5.90 x 1021 atoms/g. This leads to a damage rate of 302 atoms/α. Subsequent experiments on lower levels of irradiation should provide better constraints on the damage per α. Also, X-ray diffraction will allow changes in unit cell parameters to be quantified as a function of ‘α-particle’ only damage.
3:15 PM - BB7.3/AA8.3
Ion-irradiation-induced Damage Evolution in Titanate Pyrochlores.
Yanwen Zhang 1 , William Weber 1
1 , Pacific Northwest National Laboratory, Richland, Washington, United States
Show AbstractPyrochlore materials, due to the remarkable elemental versatility in the A2B2O7 crystal structure, are considered for a wide range of applications, such as fuel cells, catalysts, inert fuel matrices, and waste forms for the immobilization of actinides. Considerable self-radiation damage from alpha-decay in actinide-bearing phases can result in amorphization, macroscopic swelling and significant increases in dissolution rates, and these changes in structure and chemical durability affect long-term performance of the actinide waste forms. Irradiation effects in Ho2Ti2O7 and Sm2Ti2O7 single crystals were studied using 1.0 MeV Au+ ion at room temperature, which provides a reasonable simulation of the damage evolution behavior due to alpha recoils. Damage evolution as a function of irradiation dose was investigated by Rutherford backscattering spectroscopy along the <001> direction. A disorder accumulation model, with contributions from the amorphous fraction and the crystalline disorder, has been fit to the damage accumulation data. The damage evolution behavior indicates that the relative disorder on the Ho sublattice follows a nonlinear dependence on dose and that defect-stimulated amorphization is the primary amorphization mechanism, which is similar to the irradiation behavior previously observed in Sm2Ti2O7. A slower observed damage accumulation rate for Ho2Ti2O7, as compared with damage evolution in Sm2Ti2O7, is mainly attributed to a lower effective cross section for defect-simulated amorphization. The critical dose for amorphization under 1.0 MeV Au+ irradiation in Ho2Ti2O7 and Sm2Ti2O7 crystals are in good agreement with TEM results for polycrystalline Sm2Ti2O7 and Gd2Ti2O7 under 0.6 MeV Bi+ irradiation and for 244Cm-doped Gd2Ti2O7, indicating that amorphization due to heavy-ion energy deposition is relatively independent of dose rate and ion mass near room temperature. During thermal annealing in an 18O environment, the increase in exchange near the critical temperature for amorphization between 16O in the Sm2Ti2O7 sample and 18O suggests a correlation between oxygen vacancy mobility and the critical temperature.
3:30 PM - BB7.4/AA8.4
Experimental Insight into the Radiation Resistance of Zirconia-based Americium Ceramics.
Renaud Belin 1 , Philippe Martin 1 , Philippe Valenza 1 , Andreas Scheinost 2
1 DEN - DEC - SPUA - LMPC, CEA Cadarache, Saint Paul Lez Durance, 14999, France, 2 Institute of Radiochemistry, Forschungszentrum Dresden-Rossendorf (FZD), Dresden Germany
Show AbstractCeramics intended for use as nuclear fuels, transmutation targets or actinide immobilization matrices have to endure severe conditions including internal radiation. Resistance to amorphization and limited swelling are therefore key requirements. While zirconia-based materials with defect-fluorite structure have shown high tolerance against external ion-beam irradiation, few experimental studies have demonstrated that these structures also resist under more realistic conditions, i.e. with homogeneous internal radiation from α-emitting actinides within the structure. Here, we provide for the first time experimental insight into the radiation-resistance mechanisms of americium pyrochlore 241Am2Zr2O7. The presence of 62 wt% of 241Am (t1/2=432 y) in the structure provided a high dose rate which enabled us to follow within only four years an α self-irradiation-induced aging process corresponding to much longer time periods in the case of immobilization ceramics [1]. X-ray diffraction (XRD) and X-ray absorption spectroscopy (XAS) analyses were used to selectively probe the long-range and the short-range order and the local environment of both cations [2,3].The phase transition from the pyrochlore to the defect-fluorite structure was accompanied with an unusual negative lattice expansion. Once the fluorite structure was reached, neither volume changes nor amorphization were observed over a time course of 4 years. The cumulated dose during this period was 9.4×1018 α-decay events.g-1; corresponding to 0.80 dpa. This is equivalent to the dose accumulated by multiphase ceramics with 20wt% 239Pu during 1,000 years of storage [1]. The defect-fluorite structure is therefore remarkably resistant to such radiation doses.Combining XRD Rietveld refinement and XAS analysis allowed a thorough understanding of how lattice point defects are formed and recombined during the order-disorder transition. The disorder relaxation proceeds through the simultaneous formation of cation antisites and oxygen Frenkel pairs, in line with former molecular dynamics studies. Moreover, EXAFS analysis revealed a disruption in the long-range order and a markedly different development in the local environments of zirconium and americium: while Am-O polyhedra show an increasing disorder with increasing exposure, the Zr-O polyhedral units remain highly ordered. However, they rotate along edges and corners, thereby reducing the structural strain imposed by the growing disorder around americium. We believe it is this particular property of the compound that provides the remarkable resistance to radiation, making it attractive as a durable actinide immobilization host and as a stable transmutation target.[1] Weber W. et al. 1998 J. Mater. Res. 13 1434-1484[2] Martin P. et al. 2007 J. Alloy. Compd. 444-445 410-414.[3] Belin R.C. et al. 2009 Inorg. Chem. 48(12) 5376-5381
3:45 PM - BB7.5/AA8.5
Ion Beam Irradiation of Lanthanum Compounds in the Systems La2O3-Al2O3 and La2O3-TiO2.
Karl Whittle 1 , Gregory Lumpkin 1 , Mark Blackford 1 , Katherine Smith 1 , Nestor Zaluzec 2
1 Materials Engineering, ANSTO, Menai, New South Wales, Australia, 2 Electron Microscopy Center, Argonne National Laboratory, Chicago, Illinois, United States
Show AbstractThin crystals of La2O3, LaAlO3, La2/3TiO3, La2TiO5, and La2Ti2O7 have been irradiated in situ using 1 MeV Kr2+ ions in the Intermediate Voltage Electron Microscope-Tandem User Facility (IVEM-Tandem), at the Argonne National Laboratory (ANL). We observed that La2O3 remained crystalline to a fluence greater than 3.1 × 1016 ions cm-2 at a temperature of 50 K. The four binary oxide compounds in the two systems were observed through the crystalline-amorphous transition as a function of ion fluence and temperature. Results from the ion irradiations give critical temperatures for amorphisation (Tc) of 647 K for LaAlO3, 840 K for La2Ti2O7, 865 K for La2/3TiO3, and 1027 K for La2TiO5. The Tc values observed in this study, together with previous data for Al2O3 and TiO2, are discussed with reference to the phase diagrams for the La2O3-Al2O3 and La2O3-TiO2 systems and the different local environments within the four crystal structures. Results suggest an observable inverse correlation between Tc and melting temperature (Tm) in the two systems. More complex relationships exist between Tc and crystal structure, with the stoichiometric perovskite LaAlO3 being the most resistant to amorphization
4:30 PM - **BB7.6/AA8.6
Response of Complex Oxides to Extreme Conditions of Irradiation and Pressure.
Maik Lang 1
1 Geological Sciences, University of Michigan, Ann Arbor, Michigan, United States
Show AbstractComplex oxides such as A2B2O7 and A2BO5 exhibit a variety of properties that find application in a number of different technologies, from electrolytes in solid oxide fuel cells to actinide-bearing phases. Ion beam irradiations (energy: GeV) have been used to systematically study the effects of pure electronic excitation and ionization in Gd2Zr2-xTixO7 and Gd2TiO5. Synchrotron X-ray diffraction, Raman spectroscopy, and transmission electron microscopy (TEM) revealed radiation-induced phase transitions that include the crystalline-to-amorphous and the order-disorder structural transformations, as well as the formation of high-temperature and high-pressure phases. The extreme conditions prevailing in a single ion track trigger the formation of equilibrium- and non-equilibrium phases, and entire phase diagrams are accessible at the nanoscale. The specific type of material modification can be manipulated by changing the target composition (e.g., Ti-content), beam parameters (e.g., energy loss dE/dx), and external irradiation condition (e.g., pressure and temperature). As an important application, we show that the combination of relativistic heavy ions and high pressure can result in the formation and stabilization of a new metastable Gd2Zr2O7 phase that cannot be obtained by irradiation or pressure applied separately. TEM and quantum-mechanical calculations suggest that these novel structural modifications are caused by the formation of nanocrystals that change the energetics of the phase transformations. This result highlights the importance of the combined use of high pressure and high-energy ion irradiation as a new means for manipulating and stabilizing novel materials to ambient conditions that otherwise could not be recovered.
5:00 PM - BB7.7/AA8.7
Comprehensive Description of the Ion Irradiation-induced Microstructural Modifications in Nuclear Materials: Fluorite-structured Oxides as a Test Case Study.
Aurelien Debelle 1 , Frederico Garrido 1 , Lionel Thome 1
1 CSNSM - CNRS, Université Paris-Sud 11, Orsay Cedex France
Show AbstractMaterials for nuclear applications are inherently submitted to harsh radiative environments. In particular, they must face irradiation with various types of ions in a broad energy range, from low-energy heavy recoil nuclei arising from the alpha decay of actinides, to swift fission fragments. In the majority of the cases, ion irradiation induces damage in the material, which may be deleterious for its intrinsic physical and chemical properties for which it has been chosen. In the case of low- and medium-energy (tens of keV to few MeV) ion irradiation, it has been shown by RBS/C that the damage buildup occurs in several steps as a function of the accumulated ion fluence [1]. Each step is characterized by the presence of a predominant type of defects. The transition from one step to the following one seems to be triggered by radiation-induced defect re-organization which leads the material to a more stable energetic state [2]. Therefore, the ability to monitor the microstructural transformations and to determine the key parameters that trigger them definitely appears as important issues for a comprehensive knowledge of the behavior of nuclear materials under irradiation. The radiation-induced defects generate elastic strains and stresses which can in turn play a role on the defect behaviour. In the case of crystalline structures, XRD is one of the most relevant techniques to determine the strain/stress state of materials. It is here demonstrated that XRD measurements combined with RBS/C and TEM experiments allow getting a comprehensive description of the behavior of nuclear materials under irradiation. To illustrate this statement, two test case nuclear materials, namely cubic zirconia and urania, have been studied. XRD allowed determining both the strain/stress state of irradiated materials, and the variation of the strain with the irradiated depth. These depth strain profiles have been compared to SRIM-predicted and experimental damage (channeling) and implanted specie (RBS) distributions. Furthermore, the possibility to monitor the damage level as a function of the ion fluence has been established [3]. A multi-step damage buildup, similar to the one obtained by RBS/C, has been found. Defect clusters inducing large strains are formed in the first step, while dislocation loops develop at higher fluences to relieve the stored elastic energy. Thus, this elastic energy term seems to drive microstructural transformations under irradiation. [1] K.E. Sickafus et. al., J. Nucl. Mater. 274, 66 (1999).[2] J. Jagielski et. al., Appl. Phys. A97, 147-155 (2009).[3] S. Moll et. al., J. Appl. Phys. 106, 073509 (2009).
5:15 PM - BB7.8/AA8.8
Real-time Observation of Radiation Damage in Zircon by High-energy Electron Beams.
Nan Jiang 1
1 Physics, Arizona State University, Tempe, Arizona, United States
Show AbstractRadiation effects in zircon have been extensively studied because of the potential application of zircon for immobilization of plutonium. In natural zircon, which contains traces of uranium and thorium, α-decay events degrade the zircon structure resulting in a crystalline-to-amorphous transformation. The same transformation has also been observed and studied in Pu-doped zircon, neutron irradiated and ion-beam irradiated zircon. The amorphization process is generally believed to result from the spontaneous local collapse of the crystal structure due to a local high defect concentration. The direct observations of the crystalline-to-amorphous transformation were based on transmission electron microscopy (TEM) observations. In high-resolution electron microscopy (HREM) images, regions with mottled contrast were observed, and attributed to amorphous or disordered volumes in the lightly damaged zircon, while small crystallites were seen in an amorphous matrix in the heavily damaged zircon. Most interestingly, the HREM images showed that microstructural evolution in the self-irradiated natural zircon was very similar to that in the ion-beam damaged zircon, although the damage accumulation rate in the former was several orders of magnitude lower than that in the ion-beam irradiation experiments. The conclusions largely supported and encouraged the use of ion-beam irradiation to simulate the α-decay and fission track damage effects in zircon. Nevertheless, the atomic-level structural changes and the structure of damaged “amorphous” zircon are still not well understood. In review of various data obtained from a variety of analytical techniques, information about the composition of “amorphous” zircon, at least at the nanometer scale, was missing. Does heavily damaged zircon have the same composition as undamaged zircon? Therefore radiation damage by high-energy (200 keV) electrons in zircon has been studied thoroughly using imaging, diffraction and electron energy-loss spectroscopy (EELS) techniques in TEM. Both structural and compositional changes during the damage were measured using the above techniques in real time. It is found that the electron-beam induced damage in zircon is more or less similar to that induced by ion-beam irradiation or even α-decays in natural or Pu-doped Zircon. The appearance of the patches is very similar to that observed as mottled diffraction contrasts in α-decay-damaged and ion-beam irradiated zircon. However, the damage by electron beam was mainly caused by the preferential sputtering of O. The loss of O occurred initially within small sporadic regions with dimension of several nanometers, resulting in the direct transformation of zircon into ZrxSiy. These isolated patches gradually connect each other, and eventually cover the whole area of the electron beam.
5:30 PM - BB7.9/AA8.9
Key Role of the Cations Interstitial Structure in the Radiation Resistance of Pyrochlores.
Alain Chartier 1 , Gilles Catillon 2 , Jean-Paul Crocombette 3
1 , CEA-Saclay, Gif-Sur-Yvette France, 2 , Université Paris-Est, G2I, Marne la Vallée France, 3 DEN/DANS/DMN/SRMP, CEA-Saclay, Gif-Sur-Yvette France
Show AbstractAmong the structurally related fluorite compounds, the pyrochlore oxides (A2B2O7) have many technological applications due to the variety of their properties. A fundamental understanding of their response to irradiations is also of utmost interest for the nuclear industry as they can be used as an inert matrix. We focus here on the Gd2ZrxTi2-xO7 (0 < x < 2) solid solution as it is a textbook example of the response of pyrochlores under irradiations. Its response varies widely with composition, from no amorphization in Gd2Zr2O7 at any temperature (down to 30 K) to complete amorphization in Gd2Ti2O7 up to a so-called critical temperature, beyond which no amorphization is possible.We show in the present work [1] that continuous Frenkel pairs’ (FP) accumulation within molecular dynamics can quantitatively reproduce the irradiation response of the whole solid solution Gd2ZrxTi2-xO7, as a function of composition and temperature. The calculated critical temperatures TC for amorphization are in quantitative agreement with the experimental measurements.In fact, the decrease of the radiation resistance as a function of the increase of Ti content in Gd2ZrxTi2-xO7 is shown to be driven by a change in the configuration of the B interstitials. The stabilization of Ti-Ti dumbbells in the partially disordered fluorite structure hinders their recombination with vacancies. Conversely, the isolated Zr interstitials readily recombine with vacancies. So, as the amount of Ti increases in the B site of Gd2B2O7, the stability of the B-B dumbbells increases in the fluorite structure and prevents the reconstruction of the structure which eventually collapses to the amorphous state. Present calculations also unify the previous ground states based criteria from Sickafus [2] and introduce the recombination of the B-B dumbbells as the missing phenomenon in the rate equations.References [1] A. Chartier, G. Catillon, and J.-P. Crocombette, Phys. Rev. Lett. 102 (2009) 155503.[2] K. E. Sickafus, L. Minervini, R.W. Grimes, J. A.V. Ishimaru, F. Li, K. J. McLellan, and T. Hartmann, Science 289, 748 (2000).
5:45 PM - BB7.10/AA8.10
In-situ Observations of Ion-induced Phase Transformations Using a Reflection High-energy Electron Diffraction (RHEED) System.
Jonghan Won 1 , James Valdez 1 , Igor Usov 1 , Kurt Sickafus 1
1 Materials Science and Technology Divison, Los Alamos National Laboratory, Los Alamos, New Mexico, United States
Show AbstractA novel method was developed to monitor in-situ ion irradiation damage in crystalline materials. The new method uses reflection high-energy electron diffraction (RHEED) to collect electron diffraction patterns while performing ion irradiation. In our RHEED experiments, we use a ~50 keV electron beam impinging on a crystalline substrate at grazing incidence. The diffracted electrons are collected on a phosphor screen with a cold cathode device. In this presentation, we report on ion irradiations of a single crystal (LaAlO3). Irradiations were performed using 100 keV Ar ions at room temperature. Before the single crystal was bombarded with Ar ions, a diffraction pattern with strong diffraction maxima was observed. During the Ar ion irradiation, electron diffraction patterns were collected from the near surface region (continuously). The initially strong diffraction patterns gradually faded in intensity until no diffraction spots were visible on the phosphor screen at the fluence of 6-8x1014 ions/cm2. To confirm the presence of the ion-induced amorphization in the near surface region, we used grazing incidence X-ray diffraction (GIXRD) and electron diffraction in a transmission electron microscope (TEM). Both characterization techniques corroborated the RHEED in-situ measurement, showing the existence of an amorphous region produced by the 100 keV Ar ions. Results from this investigation showed that by monitoring RHEED patterns during ion irradiation, it is possible to determine precisely threshold conditions for changes in material crystal structure, as a function of ion irradiation fluence, temperature, etc.